Preface
Heat and light are types of radiation that people can feel and see, and therefore the two have been recognized "elements" of the Universe for a long time. Contrary to this, its prime "element", radioactivity, which results in radiation that human senses cannot detect, was been discovered only a century ago. Ever since its discovery, the influence of radioactivity on man and his well-being has been in question. After a period of great scientific accomplishments, which led to the understanding of the atom, came the bomb. This was the first big scare. Later development of technology, especially in energy production and weapon testing, has led to occasional contamination of the environment by radioactivity. This book is based on the author's experience gained by working in institutions where radioactivity was an important issue: research laboratories, both in institutes and universities, as well as in the International Atomic Energy Agency in Vienna. The book starts with a chapter describing the occurrence of radioactive nuclides in nature, followed by chapters on man-modified and man-made radioactivity. The big challenge has always been the measurement of radioactivity. New detection methods are still being developed nowadays, all based on the physics of radiation interaction with matter. Although radiation has found applications in almost all aspects of human activities, most of the ionizing radiation that people are exposed to still comes from natural sources. The health effects of radiation are relatively well understood and can be effectively minimized through careful safety measures and practices. Radiation sciences continue to push forward the frontiers of understanding and to expand the capabilities of technology. The scare of the bomb, however, still remains with us. It is good to know that at least two international organizations, IAEA and CTBTO, are taking care that we continue to use radioactivity for the improvement of the quality of life only. In the preparation of this book the author has used material from many of his colleagues. Whenever the reference to their work was available it has been mentioned in the text. Some of the information has been left with the author based on numerous discussions and meetings only. In the final presentation of the material I was assisted by Miss Jasmina Obhodas who prepared all the drawings and Mrs. Ljiljana Liscevic who did all of the typing required. The relevant chapters have been scrutinized at the IAEA and CTBTO and I am grateful for their very useful remarks. Ms Reina Bolt and her team at Elsevier have done the final touch. Finally, I should thank my wife Georgia for understanding my late comings home.
CHAPTER 1
Introduction
Radioactivity is a part of nature--in the process of element formation by nuclear reactions taking place in stars, both stable and radioactive isotopes of elements are formed. The isotopic composition of elements is characterized by properties of nuclear reactions that led to the formation of the elements. Elemental composition of the planet Earth, thought to be about 4.5x 109 years old, although not yet in chemical equilibrium, reflects the composition of the material from which it was formed. Therefore, a number of radionuclides occur in nature, having long half-lives (longer than the age of Earth). In addition there are natural processes which continuously produce new radioisotopes. Recently, human activities have also contributed to the increased concentration of some of the radionuclides. Ever since its discovery, the influence of radioactivity on man and his wellbeing has been in question. The development of technology, especially the use of nuclear processes like fusion and fission, in both energy production and weapon manufacturing and testing, has led to occasional contamination of the environment by radioactivity. Numerous sources of ionizing radiation can lead to human exposure: natural sources, nuclear explosions, nuclear power generation, use of radiation in medical, industrial and research purposes and radiation-emitting consumer products. Before assessing the radiation dose to the population, one requires a precise knowledge of the activity of a number of radionuclides. The basis for the assessment of the dose to the population from a release of radioactivity to the environment, the estimation of the potential clinical health effects due to the dose received and, ultimately, the implementation of countermeasures to protect the population is the measurement of radioactive contamination in the environment after the release. The types of radiation one should consider include: 1. Alpha radiation which consists of heavy positively charged particles emitted by atoms of elements such as uranium and radium. Alpha radiation can be stopped completely by a sheet of paper or by the thin surface layer of our skin (epidermis). However, if alpha-emitting materials are taken into the body by breathing, eating, or drinking, they can expose internal tissues directly and may, therefore, cause more biological damage.
2
Chapter 1
2. Beta radiation which consists of electrons. They are more penetrating than alpha particles and can pass through 1-2 centimetres of water. In general, a sheet of aluminum a few millimetres thick will stop beta radiation. 3. Gamma rays which are electromagnetic radiation similar to X-rays, light, and radio waves. In general, gamma rays, depending on their energy, can pass right through the human body, but can be stopped by thick walls of concrete or lead. 4. Neutrons which are uncharged particles. Therefore, they do not produce ionization directly. But, their interaction with the atoms of matter can give rise to alpha, beta, gamma, or X-rays which then produce ionization. Neutrons are penetrating and can only be stopped by thick masses of concrete, water, or paraffin. The two basic quantities in the assessment of radiation levels and effects are the activity of a radioactive material and the radiation dose. The activity of a radioactive material is the number of nuclear disintegrations per unit time, unit becquerel (Bq). One becquerel is one disintegration per second. The term radiation dose can mean several things (e.g. absorbed dose, dose equivalent or effective dose equivalent). The absorbed dose of radiation is the energy imparted per unit mass of the irradiated material. The unit of absorbed dose is joule&g, for which the special name gray (Gy) is used: 1 rad = 0.01 joule/kg = 0.01 Gy. It is the purpose of this book to present the facts about the presence of radionuclides in nature. The use of technology can significantly modify the exposure to natural radiation. Among the human activities which should be considered in this context are: (i) the electricity generation by coal-fired power plants, (ii) the use of phosphate fertilizers, and (iii) many consumer products. Man-made radioactivity has found many useful applications in everyday life. The best known are medical applications. The use of radionuclides and radioactivity in diagnosis and treatment of diseases is well established practice. The determination of radionuclides in environmental samples is an important task in relation to the protection of human health. This is especially the case when there is an accidental release of radioactivity into the environment, as was the case with the Chernobyl accident. Assessment of the situation requires knowledge of the type and form of released radionuclides and reliable, practical techniques for the analysis of various radionuclides. All the instruments for the detection and measurement of radioactivity are based on the physical processes of radiation interaction with matter. The understanding of these processes has led to the development of many types of radiation detectors. The methods used for the measurement of radionuclide concentrations in various matrices are numerous. Some are better then others, but the best is always a combination of several techniques. With the development of understanding of radiation interaction with living matter, better safety standards have been developed. Dosimetry is an important factor in the beneficial use of radiation.
Introduction
3
The most frightening experience in the history of humankind was the use of the nuclear bomb. The development of this device has resulted in many disturbed sites on our globe. It has also left us with the fear that this powerful weapon could be developed and used again. Mankind is trying to prevent this through the international treaties being signed by a majority of states. Let us hope we shall continue to live in this radioactive environment by using radioactivity to improve the quality of our lives.
CHAPTER 2
Radioactive Nuclides in Nature
Except for the simplest nucleus, that of hydrogen, all other nuclei consist of neutrons and protons. The ratio of neutrons to protons is unity for lighter isotopes on the so-called "stability line" and increases gradually as one approaches the behaviour elements at the end of the periodic table or moves off the "stability line". As this ratio increases, a stage is reached where the nuclide is no longer stable. The heaviest stable nuclide is 209v.~. 83 t~l. Nuclides heavier than this are unstable because they have excess energy to dissipate. Unstable nuclides are called radionuclides and they dissipate their surplus energy by the emission of radiation. This process is called radioactivity or radioactive decay. The decay of a radionuclide is a statistical process in the sense that it is not possible to predict exactly when a particular nucleus will disintegrate. One may, however, ascribe a probability that a nucleus will decay in unit time. This probability is known as the radioactive decay constant (transformation constant), k, of the radionuclide. The number of atoms of a radioactive substance disintegrating per unit time, dN/dt, which is referred to as the activity of the substance, is proportional to the total number, N, of radioactive atoms present at time t; the constant of proportionality being k. Thus, dN dt
=LN
(2.1)
Integrating this equation, one obtains N = Noe-x'
(2.2)
where N Ois the initial number of radioactive atoms present, and N, as already stated, the number of radioactive atoms at time t. Rewriting eq. (2.1), it follows that dN -~ dt
= kN
= kN
oe-x'
(2.3)
6
Chapter 2
Equation (2.3) indicates that the number of radioactive atoms present as well as the disintegration rate (activity) decrease exponentially with time. The time taken for half the radioactive atoms originally present to decay is called the half-life of the radionuclide. Substituting N = N O/2 and t = tl/2 in eq. (2.2), one obtains N 0 / 2 = N 0e-~'''2
(2.4)
or
~,t~/2 = ln2 = 0.693
(2.5)
or
tl/2 = 0.693/X
(2.6)
The number of radioactive atoms present and hence the rate of disintegration decreases to one-half in one half-life, to one-quarter in two half-lives, to one-eighth in three half-lives, and so on. The half-life is characteristic of any particular radioisotope. Another useful quantity is the mean life or the average life of a radionuclide which is the reciprocal of the decay constant, t m = 1/X. A radionuclide, upon undergoing disintegration of a particular type, yields a specified nuclide. The original radionuclide is called the parent and the decay product is called the daughter. The daughter may also be a radionuclide. A succession of nuclides, each of which transforms by radioactive disintegration into the next until a stable nuclide results, is called a radioactive series. Examples of such series are the uranium series and the thorium series. Radioactive equilibrium refers to that state in which the ratios between the amounts of successive members of the series remain constant. Under these conditions the disintegration rates of the parent and all the subsequent radioactive daughters will be the same. A steady state implies that the activity of the daughter and its precursor are equal and do not change in time. However, it is easy to see that in a finite time interval such a state of so-called secular equilibrium can only be approached and never be reached due to the fact that the primordial nucleus can never have a steady state. As a result its daughter product can never reach a steady state, etc. On the other hand, when ~-/~i-~ >> 1 a state of quasi equilibrium may be reached where the activity ratio of mother and daughter nuclei is given by Ai -
Ai_l
~i
(2.7)
~ i -- ~ i-1
Wherever in the following the expression secular equilibrium is used it should be understood that we refer to this state of quasi-equilibrium. Furthermore, we prefer to
Radioactive Nuclides in Nature
7
express the abundances of nuclides in terms of activity ratios rather than in concentration ratios. The reason for this is that activity ratios can always be expressed as convenient numbers (between zero and one in a developing decay series), whereas the ratio between concentrations is proportional to the half-lives of the nuclides, often yielding inconveniently large or small numbers. When the production and removal of nuclei in a radioactive decay series is the result of radioactive decay only, the time development of the number of nuclides N i of any isotope i in the series is given by competition between its radioactive decays dNi dt
: - ~,i N i -k- ~ i _ l N i_ 1
(i = 1..... n)
(2.8)
with the boundary condition N; (t = 0) = N~o
(i = 1,. ..,n)
(2.9)
is the decay constant of the i-th member in the series (i.e. the probability per unit time that a nucleus will decay, related to the half-life tl/2 by the expression ~,i = ln(2)/t~/2). By definition the n-th member (i.e. the end member) is stable, hence ~,~ = 0. Furthermore, i = 1 denotes the primordial nucleus of the chain (which implies that N Odoes not exist). Equations (2.8) and (2.9) describe the time development of the number of nuclei of any isotope i in a radioactive decay series by means of n coupled linear inhomogeneous differential equations. The general solution of any of these equations is the summation of the general solution of the homogeneous equation dN/dt + )~N i = 0
(2.10)
given by N i = Ci, i e -~.,t
(2.11 )
and a particular solution of the inhomogeneous equation, for which we use the trial function i-1
(2.12)
N i ( t ) = ~_~Ci, je -;~'t
j-1
Substituting expression (2.12) in eq. (2.8) and solving for thej-th term yields = C i,j = 0
C;_~,j
(j < i)
(j > i)
while boundary condition (2.9) yields
(2.13a) (2.13b)
8
Chapter 2
i-1
C i , j -- N i0 -
~C
(2.13c)
i,j
j-1
Hence the general solution to (2.8) is given by i
(2.14)
N i ( t ) = ~_~Ci, je -~ j=l
with C~,~determined by the recursive relations in eq. (2.13). This may be conveniently written in a matrix equation N(t) = M . E(t);
N ( t = 0) = N o
(2.15)
where N(t) = {N~(t), N2(t) ..... N,(t)}
(2.16a)
E(t)- {e-X't, e -x2t , . . . ,
(2.16b)
e -x"' }
(2.16c)
N o = {Nl(0), N2(O) ..... Nn(O)}
C
m
Cl,1
0
C2,1
C2, 2
..... 0
0
.....
0
0
~
o
o
9
~
~
'
(2.16d)
__
C .-1,1
C .-1.2
C._ 1
C., 2
..... .....
C .-1, n-I
0
C ..... l
C.,n
If the abundances of nuclides present in a decay series are only subjected to the law of radioactive decay (no chemical or other physical processes are involved), the development in time to a state of quasi-equilibrium is governed by eq. (2.8), no matter how complicated the initial conditions are. If, for any reason, this state of equilibrium has not yet been reached and the initial abundances, given by eq. (2.9), of the various nuclides are known, the elapsed time can be deduced from the degree of disequilibrium. Radionuclides which can be found in the environment can be divided into three groups: 1. Naturally occurring nuclides of very long half-life which have persisted since the formation of the Earth, and their shorter lived daughter nuclides. 2. Naturally occurring nuclides which have short half-lives on the geological time scale, but which are being continuously produced by cosmic-ray radiation. 3. Radionuclides released into the environment due to man's activity and accident.
Radioactive Nuclides in Nature
9
2.1 LONG-LIVED RADIONUCLIDES AND THEIR SERIES Some of the long-lived naturally occurring radionuclides are shown in Table 2.1. Some elements in this table result in non-negligible doses to man. For example, potassium which is an essential element is under close homeostatic control in the body. The average mass concentration for an adult male is about 2 g of potassium per kg of body weight. The isotopic ratio of 4~ is 1.18 10 -~ and the average activity mass concentration of 4~ in the body is about 60 Bq kg -~. The radionuclide is both a beta and gamma emitter; consequently the whole body is uniformly irradiated. 4~ is the principal naturally occurring source of internal radiation arising from ingestion. Stable potassium enters the body mainly via foodstuffs at the rate of about 2.5 g per day. Specific locations in the body where potassium is preferentially concentrated (such as the bone marrow) receive the highest doses. Other nuclides in Table 2.1 which are of particular interest are uranium and thorium isotopes and their series. Uranium 238, uranium 235 and thorium 232 decay series are schematically presented in Fig. 2.1. A number of radionuclides are formed during these Table 2.1 Some natural radionuclides with long half-lives Radionuclide
Half-life
Specific activity/
(109 years)
Ci (g of element)-1
4~
1.27
8.3•
-1~
5~V
6•
2.8•
-14
87Rb
47
2.5•
llSIn
6•
5•
-8 -12
138La
110
2.1xlO -12
142Ce
6x10 6
5.7x10 -14
1475m
110
3.4x10 -9
1485m
1.2x 10 4
2.2x 10-l~
1495m
4x105
8.2x10 -13
152Gd
1.1xl05
4.1x10 -12
Radioactivity 13
[3 13 [3
c~
o~
I74Hf
4.3x 106
8.4x 10-14
(X
144Nd
5x106
1.2x10 -13
c~
19~
700
3.3x10 -13
ot
192pt
106
1.4x10 -14
O~
2~
1.4x 10 8
1.8x 10-16
(X
232Th
14
1.1 x 10-7
o~
235U
0.71
1.5x 10-8
O~
238U
4.5
3.3 x 10 -7
Of,
10
-iII
,-IIi
II
,
i
,
,_..,
~o
\
[,,-, ~
F-
\ ,,,,,= '-~::~
" ~~
\
v
I"
,... ,--,
r
,
,,.
\ ~E [.T.
\ 9,--' E
o~
0"~ I~,,,, o o
v
7~
\
\
\
\
"
E e-i
E
\
\
E
.=:
2
r
Chapter 2
_LEd
0
r
r
r
.=
r
r r
o
0
r o
0
r 0 .,..~
~
Radioactive Nuclides in Nature
11
Table 2.2 Masses of the various daughters in secular equilibrium with 1 g of 238U Isotope
Mass (g)
238U
1.0
234Th
1.4x10 -ll
234mpa
4.8x10 -16
234U
5.4x10 -6
23~
1.8x 10-5
226Ra
3.3x 10-7
222Rn
2 . 2 x 1 0 -1~
218po
1.2x10 -15
214pb
1.0•
-14
214Bi
7.4•
-15
212pb
4.1•
-9
21~
2.7•
-12
21~
7.4•
-11
decays. For the illustration, Table 2.2. shows the masses of the various daughter nuclei in secular equilibrium with 1 g of 238U. Of special interest are the gases radon (222Rn) and "thoron" (22~ which are formed as progeny of uranium and thorium in rocks and soil. They are emitted from the ground into the atmosphere, where they decay and form daughter products, isotopes of polonium, bismuth and lead, which either remain airborne till they decay, or are deposited in rain and by diffusion to the ground. Let us discuss uranium (238U) and thorium (232Th) decay chains in some detail.
2.1.1 Uranium decay chain The 238Udecay chain comprises eleven discrete decay steps to stable lead. During decay each nuclide will emit characteristic radiation. The decay will result in the emission of alpha particles, beta particles and gamma photons with characteristic energies and probabilities of emission. The emission may involve one energy or a mixture of energies. The majority of the nuclides in the chain have short half-lives; only five nuclides have half-lives exceeding one year: 238U, 234U, 23~
' 226Ra ' 21~
The half lives range from 22.3 to 4.5• have half-lives longer than 30 minutes:
years. Of the remaining nuclides only three
12
222Rn,
Chapter 2
2~~ and 21~
Their half-lives range from 3.82 to 138 days. There are eight nuclides which decay primarily through alpha emission: 238U, 234U, 230Th ' 226Ra ' 222Rn ' 218p0 ' 214p0 ' 21~
These emissions are either pure alpha emissions or accompanied by insignificant amounts of gamma photons. Each alpha is emitted with a unique characteristic energy; for the above nuclides the energy range is between 4.2 and 7.69 MeV. Alpha particles have a limited range in air of the order of 2.5 cm at 4 MeV and 6 cm at 7 MeV. An alpha-emitting radionuclide may emit particles with only one energy, for example: 222Rn 5.49 MeV,
218po 6.00 MeV,
214po 7.69 MeV
or may emit alpha particles with slightly different energies and a characteristic probability of emission for each, e.g." 234U4.72 MeV (28%) and 4.77 MeV (72%); 226Ra 4.78 MeV (95%) and 4.60 MeV (6%). In terms of their potential radiological hazards the above alpha emitters are divided into five long-lived alpha-emitters i.e. 238U, 234U, 230Th ' 226Ra ' 21~
and radon and its two short-lived alpha-emitting daughter products i.e. 222Rn
and 218p0, 214po.
There are other alpha-emitting nuclides in the 238U decay chain but their contribution during decay is minute and for hazard and hazard assessment purposes they are not of significance. 2~~ although a gamma emitter, is also included with the long-lived alpha emitters as it rapidly decays to 21~ an alpha emitter. There are six nuclides whose decay is accompanied by significant beta emissions these are: 234U, 234mpa ' 214pb ' 214Bi ' 21~ ' 21~
The following are daughter products of 226Ra: 214pb ' 214Bi ' 2J~ ' 21~
Beta particles are emitted in the form of a spectrum with a characteristic maximum energy (ernax) and average energy; again, particles may be emitted with different energies each with their own characteristic probability, for example:
13
Radioactive Nuclides in Nature
214pb:
214Bi:
0.65 MeV 0.71 MeV 0.98 MeV
(50%) (40%), (6%);
and
1.0 MeV 1.51 MeV 3.26 MeV
(23%), (40%), (19%).
and
The energy range for the emax value in the complete 238Udecay chain is very wide and ranges from 0,02 MeV to 3,26 MeV. There are three beta emitters in the 238Udecay chain which by virtue of the energy of emission and abundance contribute a large proportion of the total beta emissions: (i) Protactinium 234pam: In terms of its emax value (2.29 MeV) and overall abundance (98%) 234mpa is the strongest single beta emitting nuclide in the uranium decay chain. It also emits gamma photons of low energy < 0.1 MeV and abundance <4%. (ii) 226Radecay products: Another group of strong beta emitters are associated with the daughter products of 226Ranamely" 214pb,214Bi,21~ 21~ Their emax values range from 0.02-3.26 MeV. Of this group 2~~ and 2~4Biare the strongest and most abundant. 2~~ is an abundant emitter of low energy beta particles of less than 0.1 MeV emax;beta energies below this value are not considered of radiological significance. 214pbemits three betas with an emax from 0.65-0.98 MeV. (iii) Bismuth (214Bi)" This emits betas with three energies and abundances: 1.0 MeV (23%), 1.51MeV (40%), and 3.26 MeV (19%). 2~~ emits betas with a single emax of 1.161 MeV at 100% abundance. The majority of alpha and beta particle decays are accompanied to a greater or lesser degree by the emission of gamma photons. In general alpha emitters have very low levels of or no gamma emissions associated with them. For example 226Ra which is primarily an alpha emitter also emits 0.186 MeV gamma photons of low abundance (4%). In many cases of gamma photon emission the energy of gamma emission or the percentage abundance is very low and therefore not of great significance in terms of external exposure. The dominant contribution to gamma emissions in the 238U decay chain occurs in that part of the decay chain below 226Ra.Over 95 % of the total gamma emissions in the chain occur in this sub series of the chain. A small contribution (a few percent) is made by 2~4pb but the overwhelming majority (90%) is contributed by 2~4Bi. Its abundant emissions range in energy from 0.2 to 2.5 MeV and are highly penetrating. Those nuclides prior to 226Rain the chain are very weak gamma emitters when compared to 226Ra and its decay products. Thus it is the short-lived daughter products of radium trapped in soils and scales and other materials that delivers the majority of the gamma doses to individuals in gold mines. Therefore the higher the concentration of 226Raper
14
Chapter 2
gram and the greater the amount of material present at one location the more intense the gamma field will be at a fixed point from a material.
2.1.2 Thorium decay chain Thorium is normally present in low concentrations in gold-bearing ores; its presence is indicated by detectable concentrations of thoron daughters in return air in most gold mines that are, or were, uranium producers. In some areas thorium is present in elevated concentrations in a layer of black mineral sand above the gold bearing reef. Thorium or its decay products have also been detected in some radium bearing scales in acid plants. The parent 232Th decays to stable 208 through nine radioactive daughters. With the exception of the parent 232Th (half-life 1.4x 10 l~ years) the daughter half-lives are all less than seven years. 228Th and 228Ra have half-lives measured in years (1.9 to 6.7 years). The remaining nuclides' half-lives varying from nanoseconds to 10.64 hours. There are seven radionuclides which decay primarily through alpha emission. The most important long-lived alpha emitters are Z32Thand 228Th. 228Ra, with a half life of 3.64 days, is not usually included as a long-lived alpha emitter when unsupported by 232Th; however, if incorporated into the body it has some radiological significance. Thoron gas (22~ decays producing a number of short-lived alpha emitters of radiological significance: 216po, 2~ZBiand 2~2p0. Most of the alpha emitters of the 232Th decay chain are pure alpha emitters or have low levels of associated gamma emission. An exception is 2~2Bi which has significant amounts of associated beta and gamma emissions. These have a wider range and higher maximum than the 238Udecay chain and range from 3.95 to 8.78 MeV in energy. There are five radionuclides whose decay is accompanied by significant beta
emissions these are: 228Ra, 228Ac, 212pb, 212Bi and 2~ The following are daughter products of 224Ra" 2~2pb, 212Bi and 2~ Beta particles are emitted in the form of a spectrum with a characteristic maximum energy (emax)and average energy; again particles may be emitted with different energies each with their own characteristic probability. The energy range for the emaxvalue in the complete 232Thdecay chain is very wide and ranges from 0.06 MeV to 2.26 MeV. The majority of alpha and beta particle decays are accompanied, to a greater or lesser degree, by the emission of gamma photons. In many cases of gamma photon emission, the energy of gamma emission or the percentage abundance is very low and therefore not of great significance in terms of external exposure. The dominant contribution to gamma emissions in the 232Th decay 224,,-~ chain occurs in that part of the decay chain below r~a. Over 95% of the total gamma emissions in the chain occur in this sub series of the chain. The dominant gamma emitter in the chain is 2~ Its abundant emissions range in energy from 0.511-2.614 MeV. Due to their abundance and high energy, any materials---e.g, scales~in which this radionuclide is concentrated, produce very localized, intense and strong gamma fields.
Radioactive Nuclides in Nature
15
2.2 C O S M I C - R A Y - P R O D U C E D RADIONUCLIDES The second group includes radioisotopes produced by cosmic rays. The rates of production of radioactive isotopes can be estimated reasonably well from the energy spectra of primary and secondary cosmic rays and a knowledge of the corresponding nuclear reaction cross sections. Cosmic rays are conventionally divided into two classes" primary and secondary. The former are, for the most part, energetic charged particles of extra-terrestrial origin, while the latter are the products resulting from collisions of the primary cosmic rays with the atoms of the Earth' s atmosphere. After the inevitable interaction of a primary particle with some atom high in the atmosphere, the nuclear debris produced undergoes successive interactions with atoms further down in the atmosphere. In such collisions ~-mesons are created and they decay into muons and ),-rays. Muons continue moving down to sea level before they decay. On the other hand, the gamma rays produce electron-positron pairs which in turn radiate more ),-rays. The huge number of electrons created in this way are called extensive air showers. The secondary radiation led to the discovery of new particles, including the positron and various mesons. The collision of secondary neutrons with the atmospheric nitrogen produces carbon-14 (~4C) which combines with oxygen to form radioactive ~4CO:. This provides a clock for the familiar technique of radioactive dating. The total number of primary cosmic rays striking the Earth' s atmosphere is roughly 104 m -2 s-1. Although they were discovered in 1911, the origin of cosmic rays still remains a mystery to astrophysicists (Hayakawa, 1963; Ginzburg and Syrovatskii, 1964). In older theories about the origin of cosmic rays it was suggested that they underwent multiple scattering within clouds of magnetized plasma in interstellar space and so were accelerated to high energies. Present-day speculation considers the sources to be violently active celestial objects such as supernova explosions, rapidly spinning neutron stars and white dwarfs. The components of the primary cosmic rays include nuclei, electrons and electromagnetic radiation. The nuclear component of the primary cosmic rays comprises (at the top of the atmosphere) about 90% protons, 9% He, 1% heavier nuclei, and almost zero antinuclei. Comparing the distribution of elements in cosmic rays with the distribution throughout the Universe, the two are rather similar. However, there are very important differences, for example, the relative abundances of H and He are much lower in cosmic rays, while Li, Be, B and odd Z nuclei are much more abundant in cosmic rays. It is assumed that they are produced by transmutation during the passage of heavier nuclei through the interstellar medium. The abundances of the nuclei at the end of the periodic table are greatly enhanced in cosmic rays--a fact that has to be considered in theories on the origin of cosmic rays. The energy spectrum of a nuclear component extends smoothly from about 10 MeV to 102~eV (more than 13 decades in energy and 32 decades in intensity). The mean particle energy of the galactic cosmic-ray spectrum is about 1 Gev, and the number density of these particles in interstellar space is about 10 -3 m -3, almost equal to the
16
Chapter 2
Table 2.3 Half-lives and decay characteristics of cosmic-ray produced radionuclides Radionuclide
Half-life
Main radiation
1~
1.6xl06 years
13555 keV
26A1
7.2x105 years
~+ 1.17 MeV; Y 1.81 MeV, 511 keV
36C1
3.00x 105 years
[3 714 keV
8~
2.13x 105 years
K x-ray
14C
5730 years
[3 156 keV
32Si
--650 years
13210 keV
39Ar
269 years
13565 keV
3H
12.33 years
[3 18.6 keV
22Na
2.60 years
13+0.545, 1.82 MeV; 7 1.275 MeV, 511 keV
365
87.4 days
13 167 keV
7Be
53.3 days
E.C., Y 477 keV
37Ar
35.0 days
K-x-ray, Bremsstrahlung to 0.81 MeV
33p
25.3 days
13248 keV
32p
14.28 days
13 1.710 MeV
28Mg
21.0 hours
130.459, Y 1.35, 0.31, 0.95, 0.40 MeV
24Na
15.02 hours
13 1.389 MeV; Y 1.369, 2.754 MeV
38S
2.83 hours
[3 3.0, Y 1.88 MeV; 7 1.6, 2.17 MeV
31Si
2.62 hours
[3 1.48 MeV; Y 1.26 MeV
16F
109.8 minutes
[3§ 0.635 MeV; 511 keV
39C1
56.2 minutes
[3 1.91 to 3.45 MeV; 70.246, 1.27, 1.52 MeV
38C1
37.29 minutes
13§ 4.91 MeV; Y 1.6, 2.17 MeV
34mc1
31.99 minutes
[3+ 2.48 MeV; e- 0.142 MeV; Y 1.17, 2.12, 3.30 MeV; 511 keV
energy density of the electromagnetic radiation and that of the magnetic fields. The energy spectrum falls off more gently than the thermal distribution; intensity at energy E is proportional to approximately E -z6. Low energy cosmic rays entering the solar system are convected away from the Sun by the solar wind. At lower energies solar flares contribute most of the particles, which tend to mask the galactic contribution. During periods of maximum sunspot activity, solar flares sporadically "contaminate" the solar system with these low energy particles. The tracks and induced radioactivity that these particles have been found to produce just below the surface of samples of Moon rock, indicate that flares have been a
Radioactive Nuclides in Nature
17
regular feature of solar activity for at least millions of years. Solar-flare particles have a steeply falling energy spectrum and diffuse out of the solar system in a few days. Almost all the elements from hydrogen up to nickel have been detected. The experimental relative abundances of these emitted nuclei bear a strong resemblance to those found in the solar atmosphere in which they originated. There is evidence for preferential emission of heavy nuclei relative to the light ones. During the passage through the atmosphere, a number of nuclear reactions take place, which are responsible for the production of a wide variety of radionuclides (see Table 2.3). Table 2.4 lists all isotopes which occur in the four heavy element decay series. The table also lists a selection of other important radioisotopes which occur naturally or are formed by either fission or irradiation.
Table 2.4 Data on some of the principal radioisotopes Element
Isotope Mass No.
Type of Decay
Half Life
Californium
252
cxy + fission
2.638y
Berkelium
247
o~'f
1380y
Curium
248 246 245 244 243 242
oty o~y o~y o~y o~y o~y
3.39• 4730y 8500y 18.11y 28.5y 162.8d
Americium
243 242 242m 241
cx7 cx7 ~y cxy
7380y 16.02h 152y 432.2y
Plutonium
244 242 241 240 239 238 236
cxy cx7 13(~)y cxy oty o~y o~y
8.26• 107y 3.763• 14.4y 6537y 24065y 87.74y 2.851y
Neptunium
240 239 238 237 235
13y I]Y 13'y o~y cxT
7.4m (meta stable) 65m 2.355d 2.117d 2.14• 3.96.1d
(continued)
18
Chapter 2
Table 2.4 (continuation) Element
Isotope Mass No.
Type of Decay
Half Life
Uranium
240 238 237 236 235 234 233 232
137
oc7
14.1h 4.468x 109y 6.75d 2.3415x107y 7.038x108y 2.445x105y 1.585x 105 y 72y
oc7
1.17m 6.7h 27.0d 3.276x104 y
Protactinium
Thorium
Actinium
234m 234 233 231 234 232 231 230 229 228 227 228 227 225
oc7
oc7 ~7 ~7
~7
24.1 d 1.405x101~ y 25.52 h 7.7x104 y (Ionium) 7340 y 1.9138 y 18.718 d 6.13h 21.773 y 10d
Radium
228 227 225 224 223
5.75 y 1600 y 14.8 d 3.66 d 11.434 d
Francium
223 221
21.8m 4.8m
Radon (or Emanation)
222 220 219 218
Astatine
218 217 216 215 211
~7 ~7
~7 sT ~Y oc, EC, y
3.8235 d 55.65 s 3.96 s 35 ms 2 s (oc94%) 32.3 ms 300 ms 0.1 ms 7.214 h (EC c 55%)
(continued)
Radioactive Nuclides in Nature
19
Table 2.4 (continuation) Element
Isotope Mass No.
Type of Decay
Half Life
Polonium
218 216 215 214 213 212 211 210
~( ]3)7 cx7 o~( ]3)7 cz7 cz7 cz7 ct7 cz7
3.05 m (ct99.9%) 0.15s 1.78x 10-3 s (almost all
Bismuth
214 213 212 211 210 209
13( cx)7 ]3( oOy or( [3)y ~( 13)7 [3( oOy
19.9m 45.65 m 60.55 m 2.14m 5.012 d stable
Lead
214 212 211 210 209 208 207 206
]37 ]37 ~7 ]37 13
26.8 m 10.64 h 36.1 m 22.3 y 3.253 h stable stable stable
Thallium
210 209 208 207 206 205 204 203
l~T ~], 137 1~7 ]]7
1.3 m 2.2 m 3.07 m 4.77 m 4.20m stable 3.779 y stable
Europium
155 154
[37 137
4.96 y 8.8y
Samarium
147
cx
1.06x1011 y (naturally occurring)
Promethium
147
[3
2.6234 y
Praseodymium
144
[37
17.28
Cerium
144 141
[37 ~7
284.3 d 32.51 d
Lanthanum
140
]37
40.2 h
164.3 lxs 4.2 txs 0.305 gs 516 ms 138.38 d
(continued)
Chapter 2
20 Table 2.4 (continuation) Element
Isotope Mass No.
Type of Decay
Half Life
Barium
140
~Y
12.79 d
Caesium
137 136 135 134 132
I]y 13y ~ [37 positron, EC, 7
30.0 y 13 d 2.3x106 y 2.062 y 6.5 d
Iodine
132 131 129 125
137 137 [37 EC
78h 8.04 d 1.7• y 60.14 d
Tellurium
132 129m 125m
[37 137 13
79h 34 d 58d
Antimony
125 124
13y 13Y
2.77 y 60.2 d
Silver
110m
[~y
249.9 d
Rhodium
106
137
29.9 s
Ruthenium
106 103
13 I~Y
368.2 d 39.28 d
Technetium
99
13Y
2.13x105 y
Molybdenum
99
[37
66.69 h
Niobium
95
[37
35.15 d
Zirconium
95
[37
63.98 d
Yttrium
91 90
13 13
58.51 d 64.0 d
Strontium
90
13
29.12 y 50.5 d
89
Rubidium
87
13
4.7x101~ y (naturally occurring)
Zinc
65
I]7
243.9 d
Nickel
63
[37
96 y
Cobalt
60 58 57
137 positron, y, EC
5.271 y 70.8d 270.9 d
Iron
60 59 55
EC, x-ray emission
lxl05 y 44.529 d 2.7y (continued)
Radioactive Nuclides in Nature
21
Table 2.4 (continuation) Element
Isotope Mass No.
Type of Decay
Half Life
Manganese
54
EC, y
312.5 d
Chromium
51
EC, T
27.704 d
Calcium
45
[3
163 d
Potassium
40
~y
1.28x 109 y (naturally occurring)
Chlorine
36
[3
3.01 x 105 y
Sulphur
35
[3
87.44 d
Phosphorus
33 32
[3 13
25.4 d 14.29 d
Sodium
22
positron
2.602 y
Carbon
14
[3
5730 y
Hydrogen
3
[3
12.35 y (Tritium)
2.3 RADIOACTIVITY IN ROCKS AND SOILS Igneous and metamorphic rocks are crystalline rocks formed under conditions of high temperature and pressure, conditions in which rocks can melt and/or recrystallize. Uranium and Th are "incompatible" elements in that their large size does not allow them to fit in well in most mineral structures. As a result they are excluded from the earlier formed high temperature-high pressure minerals and tend to be enriched in alkalic rocks which are late in forming (Gascoyne, 1982). Because U and Th behave in a similar way under these conditions, the ratio between them is usually between 3 to 4 (Th to U) in common igneous rocks. The abundance of both elements tends to increase with content of silica so that granites generally have the highest content of U and Th (although seldom exceeding 30 ppm U) and silica-poor rocks generally have contents of U of 0.1 ppm or less (Rogers and Adams, 1969). Uranium and Th and their daughters, along with radioactive potassium, 4~ are the major heat producers in the crystalline crystal rocks. In the earlier history of the earth, up to perhaps 1.4x 109 years ago, there was almost no oxygen in the atmosphere and uranium oxides could exist at the surface of the earth as grains or nuggets without being oxidized from their 4+ state into the soluble 6+ state. As such, these uranium oxide particles could travel in streams and because of their density they were segregated from less dense materials in streams, in the same way that Au collects in placer deposits. With time these stream deposits were buried, thrust deep into the ground and metamorphosed into the type of accumulation called a quartz
22
Chapter 2
pebble conglomerate deposit. These deposits (which apparently developed between 2.8 and 2.2 billions years ago) are the major source of U in the important economic mineral provinces of South Africa and Canada (Robertson, 1974). For the past 1.4• 9 years or so, U weathered from igneous, metamorphic, or sedimentary rocks generally has been dispersed in the form of the dissolved ion, U +2or one of its complexes. The reaccumulation of U generally has been facilitated by reduction reactions which result indirectly from biogenic processes. The breakup of pre-existing rocks provided the raw materials for the building of new rocks and those built at, or very near the surface of the earth are sedimentary rocks. There are a number of such rocks where U has become enriched. As rocks are uplifted and brought near the surface of the earth, agents of the atmosphere and hydrosphere attack and create a weathered zone where soils can be formed. In one sense, this could be considered a steady state system with inputs of fresh rock below and outputs of dissolved constituents and weathered rock fragments at the surface. With some parent rock types and under specific environmental conditions, the relative tendency of the decay series nuclides to be leached or to accumulate may be enhanced. For instance, the development of terra rossa type soil on limestones can lead to a very strong disequilibrium of U-series radionuclides. Soils from islands in the South Pacific Ocean, the Caribbean and south Florida have been found to have 23~ ratios exceeding 5 and in some cases exceeding 20. A large deficiency exists between the activity of 2~~ as compared to 226Ra which suggests that this soil is a powerful emanator of 222Rn to the atmosphere even though its modest U content would appear to suggest otherwise. The converse of this situation has been measured in caves where the activity of 2~~ exceeds that of 226Ra in the cave floor materials due to the fall-out of the daughters (including 2~~ The majority of the external gamma dose rate above typical soils (95%) arises from primordial radionuclides trapped in the soil. The main contributors to this component are as follows:
4~ 232Th series: 238U series:
35% 50% 15%.
The main contributors to the gamma dose rate in terms of specific radionuclides are 2~ and 228Ac from the 232Th decay chain" while for the 238Udecay series about 99% of the dose rate is due to 214pb and 214Bi. Gamma energies range up to 2.6 MeV and are partly attenuated by the soil with the result that for a typical exposure situation above ground the predominant contribution to the gamma dose rate arises from radioactive material in the top 30 cm of soil. Typical world averages and ranges in normal soils (wet weight) for uranium and thorium are: 238U, 0.025 Bq g-1 (0.01-0.05); 232Th, 0.025 Bq g-1 (0.07-0.05). However, values orders of magnitude higher can occur in oils in specific localized areas of the world.
Radioactive Nuclides in Nature
23
There are two major factors accounting for variability: (a) radionuclide concentrations and (b) shielding. The dose rates will vary according to the geology of an area and the radionuclide composition of the rocks and soil. The highest levels are generally found in igneous rocks and this is related to the quantity of silicates being highest in acidic rocks. In general sedimentary rocks have lower concentrations of radioactive materials than igneous rocks though some shales and phosphate rocks can exhibit significantly elevated levels. The radioactivity in soils is primarily that of the rock from which it was derived, however this can be diminished or augmented by weathering, sedimentation, leaching/ sorption and precipitation from moving ground water, dilution with other materials and increased porosity. Shielding is another factor to be considered. For example 20 cm of snow cover will reduce the gamma dose rate by half. During wet weather, attenuation due to water logging in the topsoil can reduce local gamma dose rates by up to 20%. There are locations of locally elevated terrestrial radiation mainly due to uranium and thorium mineralization. Examples of locally elevated dose rates are: 1. India: In the coastal areas of Kerala and Tamil Nadu in India a narrow coastal strip is rich in monazite, ilmenite, rutile and zircon. In one part of the strip, populated by about 1,000,000 persons, the thorium concentration averages between 8% and 10.5% in patches; the average dose rate for the regions is about 1.3 ~tGy h-', with maxima up to 6 ~tSv h-'. 2. Brazil: In Brazil three coastal towns are built on monazite sands. In one of these, with a resident population of 12,000 and visiting population of 30,000 each summer, average dose rates of 1-2 ~tGy h-' are measured in the streets and up to 20 ~tGy h-' on selected spots on the beach. Also in Brazil, two volcanic regions exhibit high dose rates. In one a mineral called Pyrochlore containing 60% niobium dioxide, 2% thorium and 1.3% uranium oxide produces localized dose rates up to 4 ~tGy h-'; in the second region (Pocos de Caldas), an uninhabited hill has dose rates as high as 28 ~tGy h-'. 3. In Ramsar, Iran, dose rates of 2-50 ~tGy h-' have been reported in a small area characterized by spring water rich in 226Ra. In France, dose rates of 2 ~tGy h-' are not uncommon and very localized values as high as 100 ~tGy h-' have been recorded.
Thus terrestrial radiation levels can vary enormously around the world and result in significant doses, orders of magnitude above the average values reported for man. The effects of soil properties such as pH, soil texture, and organic matter content on the bioavailability of U, Ra, and Th have not been well studied. However, the effects of these soil properties may be deduced by sampling results of plant uptake experiments where these soil properties have been reported. Radionuclides or other ions which are soluble, and thus present, in soil solution are most readily sorbed by plants. Therefore, soil reactions which affect retention and solubility will directly affect the bioavailability of these radionuclides.
24
Chapter 2
The most likely valence states of U and Th in soil are +VI and +IV, respectively, while that of Ra in soil is +II (Bondietti and Sweeton, 1977). Garten et al. (1981) related patterns of extractability of U and Th from soils to their probable valence states in soil. In the absence of large amounts of soil organic matter, U generally is considered to be mobile and transported as a divalent uranyl (UO2,+) ion (Schulz, 1965). Uranyl carbonate complexing results in a wide range of U solubility. Hostetler and Garrels (1962) reported that U is transported in acid oxidizing solutions as (UO ~+) or UO2(OH) § ions, as a UO2(CO 3)22complex in neutral solutions, and as a UO2(CO 3) 4-3complex in alkaline solution. In contrast, the reduced tetravalent form of U behaves similarly to immobile tetravalent Th (Hansen and Huntington, 1969). Talibudeen (1964) suggested that the Th/U ratio may provide information about weathering intensity in soils, since Th is much less mobile than U. Plant uptake of these radionuclides is somewhat affected by soil pH, depending upon the plant species. These differences may be related to indirect effects of competition from the higher levels of available Ca and Mg which increase with soil pH level. The effects of soil texture on the movement of several radionuclides in soil appear to be similar to those of divalent Ca and Mg. Divalent cations are sorbed by soil clays through exchange mechanisms, so relative rates of sorption generally increase with clay content of soils. Since mobility in soils is considered inversely related to sorption, downward movement of applied radionuclides would be lower in fine textured (high clay) soils. Kiss et al. (1988) reported that soils under native grasslands in Saskatchewan with higher 4~ activities usually had higher 214Bi and 2~ activities. The activities of all three radionuclides decreased as the clay content of surface soils decreased. The latter radionuclides are daughters of 238U and 232Th, respectively. Uranium concentrations in soil were related to soil organic matter content by Talibudeen (1964). Hansen and Huntington (1969) reported distinct Th accumulations in soil horizons immediately below layers with a high organic matter content. Soil organic matter apparently strongly complexed with tetravalent Th, thus increasing the mobility of Th in soil. They also found that Ra was distributed more irregularly than Th, reflecting the effects of the soil chemistry of 238U, 234Th, and 226Ra with time. Radium distributions were explained in terms of U retention and Th mobilization by soil organic matter, Ra uptake by plants, and by time. Hansen and Stout (1968) reported that upper horizons of upland soils had higher concentrations of U than Th, while the concentrations were higher in the B-horizons. Alluvial soils tended to have higher and more evenly distributed concentrations of Th, indicating its greater mobility as organic complexes.
2.4 RADIOACTIVITY IN THE WATER Rivers erode soil which contains radionuclides, and reach lakes and oceans; atmospheric depositions can also occur on their surfaces; and groundwater containing some radionuclides can reach them.
Radioactive Nuclides in Nature
25
In restricted marine and estuarine environments a build-up of clays and organic material can occur which, when buried at depths of up to a few km, can develop into organic shales. Uranium transported into such an environment is reduced to the immobile 4+ state or is adsorbed onto the organics and/or the clays. Some of these shales are quite extensive, an example being the Chattanooga shale of the eastern United States which has an average U concentration of 79 ppm (compared with average shale concentration of 2 ppm (Rogers and Adams, 1969) and once was considered a possible major source of uranium in the United States. Sandstones are the sites for most major economic accumulations of U in the United States and in Australia. The classic situation occurs where U-bearing water flows down through a sandstone aquifer and encounters a redox barrier where U and certain other multivalent elements can be reduced and precipitated, forming coatings on sand grains. As erosion occurs and the land surface is lowered, the redox barrier moves downflow as organic material is oxidized. The U accumulation, now in oxidizing waters, becomes mobilized and moves to the new site of the barrier. At the same time, new U is being freed and transported from the surface sediments. In this way, U ore bodies are formed and grow. Uranium-series daughters which are liberated by recoil and/or chemical differences may be separated from the mass of U and be transported, thus forming a plume of radioactivity, deficient in parent 238U but elevated in one or more daughters. Such plumes are related not only to ore bodies but also can result from quite minor accumulations of radioelements as well. Even before sediments are lithified into rocks, natural processes can promote the accumulation of radioactivity. Where radioelements reside in erosion resistant minerals, sediments incorporating these minerals can be the source of elevated radioactivity readings. For instance, in northeastern Florida there are sands which are enriched in the mineral monazite which contains a significant amount of Th. The enrichment occurs because the high specific gravity of monazite allows its concentration along beaches where lighter materials are swept away. Fossil beaches well inland of the present coast also have such zones of concentration. Similar beach deposits are also known in Australia and Namibia. Therefore, water contains a small and variable quantity of natural radioactivity from the decay of uranium and thorium and their daughters, together with 4~ Background radiation has been increased during recent decades as a result of man's exploitation of nuclear fission. The original--and still the major--artificial input to the hydrosphere is from the fallout of fission products from nuclear weapon testing, and the presence, in rain and fiver water, of 9~ and 137Cs (and other radionuclides) has been well documented (Preston et al., 1967; Cambray et al., 1983). Other sources of radioactivity now include the discharge of small quantities of liquid radioactive waste from the operation of nuclear-powered electricity generating stations and research establishments, the use of radioactive materials in industry and medicine, and also from the use of tracers for the investigation of water and sediment movement. Water serves as the major transport medium of the U-series nuclides. The ocean and its sediments serve as important reservoirs and as a redistribution system for these nuclides. The inputs to the marine system are the river waters and the sediments they
26
Chapter 2
carry. The range of dissolved U concentration of rivers is from 0.29 to 0.005 dpm/1 (0.13 to 0.002 pCi/1 or 0.4 to 0.01 ~tg/1) (Scott, 1982) and the concentration of ocean water is 2.4 dpm/1 (1.08 pCi/1 or 3.25 ~tg/1). Thorium is at exceedingly low levels in river and ocean waters but is transported adsorbed on particles carried by rivers and deposited on the sea floor. The Th generated by the dissolved U in the oceans also quickly finds its way to the bottom by hydrolysing and adsorbing onto sinking particulate matter. Radium which is often adsorbed onto river-borne particulate matter is desorbed when the particulates encounter the high salinity of the ocean. The most important source of Ra to the ocean is ocean floor sediments from which Ra diffuses into the seawater. Polonium and Pb, like Th, are very particle-reactive and quickly adsorb onto particulate material and are deposited on the sea floor. Radon is generated at low levels throughout the water column and also diffuses from the bottom sediments but the level of Rn in seawater is quite low, especially when compared with that found in groundwater. Groundwater is the chief medium facilitating radionuclide separation, and, as such, the various concentrations vary over a wide range of values. For instance, U has been measured at concentrations from 10 -3 ppb to hundreds of ppm, a range of over eight orders of magnitude. In contrast, all Th isotopes are always at very low concentrations in water. In non-saline groundwaters, Ra usually does not migrate very far from the area in which it is produced because of its probability of being adsorbed. Studies in South Carolina showed that Ra activity in crystalline rock aquifers was less than that of coastal plain aquifers, presumably because the crystalline rocks have a higher cation exchange capacity (King, 1982). Despite the somewhat inhibited migration of Ra, all but six states have public water supplies which are known to exceed the maximum contamination limit (5 pCi/1) (Hess et al., 1985). The number of domestic supply wells exceeding the limit is unknown but is likely to be substantial. Radon almost always has the greatest activity of all decay-series nuclides in groundwater, from a few pCi/1 to well over a million pCi/1 from granitic terrains in New England (Hess et al., 1985). The control of the uses of radioactivity, and the limitations of discharges to the environment should ensure that the levels in water are below limits derived from the International Commission on Radiological Protection (ICRP) recommendations. Where appropriate, the radioactive content of water is measured by the operator who is authorized to discharge radioactivity, and the results are checked by the appropriate authorizing Government Departments; in addition, tracer experiments to follow water movement are usually carried out by specialist groups with the appropriate measuring equipment. The measurement of the radioactive content of water is carried out by some government' s Water Authorities as a check on trends and natural levels to be expected in the environment. Of special interest is the concentration of radioactivity in seawater by the living species. This is shown in Table 2.5. The concentration factors for some radionuclides in some species can be as high as hundreds of thousands.
27
Radioactive Nuclides in Nature
Table 2.5 Concentration factors (Bq kg-l biota/Bq L -1 water) in marine biology for long-lived isotopes B iota Element
Cobalt Strontium Technetium Iodine Caesium
Plants Isotopes
6~ 89Sr,9~ 99Tc
1291
134Cs,135Cs,
Animals
plankton a l g a e
Molluscs Cephalo- Crusta- Fish ceans plankton (ex Cepha- pods lopods)
5000 3 5 1000 20
1000 1 100 3000 30
2000 1 100 3000 30
5000 1 1000 10 30
1000 100
1000 100
7000 400
Phyto-
Macro-
Zoo-
200 2
l0
5000 2 1000 10 30
1000 2 30 10 100
1000 100
300 10
137Cs Europium Neptunium
155Eu 237Np
10000 100
Plutonium
239pu,239pu,
100000 1000
1000
3000
50
300
40
200000
2000
20000
100
500
50
24~ Americium
241Am
2000
2.5 R A D I O A C T I V I T Y IN T H E A I R The concentration of atmospheric radionuclides has a special distribution which depends on latitudes and altitudes. Cosmogenic radionuclides have higher production rates in the stratosphere than in the troposphere, because of a higher intensity of cosmic rays in the stratosphere. Fallout nuclides have higher concentrations in mid-latitude of the Northern Hemisphere, because most atmospheric nuclear explosion experiments were made there. Radon and thoron and their decay products are the most important sources of radiation exposure to the general public, contributing on average about half of the total effective dose equivalent received from natural and man-made radioactivity (UNSCEAR, 1988; NCRP-94, 1987; NCRP-97, 1988); see Table 2.6. Radon (222Rn or Rn) is a radioactive noble gas, which is chemically relatively inert. The gas originates from traces of uranium (238U or U) in various materials, e.g. granites. The Rn atom is formed in the mineral grains of the material. A certain fraction of atoms formed escapes into the material pore space and diffuses to the material surface to be released in to the surrounding air. Radon decays radioactively into a series of decay products, or Rn daughters (RnD), which are themselves radioactive. Inhalation and subsequent deposition in the respiratory tract of these RnD constitutes a radiation hazard, due to the irradiation of the respiratory tract tissue by decay of deposited RnD.
28
Chapter 2
Table 2.6 Average individual radiation exposures from various sources
Natural
Man-made
Radiation Sources
Effective Dose equivalent per year/m Sv
Cosmic rays at sea level Radon (222Rn and 22~
0.37
Potassium (4~
Other natural sources
0.30 0.40
Medical use of radiation Nuclear explosives testing Nuclear power production
0.4-1.0 0.01 0.002
1.30
The fact that Rn is omnipresent in all soils, and is measurable down to very low concentrations, makes it an ideal, naturally occurring tracer gas. This property may be used e.g. in exploration techniques, or in the detection of subsoil fissures and aquifers. Radon is the first gaseous isotope in the 238Uchain. Analogously, thoron (22~ and actinon (2~9Rn) occur as the first gaseous isotopes in the 232Th and 235Udecay series. Following the decay of Rn through alpha emission, there are three short-lived isotopes in quick succession, viz. 2~8p0 (historically RaA), 214pb (RaB) and 214Bi (RaC). The latter is followed by 214p0 (RaC'), which has a half-life of 163 s, and the activity of this isotope is taken to be equal to that of its immediate predecessor, 214Bi. It is seen that RaA and RaC' are o~ emitters, while RaB is a 13 radiation emitter (see Table 2.7). All three, but largely RaA and RaC' contribute to the inhalation hazard associated with RnD, since the radiation effects of ~ radiation on respiratory tract tissue is by far more pronounced. The RnD atoms are all metals, and it is known that they are oxidized to the metal oxides e.g. Po---~PoO x rapidly after formation. A Rn atom is formed upon decay of the predecessor 226Ra atom in the mineral grains of the U beating material. When the decay happens close to the grain surface, the Rn atom can, by virtue of its recoil energy, be ejected from the mineral grain into the material pore space. The moisture content of the material has an effect on this process, as a thin moisture layer in the grain surface can retard the recoiling Rn atom so that it ends in the pore space. If no layer exists, the atom may be ejected into an adjacent grain, and not be available in the pore space. If the pores are filled with water, the atom ends in the water layer. It follows that only a certain fraction of all the Rn atoms formed in the material grains can escape into the pore space. This is the emanation fraction of the material. In the pore space the Rn atom diffuses through the tortuous pore paths. This process is governed largely by diffusion, although convective processes may also assist in the transport e.g. if pressure differentials exist across the material. Water content again
29
Radioactive Nuclides in Nature
Table 2.7 The naturally occurring isotopes of radon and their major decay products Isotope
Mode of Decay
Half-life
Energy/MeV
Radon-22 (Rn)
ot
3.823 d
5.4897
Polonium-218 (RaA)
c~
3.05 min
6.0026
Lead-214 (RaB)
13
26.8 min
0.67,0.73 0.35192,0.29522,0.24192 1.54,3.27, 1.51 0.60932,1..7645,1.12028
7 Bismuth-214 (RaC)
[3 Y
19.8
Polonium-214 (RaC)
ot
163
7.6871
22.3 y
0.017,0.061 0.046539
Lead-210 (RaD)
[3 7
Bismuth-210 (RaE)
l] 7
5.01 d
1.161 0.2656,0.3046
Polonium-210 (RaF)
~
138.38 d
5.3044
Lead-206
-
Stable
Radon-220 (Tn)
ot
55.6 s
6.2883
Polonium-216 (ThA)
c~
0.15 s
6.7785
Lead-212 (ThB)
13 Y
10.64 h
0.331,0.569 0.23863,0.30009
Bismuth-212 (ThC)
I] (64%) Y ot (36%)
60.6 m
0.67,0.93,1.55,2.27 0.7272 6.051,6.090
Polonium-212 (ThD)
c~
0.298 ~ts
8.7844
Thallium-208/ThC)
13
3.053 min
1.796, 1.28, 1.52 2.6146, 0.5831, 0.5107
Lead-207
-
Stable
Radon-219 (An)
o~
3.96 s
6.8193, 6.553, 6.425 0.27120, 1.4017
1.780 m
7.386 0.4048, 0.4270
36.1 m
1.38 0.4048, 0.8318,0.4270
2.14 m
6.623,6.279 0.3510
4.77 m
1.43 0.8978
7
Y Polonium-215 (Ac)
ot Y
Lead-211 (AcB)
13 Y
Bismuth-211 (AcC)
o~ Y
Thallium-207 (AcD)
[3 Y
30
Chapter 2
affects the transport to a very large degree, since the diffusion of Rn in water is about 100 times slower than in air. The diffusive transport of the Rn through the material is characterized by the diffusion length. This quantity, deduced from the diffusion coefficient, represents a typical distance that a Rn atom will move in the material before it decays. For typical soils (moist) the diffusion length is 20-50 cm, while for coarse, dry sands, the value may be as high as 1.5 m. In dense material, such as dense granite, the diffusion length is only of the order of 10-15 cm. In air, the diffusion length is 2.18 m. The diffusion coefficient, which determines the diffusion length, is an effective diffusion coefficient for the material, and incorporates all mechanisms which contribute to the diffusion rate, e.g. adsorption, and moisture content. If the Rn atom reaches the material surface before it decays, it escapes from the material into the surrounding air. This process is termed exhalation, and is characterized by the exhalation rate, which is the flow of Rn atoms from the surface of the material (atoms/m2/s). The materials from which the Rn exhales, can be any U-containing material. Ordinary building materials, such as cement, originate in granite quarries, where the granite may contain elevated quantities of natural U. These materials will then emit Rn into the buildings containing them. The large concrete floor slab is a typical example. Building material additives, such as fly ash from coal burning power stations can contain significant U levels from the original coal. The building material into which the fly ash is blended will then emit Rn. Other examples include tailings from mining operations. When large in extent, these tailings impoundments can represent significant sources of atmospheric radon, leading to potential radiological exposure of members of the public. In underground mines, the walls of the workings e.g. tunnels and slopes, can exhale significant amounts of Rn gas. Due to poor ventilation, the Rn levels can grow to high concentrations and elevated radiation doses to mine workers (up to 50 or more mSv/a) may result. It has been shown that the major source of Rn in the indoor environments is the subsoil region. From here, Rn can diffuse (or be convected) into the building through cracks, water pipe entrances etc. Another source of radon gas occurs when Rn gas is dissolved in water, and the water is disturbed, e.g., in a shower. Upon dispersal into droplets, the radon dissolved in the water is released into the air. Underground water, where high pressures exist, may contain significant amounts of Rn (up to 370 kBq m-3). Such borehole water may therefore be significant sources of Rn in houses and other buildings. The radon daughters (RnD) are metal atoms, Po, Bi and Pb, and these atoms are oxidized to the metal oxides very rapidly after formation, e.g. Po---~PoO2. The shortlived RnD have relatively high decay energies. Since these isotopes are formed in succession in a decay series, it follows that for a given mother radon concentration, there will be a certain ratio of the various decay products with regard to each other. This ratio will depend on the age of the mixture of Rn and RnD, since the decay process is time dependent. If a fixed number of Rn atoms are sealed in a closed container at time
Radioactive Nuclides in Nature
31
t = 0, the various RnD will grow in time. The 2~8po grows into the level of the Rn first, followed by 214pb and 2~4Bi. The decay products reach equilibrium with the mother Rn after about four hours, and then decay at the same rate as the radon. The equilibrium state of the Rn and RnD at any stage is characterized by a so-called equilibrium or F factor. An F factor of one represents a full equilibrium between Rn and RnD, while values less than one represent realistic everyday situations. The F factor in outside air is typically 0.7-0.8. In indoor environments it varies between values of 0.33 and 0.45. In underground mining situations, where the ventilation rate of the working is often very variable, the F factor varies between 0.1 and 0.7. In the former case the air is said to be young, and in the latter case the air, or Rn, is said to be old. There can be large short-term localized variations around "average" Rn concentration directly related to atmospheric stability conditions. Strong inversion layers close to the ground can act to trap radon and daughters and result in increases in gamma dose rate up to 25% or more. Similar temporary increases can occur in heavy rainstorms when the radon daughters are washed out onto the ground surface. 222Rnconcentrations in outdoor air are largely dependent on the 226Raconcentration in the soil, the exhalation rate from the soil and atmospheric dispersion factors. Values over large areas of water will be lower than over land areas: and values near surface uranium deposits up to several orders of magnitude higher. Diurnal and seasonal variations are of the order of 2.4 times the lowest values. Indoor concentrations are more variable, in particular in northern climates where low ventilation rates are encountered inside sealed and insulated buildings. As a result the majority of exposure arises indoors, and may be many times the outdoor contribution.
REFERENCES Bondietti, E.A. and Sweeton, F.H., Transuranic speciation in the environment, pp. 449-476. In: M.G. White and P.B. Dunaway (ed.) Transuranics in Natural Environments NVO-178, National Technical Information Service, Springfield, VA, 1977. Cambray, R.S., Lewis, G.N.J., Playford, K., and Eakins, J.D., Radioactive Fall-out in Air and Rain: Results to the end of 1982. AERE R10859, HMSO, 1983. Garten, C.T. Jr., Bondietti, E.A. and Walker, R.L., Comparative uptake of uranium, thorium, and plutonium by biota inhabiting a contaminated Tennessee floodplain, J. Environ. Qual. 10 (1981) 207-210. Gascoyne, M., Geochemistry of the actinides and their daughters. In: M. Ivanovich and R.S. Harmon (ed.) Uranium Series Disequilibrium: Applications to Environmental Problems. Oxford University Press, Oxford, 1982. Ginzburg, V.L., and Syrovatskii, S.I., The Origin of Cosmic Rays, Pergamon Press, New York, 1964. Hayakawa, S., Prog. Theor. Phys. 13 (1955) 464. Hansen, R.O. and Huntington, G.L., Thorium movements in morainal soils of the high Sierra, California. Soil Sci. 108 (1969) 257-265. Hansen, R.O. and Stout, P.R., Isotopic distribution of uranium in soils. Soil Sci. 105 (1968) 44-50. Hess, C.T., Michel, J., Horton, T.R., Prichard, H.M., Conoligio, W.A., The occurrence of radioactivity in public water supplies in the United States. Health Phys. 48 (1985) 553-586.
32
Chapter 2
Hostetler, P.B. and Garrels, R.M., Transportation and precipitation of uranium and vanadium at low temperatures with special reference to sandstone-type uranium deposits. Econ. Geol. 57 (1962) 137-167. King, P.T., Michel, J., Moore, W.S., Groundwater geochemistry of Ra-228, Ra-226 and Rn-222. Geochim. Cosmochim. Acta 46 (1982) 1173-1182. Kiss, J.J., Dejong, E. and Bettany, J.R., The distribution of natural radionuclides in native soils of southern Saskatchewan, Canada. J. Environ. Qual. 17 (1988) 437-444. NCRP Report No 94 (1987): Exposure of the Population of the United States and Canada from Natural Background Radiation. NCRP Report No 97 (1988): Measurement of Radon and Radon Daughters in Air. Preston, A., Jeffries, D.F., and Dutton, J.W.R., Water Research 1 (1967) 475-496. Robertson, D.S., Basel Proterozoic units as fossil time markers and their use in uranium prospection in formation of uranium ore deposits, pp. 495-512. Proceedings of a Symposium, Athens, Intl. Atom. Energy Agency, Vienna, 1974. Rogers, J.J.W. and Adams, J.A.S., Uranium, p. 92-A-1-92-G-7. In: K.H. Wedepohl (ed.) Handbook of Geochemistry. Springer-Verlag, New York, 1969. Schulz, R.K., Soil chemistry of radionuclides. Health Phys. 11 (1965) 1317-1324. Scott, M.R., The chemistry of U- and Th-series nuclides in rivers, pp. 181-201. In: M. Ivanovich and R.S. Harmon (ed.) Uranium series disequilibrium: applications to environmental problems. Oxford Univ. Press, Oxford, 1982. Talibudeen, O., Natural radioactivity in soils. Soils and Fertilizers 27 (1964) 347-359. United Nations Scientific Committee on the Effects of Atomic Radiations (UNSCEAR): Sources effects and Risks of Ionizing Radiation. United Nations, New York, 1988.
33
CHAPTER 3
Technologically Modified Exposure to Natural Radiation
There are circumstances where man finds himself in a natural radiation environment to which he would not be exposed if some kind of technology had not been developed. Examples are travelling by air, using natural gas for cooking or heating purposes, living in the neighbourhood of a coal-fired power plant. The resulting exposures have been labelled "technologically enhanced" natural radiation exposures, defined as exposures to truly natural sources of radiation (i.e., naturally occurring radionuclides and cosmic radiation) which would not occur without (or which are increased by) some technological activity not explicitly designed to produce radiation. In some cases, technology helps to reduce the natural radiation exposure. For example, when drinking water supplies are drawn from surface waters, the use of water-purification processes brings about a decrease in the concentration of radium and other naturally occurring radioactive elements. Another example is the burning of fossil fuel, which reduces the specific activity of 14C in the biosphere and therefore lowers the doses from those radionuclides. Operations and activities which act to concentrate and redistribute naturally occurring radioactive material (NORM) in the environment are numerous and further sources continue to be identified. The following sources have been identified as major contributors: 1. Mining and minerals processing facilities: Uranium mining and milling Monazite/beach sands operations Copper mines Phosphate rock mining/production of phosphoric acid Gold mines Tin mines Mines extracting Pyrochlore and Euxenite ores (Columbium/Tantalum) Aluminum mines Beryllium mines
Chapter 3
34
Iron mines Lead mines Molybdenum mines Nickel mines Silver mines Titanium mines Zinc mines Zirconium mines Coal mines Fluorspar mines Granite mines Limestone mines 2. Industrial processes: Foundries using zircon sands Sandblasting with zircon sands Operations producing building materials from mine wastes e.g. phosphogypsum and phosphate slag Fertilizer producers utilizing phosphoric acid Operations using fly ash from coal mines Scrap-yards utilizing contaminated scrap from mines Foundries and smelters using contaminated scrap Operations using lignite, pumice, scoria and mineral wool Titanium dioxide production from ilmenite Tin smelting Operations using pyrochlore in the production of special alloys The production of zirconia from baddeleyite The manufacture of glazes from zirconia The manufacture of catalysts and special glasses from rare earths Numerous localized areas of the world have been subjected to mining and mineralprocessing activities involving NORM: in some of these areas significant contamination of the surrounding environment with enhanced levels of NORM has occurred resulting in enhanced radiation exposure of the public.
3.1 R A D I A T I O N F R O M C O A L F I R E D P O W E R P L A N T S
Coal, like most materials found in nature, contains trace quantities of the naturally occurring primordial radionuclides. Therefore, the combustion of coal results in the release to the environment of some natural activity and in the re-distribution of that natural activity from deep in the earth to locations where it can modify ambient radiation fields and population radiation exposure.
Technologically Modified Exposure to Natural Radiation
35
During recent years, radioactivity released into the environment by coal power stations has received a great deal of attention from the public, as well as from governments and their agencies. Some of the reports have received much attention and have been published in journals and newspapers. One such report which has been much discussed by the general public is derived from the study conducted by the U.S. Environmental Protection Agency (EPA). The report, "Radiological Impact Caused by Emission of Radionuclides into Air in the United States", was prepared as a result of legislation that requires radionuclides to be included among other sources of air pollution. Preliminary findings suggest that there are greater risks to the public of developing cancer from radionuclides emitted by coal-fired power plants than by normally operating nuclear plants. Radionuclides~including isotopes of uranium, thorium, tritium, argon, noble gases, iodine, radon, and polonium~are released into the atmosphere from operating the various facilities. These radionuclides are dispersed into populated areas where exposure occurs by breathing or swallowing the materials. When coal is burned, the mineral content is converted to ash and slag. These wastes contain most of the radionuclides originally present, but a fraction of the ash is released into the atmosphere. The quantity released depends on the particulate control system, furnace design, mineral content of the coal, and existing emission control standards. According to the EPA figures for existing coal-fired plants, the lifetime risk to individuals nearest to the facilities ranges from 60x 10-6 to 700x 10-6. For boiling-water and pressurized-water reactors, the risks are 20x 10-6, respectively. Translated into health effects, the EPA estimates that from 0.0004 to 1.5 fatal cancers can be expected to develop per year of operation for each existing coal-fired power station. There are 250 such stations in the U.S. and the wide variation in range is due to differences in siting and power generating capacity. The accepted number of fatal cancers for each of the 69 nuclear power generating plants is 0.001/year of operation. The EPA report makes reference to a total of 250 existing and 145 new coal-fired plants, 25 boiling-water reactors (BWR), and 44 pressurized-water reactors (PWR) in the U.S. On a direct comparison at suburban sites between coal and nuclear plants, BWR facilities each can be expected to produce 0.0013 fatal cancers per year and PWR facilities, 0.0009 fatal cancers per year. Existing coal-fired plants, on the other hand, each can be expected to produce 0.10 fatal cancers per year and new coal plants, 0.017 fatal cancers per year. The principal radionuclides emitted by coal-fired stations include radon, uranium, and thorium. Those emitted by nuclear plants include the noble gases, tritium, and the halogens. Activity concentrations due to the presence of natural radionuclides in coal are shown in Table 3.1 (United Nations, 1982). This table presents results of measurements of radionuclides in coal samples originating from mines or from power plants. The most significant study is that of Beck et al. (1980) who listed the concentrations measured in almost 1000 samples obtained directly from mines providing most of the
36
[..
r~
0
M
0
0 0
0
o
g.
N
r~
~
0
.<
!~
0
0
~
~1~
0
0
r,.)
i~ .
0
;~ 0
r,.)
i.~
0
r,..) ~
r
eq V
0
V
0
0
0 ~
r
r~ V
"-
--"
0
~
0
.
,~
e~
~r~
~
0
I
I
0 r~
0
r ~'~
E
~
0
9,,,,,,
E
r,r
e~
~0
0 ~'~
0
I"~
~
I
~"
0 i,~ I i" ~ e~
o
0 ,,-~
0o
o
9
0 9
,-~
I
e',l
~.
I
7
c~
Chapter 3
~
N
;~
I'~
.1
00
0 r ,,~
E o
0 r ~
~
0
,-
06
r
m
r
~"
r,-
0
~
r
q.~
~
r
,-~
0")
,--
r
,"
eq
0
or
T e c h n o l o g i c a l l y M o d i f i e d E x p o s u r e to N a t u r a l R a d i a t i o n
o~ c~ c~
c~
c~
c~
c~
c~ c~
o~ c~
--
V
0
0
37
38
Chapter 3
coal presently used in the United States. These authors found that the activity concentrations measured in the coal samples varied over two orders of magnitude (0.7-70 Bq kg -1 for 4~ less than 3-520 Bq kg -~ for 238U, and 3-320 Bq kg -1 for 232Th). Variations can be quite large even in the same area. Gluskoter et al. (1977) obtained concentrations ranging from 4-300 Bq kg -j for 238U,0.4-10 Bq kg -~ for 232Thand 3-100 Bq kg -1 for 4~ in five different seams of coal mined in Illinois. In general, the concentrations of natural radionuclides in coal are less than those in the earth's crust. Occasionally, however, and usually as a result of leaching from abnormally radioactive overburdens of volcanic origin, very high concentrations of some radionuclides, in particular uranium, can be found in various coal deposits. Those uraniferous coals are the exception and occur almost invariably, in low-grade coal deposits. It is assumed that the average activity concentrations in coal are 50 Bq kg -~ of 4~ and 20 Bq kg -~ each of 238U and 232Th and that all the decay products of 238Uand of Z32Th are in radioactive equilibrium with their precursors, although that might not be always the case for 2~~ and 21~ (see, for example, the results of Kakinen et al. (1975) (Table 3.1). Enhanced activity concentrations of 2~~ could conceivably occur if large quantities of 222Rn diffuse from adjacent high activity rocks into a lower activity coal seam, with subsequent trapping of the decay products in the coal. In the production of electric power, coal is burned in furnaces operating at temperatures of up to 1700~ Most of the mineral matter in the coal is fused into a vitrified ash. A portion of the heavier ash, along with incompletely burned organic matter, drops to the bottom of the furnace as bottom ash or slag. The fly-ash, however, is carried through the boiler along with the hot flue gases and any volatilized mineral compounds to the stack were, depending on the efficiency of emission control devices, some fraction is collected while the rest (escaping fly ash) is released into the atmosphere. The radionuclides included in the non-combustible mineral matter are thus partitioned between the bottom ash and fly ash, except for the gases and volatilized minerals which will be incorporated directly into the flue gases. Table 3.2 presents a list of reported activity concentrations of natural radionuclides in bottom ash, collected fly ash and escaping fly ash. Owing mainly to the elimination of the organic component of the coal, there is very approximately an order of magnitude enhancement of the activity concentrations from coal to ash. Consequently, the natural radionuclide concentrations in ashes and slags from coal-fired power stations are significantly higher than the corresponding concentrations in the earth's crust. The arithmetic averages of the concentrations in escaping fly-ash from Table 3.2 are, in Bq kg -~, 265 for 4~ 200 for 238U, 240 for 226Ra,930 for 21~ 1700 for 21~ 70 for 232Th, 110 for 228Thand 130 for 228Ra. The amount of emission of radioactive isotopes into the atmosphere through the power plant chimney depends mainly on the efficiency of electrostatic precipitators or other devices used for the cleaning of flue gases. There are two types of coal burning power plants in operation in the world: 1. power plants with emission into the atmosphere of about 10% of the produced fly ash, and
Technologically Modified Exposure to Natural Radiation
39
2. Modern power plants with emission of only about 1% of the produced fly ash. The arithmetic mean for potassium, uranium and thorium activity concentration in fly ash carried by flue gases through electrostatic filters into the chimney is: 4~ 500 Bq kg -1 238U" 200 Bq kg -I 232Th: 200 Bq kg -I Based on these values one can obtain the estimate of atmospheric release. On average it is: 238U" 1500 MBq GW -1 year -~ 232Th"1500 MBq GW -1 year -~. The amounts of radioactive substance emitted in the airborne effluents of coal-fired plants have been studied by a number of research groups, including Eisenbud and Petrow (1964), Terrill et al., Martin et al. (1971), McBride et al. (1977) and Styron et al. (1980). A detailed report on radiological impact of airborne effluents of coal-fired and nuclear power plants has been published by McBride et al. (1977). Data based on the same report are also presented by Torrey (1978). Here is the summary of their reports. The radiological impact of naturally occurring radionuclides in airborne effluents of a model coal-fired steam plant of 1000 MW(e) is evaluated assuming a release unto the atmosphere of 1% of the ash in the coal burned, and compared with the impact of radioactive materials in the airborne effluents of model light-water reactors of 1000 MW(e). The principal exposure pathway for radioactive materials released from both types of plants is ingestion of contaminated foodstuffs. For nuclear plants, immersion in the airborne effluents is also a significant factor in the dose commitment. Assuming that the coal burned contains 1 ppm uranium and 2 ppm thorium, together with their decay products and using the same impact analysis methods used in evaluating nuclear facilities, the maximum individual dose commitments from the coal plant for the whole body and most organs (except the thyroid) are shown to be greater than those from a pressurized-water reactor (PWR) and, with the exception of the bone and kidney doses, less than those from a boiling-water reactor (BWR). With the exception of the bone dose, the maximum individual dose commitments from the coal plant are less than the numerical design guideline limits listed in 10 CFR 50, Appendix 1, for light-water reactors (LWRs). Population dose commitments from the coal plant are higher than those from either nuclear plant, except for the thyroid dose from the boiling-water reactor. The use of coal containing higher uranium concentrations and/or higher particulate releases (>>1%), characteristic of the present coal-fired power industry, could result in dose commitments from a coal plant several orders of magnitude higher than those estimated in this study. The study is limited to a comparison of the radiological impacts of airborne effluents from model coal-fired and nuclear power plants and does not compare the total radiological impacts of a coal vs. a nuclear economy. It is concluded that an evaluation of the radiological impact on the environment should be included in the assessment of both coal-fired and nuclear power plants.
40
T
~
.,.~
o
o
0
0
t.N
~
~ ~
t'~
0
c~
0
t~
I~
~
I~
I" ~
.~
ZZZ
.,.~
!~
1" ~
.~
0
~ ~ z - ~
~
,-.-
N
,--~
e~
E oE. - ~~ o
I
T
0
0 0
T
~
0
Chapter 3
0
c~
oo
E
~J"
('~
I
"-~
~0
,-~
q'~
,-~
,..
~ "-
~
~.~
~:~
-~
~"
~'~
"
,..~
~
~
",~"
r',-
~
~ ~
c',l
~
r~
,--
~
('q
~,~~<
~'~
~ . ~ ~ ~
~O
~l"
0
~
~'xl
Technologically Modified Exposure to Natural Radiation
0 (-~
m
(.q
I
~'.
.~<<<<
~
~
~
~.~
o
~
0 (~
~
~
~
("a
0 I 0
.~
r"-
0
I
cr
I
c'q
""
~
o ,,N
c',l
o,..~
o
~3 .o "~
E
41
42
Chapter 3
According to Styron et al., (1979) for a realistic assessment of the magnitude of release of radionuclides, special attention needs to be given to lead-210 and polonium210 since they appear to have a large potential for significant environmental impact and have not received sufficient attention in trace-element studies for power plants. Another potentially important parameter in determining radiation exposure to man centres on disposal and utilization of coal ash and refuse. Lee et al., (1977) have suggested that emanation of radon-222 from ash disposal ponds will be the most serious radionuclide problem associated with increased use of coal. A potential hazard can be associated with the use of fly ash in cement and concrete blocks and in roadway construction. The radium-226 in these concrete blocks used for home construction may constitute an important source of radon-222 dose to the public. Additional readings on the subject of radiological impact of coal burning power plants can be found in references Aigneperse et al. (1982), Blackburn and Gneran (1979), Halbritter et al. (1982); Watanabe et al. (1980), Valkovic (1983).
3.2 R A D I A T I O N E X P O S U R E DUE TO E X P L O R A T I O N AND USE OF PHOSPHATES
Phosphate rock deposits contain uranium (U), radium (Ra), thorium (Th), and other radionuclides as contaminants. Uranium in phosphate rock deposits throughout the world range from 3 to 400 mg kg -~ (Guimond, 1978). It has been estimated that 1000 kg of Florida phosphate rock contains about 100 ~tCi each of 238U and 226Raand 4 ~tCi of 23~ (Menzel, 1968). Some of these elements are retained in the H3PO 4 and the remainder are transferred to the by-products during fertilizer manufacture. For instance it is estimated that 60% of the radioactivity in mined Florida phosphate rock remains with slime and sand tailings during beneficiation (Guimond and Windham, 1975). Literature on the radionuclide content of phosphates is rather extensive. Radium, uranium, thorium and members of their decay series are the principal radioelements present in fertilizers. Radium content in fertilizers is extensively discussed by some authors (Guimond, 1990; Roessler, 1990). Other reports deal with radon release due to the use of phosphate fertilizers (Paul et al., 1984; Roessler et al., 1990). However, the majority of authors are concerned with uranium and thorium concentrations. Sedimentary phosphate ores, such as those found in Florida and Morocco, tend to have high concentrations of uranium, whereas the opposite occurs with magmatic ores, such as apatite from Kola. Typical activity concentrations of 238U are 1500 Bq kg -1 in sedimentary phosphate deposits and 70 Bq kg -1 in apatite. 238U is generally found in radioactive equilibrium with its decay products. The activity concentrations of 232Th and of 4~ in sedimentary phosphate rock are much lower than those of 238U, and comparable to those observed normally in soil. Several authors have reported on uranium concentrations in phosphates (Cathcart, 1978; Gorecka and Gorecki, 1984; Paschoa et al., 1984; Saleh and A1-Saleh, 1987; Vucic and Ilic, 1989). Radioactivity in
Technologically Modified Exposure to Natural Radiation
43
phosphate rocks due to 238U, 226Ra,232Thand 4~ is discussed by some authors (Komura et al., 1985; Landhe and Rao, 1988); others discuss U and Th activity as a function of particle size (Metzger, 1979; Metzger et al., 1980). Radioactivity released by the use of phosphate fertilizers, accumulation in soil, migration and transfer or radioactivity into plants is extensively discussed (Arkhipov et al., 1981; Aswathanarayana, 1988; Guimond and Hardin, 1989; McNabb et al., 1979; Osmond et al., 1984; Rothbaum et al., 1979; Shishkunova et al., 1989; Van Clecmput et. al., 1993). Mining and processing phosphate ores redistribute 238U and its decay products among the various products, by-products and wastes of the phosphate industry. Effluent discharges into the environment as well as the use of phosphate fertilizers in agriculture and of by-products in the building industry are possible sources of exposure to the public (Lardinoye et al., 1982; Nash, 1987). The exposure pattern, or chain of relationships between radioactivity in agricultural products and radiation dose and health effects (risk) in humans, is schematically presented in Fig. 3.1. The actual radionuclides present in some phosphate fertilizers (including by products and wastes) contribute to the radiation exposure of a population (Fisher et al., 1979; Fitzgerald and Sensitaffar, 1978; Phillip et al., 1978). Table 3.3 summarizes the radiation exposure doses due to the industrial exploitation of phosphate rock, expressed in terms of collective effective dose equivalent commitments resulting from the decision to use a unit mass of marketable ore to accomplish a defined purpose, as reported by UN Scientific Committee on Effects of Atomic Radiation (United Nations, 1982). The impact from one-year phosphate ore processing has been very crudely estimated on the basis of the following data and assumptions:
Table 3.3 Collective effective dose equivalent commitments per unit of marketable phosphate ore (10-6 Source of exposure
man
Sv t-j)
Cloud Passage Inhalation
Activity deposed Internal irradiation
External irradiation
Atmospheric discharges from an ore drying plant
0.04
0.01
0.002
0.05
Atmospheric discharges from wet process phosphoric acid plant
0.03-0.2
0.0009-0.06
0.0001-0.010
0.03-0.3
Atmospheric discharges from a uranium recovery plant
0.05
0.007
-
0.06
17000
16000
33000
Agricultural use of phosphate fertilizers
Total
From: Radiological impact of uranium recovery in the phosphate industry, Nuclear Safety, Vol. 22, No. 1 (1981).
Chapter 3
44
A. Radioactivity in "additives" Rate of application to soil Management (tilling, etc.) - Movement in soil (weathering, leaching)
-
-
B. Radioactivity in Soil
- Soil-Plant transfer - Translocation
C. Radioactivity in Plants l
- Role in animal diet - Transfer to meat, milk, eggs
D. Radioactivity in Animal Products l
J
- Role in human diet - Summation over diet components At"
E. Dietary Intake of Radioactivity - Ingestion dose conversion factors - Summation over radionuclides
F. Radiation Dose (committed dose equivalent, committed effective dose) -
Risk coefficient
G. Health Effects (Risk) Fig. 3.1. Chain of relationships, radioactivity in agricultural products to dose and risk in humans.
Technologically Modified Exposure to Natural Radiation
45
1. The world production of marketable rock is 1.3x 1011 kg in a typical year; 2. All the marketable ore is processed in ore drying plants; 3. Seventy per cent of the marketable ore production is used to prepare phosphate fertilizers; 4. Ten per cent of the by-product gypsum is used as building material in dwellings. On the basis of these assumptions, the collective effective dose equivalent commitment resulting from the 1977 production of phosphate rock is estimated to be about 3x105 man Sv. By far the most important contribution to the total dose is due to the use of by-product gypsum in dwellings. The total of the other contribution to the collective effective dose equivalent commitment is only about 6000 man Sv. Many reports deal with radiation dose estimates from phosphate industry operations (Booth, 1977; Guimond and Windham, 1975; Lindeken and Coles, 1978; Moere, 1977; Patridge et al., 1978; Pfister et al., 1976; Pfister and Pauly, 1980). Phosphate rocks vary considerably in their content of U, Ra, and Th, depending on the geographical area from which they were mined. In a survey of phosphate rock samples from all of the major phosphate rock-producing regions of the world at the time, the median contents were 59 mg kg -~ or U, 8 mg kg -1 of Ra (Menzel, 1968). It was estimated that totals of 400 ~Ci each of U and Ra and 15 ~Ci of Th had been applied per acre in some potato (Solanum tuberosum L.) fields in Maine over a 45-year period. Such additions of U and Ra are nearly equal to the total amounts naturally occurring in the soils. However, the addition of Th was much greater than the naturally occurring amount. Most crops are not as heavily fertilized with P as potatoes, although tobacco (Nicotiana tabacum L.) is usually highly fertilized with P. About 33 % of the U in beneficiated phosphate rock concentrate feed for acidulation by the wet-process method was found in the phosphogypsum (PG) byproduct. The remainder of the U was found mainly in the H2PO4, which subsequently is processed to several types of P fertilizers. Eighty-one samples of NPK fertilizers were analyzed for 238U,235U,226Ra, 228Ra,and 228Thin Finland (Mustonen, 1985). These fertilizers represented 28 grades of fertilizers used for agriculture, forests, and gardens, with a mean N - P - K grade of 16.5-6.2-11.3. The mean activities of radionuclides per unit weight of P varied somewhat among factories. At an annual rate of 30 kg P ha -~, the radionuclide application rates then were estimated. It was calculated that the annual contribution of 238Uin P fertilizer was about 0.25% of the total U in the surface 10-cm layer. Because the applied P eventually is mixed with soil at greater depth (probably a 25-cm layer), the annual U contribution would be less than the above percentage (Mustonen, 1985). It was estimated that P fertilization would cause an annual increase of about 0.04% of the total Ra concentration in tilled soils of Sweden (Mortvedt and Sikora, 1992). Few papers have been published on the effects of radioactivity in P fertilizers on agricultural crops. Winter wheat grain and straw were analyzed for 226Ra after plots were fertilized annually with P fertilizers for 11 years in Belgium (Kirchmann et al., 1980). Total P application rates for this 11-year period were 153 and 597 kg ha -~ which could provide mean annual rates of 14 and 54 kg P ha -~, respectively. They reported no
46
Chapter3
statistical differences in 226Raconcentrations of either grain or straw from the two P application rates. A zero-P rate was not included in this study for comparison purposes, but non-fertilized soil and fertilized soil were analyzed for 226Ra. It was estimated that the 226Racontamination in P fertilizers added to soil at the two P rates represented only 0.25 and 1.08%, respectively, of the total 226Rain the upper 20-cm layer of soil. They concluded that the application of P fertilizers, even at that high rate, did not significantly affect the 226Racontent of wheat grown on this soil (Kirchmann et al., 1980). Plant tissues and soil samples from nine long-term (>50 years) soil fertility plots in the United States were analyzed for 226Raby a Rn bubbler tube method and for U and Th by inductively coupled plasma (ICP) spectrophotometry. The triple superphosphate (TSP) used for these studies, made from Florida phosphate rock, had been applied at rates of about 30 kg P ha -1 annually. Results showed that there were no differences in U, Ra, or Th concentrations in corn (Zea mays L.) leaves or grain, soybean (Glycine max. L. Merr.) leaves or grain, or timothy (Phleum pratense L.) forage grown on nonfertilized or TSP-fertilized soil (Mortvedt and Sikora, 1992). In another study, there were also no differences in concentrations of corn, wheat, or soybean grain grown under field conditions, but those of immature corn forage increased with the high PG application rate. The PG contained 25 pCi g-~ of radioactivity. Analyses of soil after soybean harvest showed that the level of radioactivity in the surface 15 cm layer increased from 0.9 to 1.09 pCi g-1 with PG application but that of the subsurface layer was unaffected. In evaluating the extent to which radioactivity in PG would contribute to human radiation exposure, it was concluded that application of PG to soils at a rate of 1100 kg ha -~ once every four years would not significantly affect radioactivity in grain crops. Application rates of PG to supply recommended rates of Ca for peanuts (Arachis hypogea L.) or S for other crops range from 500 to 1000 kg ha -~ (Roessler, 1990). Uptake of these radionuclides by terrestrial plants is relatively low but varies considerably among elements. Plant uptake of U generally is greater than that of Th or Po. While plant uptake mechanisms have not been well studied, it has been shown that their uptake is competitively depressed by Ca 2+. A brief review of Ca physiology in plants may give some understanding of the intake of these radionuclides and their decay products, especially Ra 2+ (Mortvedt and Sikora, 1992; Shishkunova et al., 1989). Analysis of soils from a long-term pasture experiment in New Zealand showed that the U from superphosphate remained in the top 5 cm of soil which was high in organic matter. Phosphate deposits are also considered as uranium ore. Processes and economics of uranium recovery and its impact on phosphoric acid plants have been also studied. Many countries had programs on uranium recovery from phosphoric acid; as a consequence, recovery of uranium has been discussed by many authors (Baran and Kuca, 1988; Can, 1984; Celon and Lazarevic, 1971; De Martean and Sorale, 1983; Dennis, 1984; Efimova et al., 1987; Habashi et al., 1986; Hurst, 1983; Inone and Nakashio, 1982; MacCready et al., 1981; Petrache et al., 1987; Ring, 1975). In the work reported by Hayumbu et al. (1995) ten phosphate rock samples from Nepal, Pakistan, Tanzania, Burkina Faso, Colombia, Tunisia (Gafsa), Peru (Bayovar),
=~
i.
9 .,..
E
. ,-~
I:I.
~J
~J
E
--
~
u~
~:)
--
+I
~O
('N
~O
~ +I
0',
+I
L'N
+I
"
+I
+I
+I t'--
oC
+I ~
,--.
+I
~
,-+I
~
+I ~o
"m"
+I r---
t'~
"m"
~.~"
~ +I
~'i ~.~
+I
t~
+I
+I c,l
+I
+I 0
~
~ +I
'-~
,--. +I ~ +I
L'q
I"+I
+I
~O
+I
+1
+I Ch
+I
e~
u~ oO +I "m" l"-u'b
em 0"~
" +I
+I
+I
Technologically Modified Exposure to Natural Radiation
_~
+I ~o
9
oO
+I Cq
+I
+I u~
~
+I
i----
+I L'---
om
+' ?, @
"~
.
+I
+, +,
+I
r'-+I
+I
+I r---
~-
,--
~O
--
+I
O0
Um
+I Om
I"-I
U'~
L'N
+I
+'
+I
~
+I
+I
oc
Lr~
(N
Ii,
~
+I
+I
~
U'b +I
+, ~
+I
O0
47
48
Chapter 3
Zambia, Thailand and USA (Florida) were analyzed by using the X-ray fluorescence method for elemental composition. These rock phosphates have also been evaluated for their agronomic effectiveness by means of radioisotope techniques (Zapata and Axmann, 1991). The samples were prepared as pellets and analyzed using the emission transmission method for trace elements (Markowicz and Haselberger, 1992). All samples were excited using an Mo tube with an Mo secondary target. For quality control of concentration data analysis of standard reference material CRM 032 (phosphate rock) produced by CEC, Brussels, has been used. Satisfactory agreement between certified and measured values was obtained. The concentrations of elements determined in different phosphate fertilizers are shown in Table 3.4. Concentrations of the following elements were determined, Ti, Mn, Fe, Co, Ni, Cu, Zn, Br, Se, Rb, Sr, Y, Pb, Th and U. With the exception of uranium for which all concentrations above the detection limit are reported, only values with errors less than 25% are reported in Table 3.4. The errors on the concentrations shown in Table 3.4 were taken at 2 cj. The only elements which appear in all samples are Zn, Fe, Sr and Y. Uranium appears in concentrations above the detection limit in all but one sample (Nepal). The range and mean values (ppm) of element concentration in the studied phosphate rocks are shown in Table 3.5.
Table 3.5 The range and mean value of element concentrations in phosphate rocks Element
Range/ppm
mean/ppm
Ti
MDL-1459
558
Mn
MDL-10030
1370
Fe
1715-25000
10450
Co
MDL-127
16
Ni
MDL-106
22
Cu
MDL-396
55
Zn
17.5-2577
385
Br
MDL-10.6
2.8
Se
MDL-8.5
2.6
Rb
MDL-26
8.2
Sr
144-4405
1473
Y
23-156
82
Ce
MDL-2476
248
Pb
MDL-120
33
Th
MDL-21
6
U
MDL-103
36.8
Technologically Modified Exposure to Natural Radiation
49
In the report by Santos et al. (1995) 21~ and 2mpb concentrations in urine, hair and skin smear samples from individuals using phosphated fertilizers have been compared with a control group of occupationally unexposed individuals. Urine and hair samples of the test group showed slightly higher concentrations of 2mpo and 21~ than those observed for the control group. These concentrations remained, however, lower than those for uranium mine workers. Skin smear values indicated contamination by direct contact with dust from fertilizers and this may contribute to skin cancer induction in this risk population. Martinez-Aguirre et al. (1994) have performed an extensive study of the presence of natural radioactivity around a phosphate fertilizer factory complex situated in an estuarine area of southwest Spain. The study has concluded that the wastes from such industries are the cause of the enhancement of natural radioactivity in the immediate environment. Thus, significantly high levels of U and Th-isotopes and 226Ra are detected in water and sediment samples collected in this area. These conclusions, based on the enhanced isotopic concentrations, are further supported by the measured U, Th and Ra isotopic activity ratios being quite different from any observed elsewhere in undisturbed estuaries. These isotope activity ratios appear to be very sensitive indicators of waste disposal practices in such environments.
3.3 ENHANCED EXPOSURE TO COSMIC RAYS Between sea level and the upper limit of the earth's atmosphere there is about 1000 g cm -2 of air mass. The earth's atmosphere provides an effective shield against cosmic radiation; the dose rate doubles for every 1500-2000 meters increase in altitude. Dose rates at the edge of the atmosphere at high altitudes are around 1000 times higher; solar flares can increase these levels by orders of magnitude. High-flying aircraft such as Concorde carry radiation alarms to detect these solar surges which can result in dose rates as high as 10 mSv h -1 or more: in these conditions evasive action is required to reduce dose rates. The average crew exposure of subsonic commercial flights is estimated to be around 1 mSv per annum" for supersonic flights the average is around 2.5 mSv with a possible maxima of 15 mSv per annum. For astronauts who venture further out into the cosmos the doses incurred are higher, averaging 1 to 10 mSv per mission. Large solar events have the capacity to increase these up to 1000 mSv per mission. Population living permanently at high altitude e.g. Bogota, Quito, and Lhasa receive annual effective doses from cosmic radiation exceeding 1 mSv. The average annual effective dose resulting from external cosmic exposure to the population can vary appreciably with altitude (see Table 3.6 for some figures from the USA). The average value at sea level of 270 ~tSv per annum has an estimated individual range of from 150 to 5000 ~tSv per annum according to location and human activities and includes a 10 ~tSv contribution due to exposure during air travel.
50
Chapter 3
Table 3.6 Average annual effective dose variation with altitude Location (m)
Altitude (uSv/annum)
Average effective dose
Sea level Denver Leadville
1 1600 3200
270 500 1250
The number of passenger kilometres flown throughout the world in scheduled commercial flights was 9,341,01 1 in 1978 (International Civil Aviation Org., 1978). Taking the average speed to be 600 km h -~, a total of about 1.6 109 passenger hours was spent travelling in that year. The dose rates incurred during the flights vary according to the altitude and, to a smaller extent, to the latitude and to the solar activity. Table 3.7 shows the variation with altitude, from 4 to 20 km, of the dose rate and of the dose equivalent rate averaged over two geomagnetic latitudes and two periods of solar activity (O'Brien, 1975). The altitudes of subsonic flights depend on the type of aircraft used and on the distances covered in a given flight. They could be as low as 2-3 km for short flights and as high as 12 km for intercontinental flights, with intermediate values of 5 - 1 0 k m for medium range and continental flights (Hsu and Weng, 1976; Wachsmann and Regulla, 1978; Wallace, 1975). Table 3.7 shows that the dose rate and the dose equivalent rate vary by a factor of 20 between the altitudes of 4 and 12 km. Assuming that the average altitude of commercial flights is 8 km, the average dose rate would be 0.84 ~tGy h -~ and the average dose equivalent rate 1.35 ~tSv h -~, yielding a collective effective dose equivalent to the world population of about 2000 man Sv from air transportation in 1978. Table 3.7 Variation of the galactic dose rate and dose equivalent rate with altitude* (after O'Brien, 1975) Altitude (km)
Absorbed dose rate (~tGyh-l)
Dose equivalent rate (~tSvh-1)
4 6 8 10 12 14 16 18 20
0.14 0.33 0.84 1.75 3.01 4.61 5.92 7.09 7.72
0.20 0.51 1.35 2.88 4.93 7.56 9.70 11.64 12.75
Values averaged over 2 geomagnetic latitudes (43 ~ and 55 ~ and over two periods of solar activity (minimum and maximum).
Technologically Modified Exposure to Natural Radiation
51
Supersonic aircraft (SST), which are used on a small scale, fly at altitudes ranging up to 20 km, compared with at most 12 km for standard jet aircraft. Assuming that SSTs fly at an average altitude of 16 km, the average absorbed dose rate would be about 6 ~tGy h -! (Table 3.7). Actually, the absorbed dose rates measured on board the commercial SST flights of Air France, averaged over the years 1976-1980, amount to about 5 ~tGy h -~. Taking from Table 3.7 the quotient of the dose equivalent to the absorbed dose to be 1.6 Sv Gy -~, the corresponding dose equivalent rate is about 8 ~tSv h-~. Wallace (1975) calculated absorbed doses to passengers for a round trip, for both subsonic and supersonic transport between various city pairs. Some of these estimates are shown in Table 3.8. Doses for a round trip in supersonic aircraft are approximately 70% of those for subsonic speeds, because of the shorter flying time. However, the dose rates in supersonic aircraft are about twice as high as in subsonic aircraft. For a round trip across the Atlantic, the tissue-absorbed doses in passengers may be estimated to be about 2 10-5 Gy for an SST and 3 10-5 Gy for a subsonic aircraft, under average solar conditions. All the dose values given above refer to galactic cosmic rays. There is, in addition, a contribution due to the solar flares. From dose rate values given in the 1972 UNSCEAR report (United Nations, 1977), based on results obtained by an ICRP working group by averaging the effects of solar flares over the period 1952-1960 (International Commission for Radiation Protection, 1966), the average absorbed dose index rates from solar radiation can be estimated to be 4 10-8 Gy h -1 at 12 km and 9 10-7 Gy h -1 at 20 km. The average contribution from this source is thus small compared with that from galactic cosmic rays. Although radiation of solar origin does not contribute significantly to the average absorbed dose index rate, during an occasional intense solar flare radiation levels at these altitudes may increase by several orders of magnitude. The giant solar flare events last only for about 10 h and occur a few times in each solar cycle, and therefore are not likely to add significantly to the collective dose of the world population. It is worth Table 3.8 Comparison of calculated cosmic-ray doses to a person flying in subsonic and supersonic aircraft (average solar conditions) (after Wallace, 1975) Route
Subsonic flight at 11 km
Supersonic flight at 19 km
Flight duration (h)
Dose per round trip (10 -5 Gy)
Flight duration (h)
Los Angeles-Paris
11.1
4.8
3.8
3.7
Chicago-Paris
8.3
3.6
2.8
2.6
Dose per round trip (10 -5 Gy)
New York-Paris
7.4
3.1
2.6
2.4
New York-London
7.0
2.9
2.4
2.2
Los Angeles-New York
5.2
1.9
1.9
1.3
Sydney-Acapulco
17.4
4.4
6.2
2.1
52
Chapter 3
Table 3.9 Absorbed dose rates of astronauts on space missions (after Curtis, 1974; English et al., 1975; Radke, 1969; Grigoriev, 1976) Mission or mission series
Launch date
Duration of mission (h)
Typeof orbit
Dose (10-5 Gy)
Apollo VII Apollo VIII Apollo IX Apollo X Apollo XI Apollo XII Apollo XIV Apollo XV Vostok 1-6 Voshkad 1,2 Soyuz 3-9
Aug. 1968 Dec. 1968 Feb. 1969 May 1969 July 1969 Nov. 1969 Jan. 1971 July 1971
260 147 241 192 182 236 209 286
Earth orbital Circumlunar Earth orbital Circumlunar Lunar landing Lunar landing Lunar landing Lunar landing Earth orbital Earth orbital Earth orbital
120 185 210 470 200 -200 -500 -200 2-80 30,70 62-234
mentioning that SST aircraft carry radiation monitors, and the pilots will move the aircraft to lower altitudes when the dose rate reaches a prescribed level. When travelling into space, astronauts are subjected to primary cosmic ray particles, the radiation from solar flares, and also the intense radiation present in the two radiation belts. Savun et al. (1973) have reported measurements in the radiation belts in 1971. Measurements inside a 0.7 g cm -2 shield indicate that the maximum absorbed dose rate crossing the inner belt was 0.22 Gy h -1 and crossing the outer belt 0.054 Gy h -1. Estimated absorbed doses received by astronauts on several Apollo missions (average for the three occupants) based upon measurements carded out with tissueequivalent ionization chambers are shown in Table 3.9. (compiled from ref. Curtis, 1974; English et al., 1975; Radke, 1969; Grigoriev, 1976). A large part of this dose was received while the spacecraft was passing through the earth's radiation belts. For example, the higher dose received on the Apollo X mission was largely due to a different trajectory through the radiation belts. Analogous data in Table 3.9 from space flights of the USSR (Vostok, Voskhad and Soyuz series) indicate doses of comparable magnitude (Grigoriev, 1976). In outer space, remote from the shielding influence of the earth' s magnetic field, the absorbed dose index rate from solar protons emitted during solar flares can be very high. For example, it has been estimated that the absorbed dose indices in outer space from the solar proton event of 10 July 1959 were: from protons 3.6, 1.7, and 0.4 Gy behind shielding of 1, 2 and 5 g cm -2, respectively, and from alpha particles the corresponding values were 1.5, 0.3 and 0.05 Gy, respectively (Curtis, 1974). However, the Apollo missions did not experience any measurable solar particle events (English et al., 1975).
Technologically Modified Exposure to Natural Radiation
53
3.4 M I S C E L L A N E O U S SOURCES OF RADIATION 3.4.1 Radiation exposures due to geothermal energy production Geothermal energy is produced in Iceland, Italy, Japan, New Zealand, Russia and the United States. At the present time, it accounts for only 0.1% of the world's energy production (United Nations, 1981) but its relative importance may grow in the future as the potential resources of geothermal energy are believed to be very large. In geothermal energy extraction, use is made of hot steam or water derived from high-temperature rocks deep inside the earth. The geothermal fluids carry natural radionuclides and especially 222Rn, which is discharged into the atmosphere. From measurements of 222Rn activity concentrations in the hot stream used in three Italian power plants, the 222Rn annual releases have been estimated to be 110 TBq from the 400 MW Larderello plant, 7.0 TBq from the 15 MW plant at Piancastagnaio, and 1.5 TBq from the 3 MW plant at Bagnore (Mastinu, 1980). These figures point to an average 222Rn atmospheric discharge per unit energy generated of about 400 TBq per GWa. The corresponding collective effective dose equivalent commitment per unit energy generated is estimated to be about 6 man Sv per GWa if the assumptions used for the discharges from coal-fired power plants are applied. It is recalled that these assumptions are: equilibrium factor of 0.6 between 222Rnand its short-lived decay products; population density of 100 km -2 around the plant; effective dose equivalent per unit activity inhaled of 1.3 10-8 Sv Bq -~', indoor concentrations equal to the outdoor concentrations. Annual individual effective dose equivalents resulting from inhalation of shortlived decay products of 222Rn have also been estimated. Assuming, as in the case of discharges from coal-fired power plants, an effective stack height of 100 m and an annual average of the ground level air concentration per unit release rate of 4 10-8 Bq m -3 per Bq S-1 at 1 km from the stack, the annual effective dose equivalent resulting from atmospheric 222Rn discharges would be about 3 10-5 Sv for an individual of the critical group living around a geothermal plant of 1 GW of electrical power. It is to be noted that the existing geothermal plants have a lower power, resulting in correspondingly lower estimates of annual effective dose equivalents.
3.4.2 Consumer products The consumer products containing deliberately incorporated radionuclides can be broadly classified into five categories; radioluminous products (Maghissi et al., 1975, 1978); electronic and electrical devices (Ristagno, 1978); antistatic devices (Niemeyer et al., 1978; Webb et al., 1975); smoke detectors (Belonger et al., 1979; Johnson, 1978) and ceramic, glassware (Goldman and Yaniv, 1978), alloys, etc. containing uranium or thorium. Some of these products, such as the antistatic devices, are more widely disseminated in industry than among the general public. Table 3.10 presents some information on the number of products and the activities involved in each category in Germany (Wehner, 1978). Although the data shown are
54
Chapter 3
Table 3.10 Consumer products in Germany; data refer to the years 1973 or 1975, depending on the product (after Wehner, 1978) Type of consumer product
Produced in Germany No. of pieces Total activity and or weight radionuclide used
Exported
No. of pieces imported into Germany
Radioluminous timepieces
Devices containing scales or dials with luminous paint
14 106
40 TBq 3H 10 TBq 147pm
50%
8 105 (3H) 1 105 (147pm)
Electronic and electrical devices
High-pressure mercury lamps
7 106
15 GBq 232Th
20%
Ignition devices for fluorescent lamps
26 106
3 TBq 85Kr
50%
Electronic components containing 40 106 radioactive substances 11 106 3 106
200 TBq 85Kr 10 TBq 3H or 147pM 0.2 GBq 232Th
40%
Electronic tubes
7 105
3H, 6~ ' 63Ni 147pm' 226Ra
Antistatic devices
9
21~
Smoke detectors
1 105
226Ra' 241Am
Articles with uranium paints
3 105
0.6 GBq 238U
50%
1 106
Glassware containing uranium
4 103 kg
2 GBq 238U
50%
3 105
Glassware containing thorium
16 103 kg
7 GBq 232Th
10%
3 104
Ceramic, glassware, alloys etc. containing uranium or thorium
not contemporaneous, as some correspond to the year 1973 and others to the year 1975, Table 3.10 gives a good idea of the relative importance of each category in an industrialized country.
REFERENCES Aigueperse, J., Chalabreysse, J., Coulon, R. et al., Impact radiologique des rejets atmospheriques d'une centrale au charbon, p. 195-214 in: Health Impacts of Different Sources of Energy. IAEA, Vienna, 1982. Arkhipov, N.P., Fedorova, T.A., Fedorov, E.A., Fevraleva, L.T. and Tyumentseva L.M., Natural radionuclide content alteration in soils by regular phosphoric fertilization (in Russian). Prochvovedenie 12 (1981) 52-61. Aswathanarayana, U., Natural radiation environment in the Minjingu phosphorite area, Northern Tanzania. In: Laag, J. (ed.), Health Problems in Connection with Radiation from Radioactive Matter in Fertilizers, Soils and Rocks, pp. 79-85. Universitetsforlaget, Oslo, 1988.
Technologically Modified Exposure to Natural Radiation
55
Baran, V. and Kuca, L., Method of recovering Uranium and/or rare earths by solvent extraction (in Czech). CS patent 252226/B 1. Bayliss, R.J. and Whaite, H.W., A study of the radium alpha-activity of coal, ash and particulate emission at a Sydney power station. Air Wat. Pollut. Int. J. 10 (1966) 813-819. Beck, H.L., Gogolak, C.V., Miller, K.M. et al., Perturbations on the neutral radiation environment due to the utilization of coal as an energy source, pp. 1521-1558 in: Natural Radiation Environment III. CONF-780422 (Vol. 2) 1980. Bedrosian, P.H., D.G. Easterly and S.L. Cummings. Radiological survey around power plants using fossil fuel. EERL-71-3, 1970. Belanger, R., Buckley, D.W., and Swensen, J.B., Environmental assessment of ionization chamber smoke detectors containing Am-241. NUREG CR- 1156, 1979. Blackburn, R., and Gueran, J., Annual rate of release of radon to the atmosphere as a result of coal combustion in the United Kingdom. Radiation Phys. Chem. 13 (1979) 145-147. Both, G.F., The need for radiation controls in the phosphate and related industries. Health Phys. 32 (1977) 285-290. Camplin, W.C., Coal-fired power stations--the radiological impact of effluent discharges to atmosphere. NRPB-R 107, 1980. Can, S., Evaluation of phosphoric acid for the recovery of Uranium as a by-product. Turk. J. Nucl. Sci., 11 (1984) 108-120. Cathcart, J.B., Uranium in phosphate rock, Geological survey Professional paper, No. 988-A, Washington, D.C.: Geological Survey. Coles, D.G., Ragaini, R.C., and Ondov, J.M., Behavior of natural radionuclides in western coal-fired power plants. Environ. Sci. Technol. 12 (1978)442-446. Curtis, S.B., Radiation physics and evaluation of current hazards. In: C.A. Tobias and P. Todd (eds.), Space Radiation Biology and Related Topics. Academic Press, 1974. De Marteau J.M. and Sorale, S., Process for recovering Uranium contained in phosphate compounds. Report RFP-ABST-09. Rockwell Int. Co., Rocky Flats Plant, 1983. Delon, A. and Lazarevic, M., Possibilities for recovery of Uranium as a by-product in the production of phosphate fertilizers and tripolyphosphate. In: The Recovery of Uranium, pp. 351-361, IAEA, Vienna, 1971. Dennis, R.S., Process for recovery of Uranium from wet process H3PO4. US patent document 4.466.944/A. Washington: US Commissioner of Patents, 1984. Efimova, Z.I., Smirnov, Yu.V. and Sokolova, I.D., Development of new methods of Uranium separation from non-tradition sources. (In Russian) At. Tekh. Rubezhom. 4 (1987) 3-9. Eisenbud, M. and Petrow, H.G., Radioactivity in the atmospheric effluents of power plants that use fossil fuels. Science 144 (1964) 288-289. English, R.A., Bailey, J.V. and Brown, R.D., Application of Apollo cosmic radiation dosimetry to lunar colonization studies, p. 79-89 in: Natural Radiation Environment II. CONF-720805-P1, 1975. Fisher, H., Herzer, W. and Hettwig, B., Radiation doses due to fabrication of fertilizers in Nordenham (in German). Bremen University, Germany, 1979. Fitzgerald, J.E. and Sensitaffar, E.L., Radiation exposure from construction materials utilizing by-product gypsum from phosphate mining. In: Mogihissi et al. (eds.), Radioactivity in consumer products, pp. 351-368. Report NUREG/CP-0001. Washington: Nucl. Reg. Commission, 1978. Furr, A.K., Parkinson, T.F., Hinrichs, R.A. et al., National survey of elements and radioactivity in fly ashes. Absorption of elements by cabbage grown in fly ash-soil mixtures. Environ. Sci. Technol. 11 (1977) 1194-1201. Gluskoter, H.J., Ruch, R.R., Miller, W.G. et al., Trace elements in coal: occurrence and distribution. Illinois State Geological Survey Circular 499, 1977. Goldman, M. and Yaniv, S.S., Naturally occurring radioactivity in ophthalmic glass. In: Radioactivity in Consumer Products. pp. 227-240, NUREG/CP-0001, 1978. Goldstein, N.P., Sun, K.H. and Gonzales, L.J., Radioactivity in fly-ash from a coal-burning power plant. Trans. Am. Nucl. Soc. 14 (1971) 66.
56
Chapter 3
Gorecka, H. and Gorecki, H., Determination of uranium in phosphogypsum. Talanta 31 (1984) 1115-1117. Grigoriev, Yu.G., Problems of space radiobiology. In: A.M. Kuzin (ed.), Problems of Radioecology and Biological Effects of Low Doses of Ionizing Radiation, pp. 9-16. USSR Academy of Sciences, Sykryvkar, 1976. Guimond, R.J. and Hardin, J.M., Radioactivity released from phosphate-containing fertilizers and from gypsum. Radiat. Phys. Chem. 34 (1989) 309-315. Guimond, R.J. and Windham, S.T., Radioactivity distribution in phosphate products, by-products, effluents and wastes. ORP/CSD-75-3, 1975. Guimond, R.J. and Windham, S.T., Radiological evaluation of structures constructed on phosphate-related land. In: Natural Radiation Environment III. CONF-780422 (Vol. 2) pp. 1457-1475, 1980. Guimond, R.J., Radium in fertilizers. In: IAEA. The Environmental Behavior of Radium. pp. 113-128. Technical Report Series No. 310. IAEA, Vienna, 1990. Guimond, R.J., The radiological aspects of fertilizer utilization. In: Moghissi et al. (eds.), Radioactivity in Consumer Products, pp. 380-393. Report NUREG/CP-0001. Nucl. Reg. Commission, Washington, 1978. Habashi, F., Awadella, F.T. and Zailaf, M., The recovery of Uranium and the Lanthanides from phosphate rock. J. Chem. Technol. Biotechnol. 36 (1986) 259-266. Halbritter, G., Br~iutigam, K.R., Fluch, F.W. et al., Contribution to a comparative environmental impact assessment of the use of coal and nuclear energy for electricity generation for selected site conditions in the FRG. In: Health Impacts of Different Sources of Energy, pp. 229-247. IAEA, Vienna, 1982. Hamilton, E.I., The chemical elements and human morbidity--water, air and places--a study of natural variability. Sci. Total Environ. 3 (1974) 3-85. Hayumbu, P., Haselberger, N., Markowicz, A. and Valkovic, V., Analysis of Rock Phosphates by X-ray Fluorescence Spectrometry, Appl. Radiat. Isot., 46 (1995) 1003-1005. Hsu, P.C. and Weng, P.S., Radiation exposure during air and ground transportation. Health Phys. 31 (1976) 522-524. Hurst, F.J., Uranium from phosphate ores. Oak Ridge Nat. Lab. Report CONF-830788-2. Inoue, K. and Nakashio, F., Technology of Uranium recovery from wet-process phosphoric acid (in Japanese). Kemikaru Enjiniaringu 27 (1982) 59-65. International Atomic Energy Agency: Technical Reports Series No. 295; Measurements of radionuclides in food and the environment, a Guidebook, IAEA, Vienna, 1986. International Civil Aviation Organization. Annual report of the Council, 1978. International Commission on Radiological Protection. A report of the ICRP Task Group on the Biological Effects of High-energy Radiations. Radiobiological aspects of the supersonic transport. Radiobiological aspects of the supersonic transport. Health Phys. 12 (1966) 209-226. Jacobi, W., Schmier, H. and Schwibach, J., Comparison of radiation exposure from coal-fired and nuclear power plants in the Federal Republic of Germany. In: Health Impacts of Different Sources of Energy, pp. 215-227. IAEA, Vienna, 1982. Jacobi, W., Umweltradioaktivit~t und Strahlenexposition durch radioaktive Emissionen von Kohlekraftwerken. GSF-Bericht S-760, 1981. Jaworowski, Z., Bilkiewicz, J. and Zylica, E., 226Ra in contemporary and fossil snow. Health Phys. 20 (1971) 449-450. Johnson, J.E., Smoke detection containing radioactive materials. In: Radioactivity in Consumer Products, pp. 134-440. NUREG/CP-0001, 1978. Jordan, S. and Schikarski, W., Evaluation of radioactive and non-radioactive trace constituents emitted from fossil-fuel and nuclear power plants. In: Environmental Behavior of Radionuclides Released in the Nuclear Industry, pp. 525-535. IAEA, Vienna, 1973. Kakinen, J.W., Jordan, R.M., Lawasani et al., Trace element behavior in coal-fired power plant. Environ. Sci. Technol. 9 (1975) 862-869. Kirchmann, R., Darcheville, M. and Koch, G., Accumulation of Radium-226 from phosphate fertilizers in cultivated soils and transfer to crops. In: Natural Radiation Environment III. CONF-780422 (Vol. 2) pp. 1667-1682, 1980.
Technologically Modified Exposure to Natural Radiation
57
Klein, D.H., Andren, A.W., Carter, J.A. et al., Pathways of thirty-seven trace elements through coal-fired power plant. Environ. Sci. Technol. 9 (1975) 973-979. Komura, K., Sakanone, M., Yanagisawa, M. and Sakurai, J., Uranium, Thorium and Potassium contents and radioactive equilibrium states of the U and Th series in phosphate rock and phosphate fertilizers (in Japanese). Radioisotopes (Tokyo) 34 (1985) 529-536. Lardinoye, M.H., Weterings, K. and Berg, W.B., Unexpected 226Rabuild-up in wet-process phosphoric acid plants. Health Phys. 42 (1982) 503-514. Lee, H., Peyton, T.O., Steele, R.V. and White, R.K., Potential Radioactive Pollutants Resulting from Expended Energy Programs, SRI-EGU-4869, Stanford Research Institute, Menlo Park, CA, 1977. Lindeken, C.L. and Coles, D.G., The Radium-226 content of agricultural gypsums. In: Mogihissi et al. (eds.), Radioactivity in Consumer Products, pp. 369-375. Report NUREG/CP-0001. Washington: Nuclear Reg. Commission, 1978. Lisachenko, E.P. and Obuhova, O.L., The radioactivity of coals, ashes and slags in the USSR. Presented at the 2nd Special Symposium on the Natural Radiation Environment. Bombay, 1981. Londhe, V.S. and Rao, S.R., Study of distribution of some neutral radionuclides on processing of rock phosphate. Bull. Radiat. Prot. 11 (1988) 181-186. MacCready, W.L., Wethington, J.A. and Hurst, F.J., Uranium extraction from Florida phosphates. Nucl. Technol. 53 (1981) 344-353. Markowicz, A. and Haselberger, N., A modification of the emission-transmission method for the determination of trace elements by XRF. Int. J. Radiat. Appl. Instr. Part A. 43 (1992) 777-779. Martin, J.E., Harvard, E.D., Oakley, D.T. et al., Radioactivity from fossil-fuel and nuclear plants. In: Environmental Aspects of Nuclear Power Stations, pp. 325-337. IAEA, Vienna, 1971. Martinez-Aguirre, A., Garcia-Leon, M. and Ivanovich, M., The distribution of U, Th and 226Raderived from the phosphate fertilizer industries on an estuarine system in southwest Spain, J. Environ. Radioactivity 22 (1994) 155-177. Mastinu, G.G., Natural radioactivity levels in releases from coal-fired power plants in Italy. In: Seminar on the radiological burden of man from natural radioactivity in the countries of the European Communities, pp. 429-436. CEC report V/2408/80, 1980. Mastinu, G.G., The radiological impact of geothermal energy. In: Seminar on the radiological burden of man from natural radioactivity in the countries of the European Communities, pp. 437-445. CEC Report V/2408/80, 1980. McBridge, J.P., Moore, R.E., Witherspoon, J.P. et al., Radiological impact of airborne effluents of coal-fired and nuclear power plants. ORNL-5315, 1977. McNabb, G.J., Kirk, J.A. and Thompson, J.L., Radionuclides from phosphate-ore-processing plants: the environmental impact after 30 years of operation. Health Phys. 37 (1979) 585-587. Menzel, R.G., Uranium, Radium and Thorium content in phosphate rocks and their possible radiation hazard. J. Agric. Food Chem., 16 (1968) 231-234. Metzger, R., McKleen, J.W., Jenkins, R. and McDowell, W.J., Specific activity of Uranium and Thorium in marketable rock phosphate as a function of particle size. Health Phys. 39 (1980) 69-75. Metzger, R., Uranium and Thorium correlated with particle size in phosphate fertilizers. Trans. Am. Nucl. Soc. 33 (1979) 230-231. Mishra, U.C., Lalit, B.Y. and Ramachandran, T.V., Radioactivity release to the environment by thermal power stations using coal as a fuel. Sci. Total Environ. 14 (1980) 77-83. Moere, H., Distribution of natural radioactivity through the use of phosphate containing fertilizers in agriculture (in Swedish) Report SSI:020, Stockholm: National Institute of Radiation Protection. Moghissi, A.A. and Carter, M.W., Public health implications of radioluminous materials. DHEW (FDA) 76-8001, 1975. Moghissi, A.A., Carter, M.W., Simpson, R.E. et al., Evaluation of public health implications of radioluminous materials. In: Radioactivity in Consumer Products, pp. 256-276. NUREG/CP-001, 1978. Mortvedt, J.J. and Sikora, F., Heavy metals, radionuclides and fluorides in phosphorus fertilizers. In: Sikora, F. (ed.), Future Directions for Agricultural Phosphorus Research, pp. 69-73. TVA Bulletin Y.224. National Fertilizer and Environmental Research Center, Muscle Shoals, AL, 1992. Mustonen, R., Radioactivity of fertilizers in Finland. Sci. Total Environ. 45 (1985) 127-134.
58
Chapter 3
Nash, J.D., Working with the phosphate factor. Environment 29 (1987) 13. Niemeyer, R.G., Case, F.N. and Cutshall, N.H., Evaluation of Polonium-210 static eliminators. In: Radioactivity in Consumer Products, p. 423-433. NUREG/CP-0001, 1978. Nishiwaki, Y., Tsunetoshi, Y., Shimizu, T. et al., Atmosphere contamination of industrial areas including fossil-fuel power stations, and a method of evaluating possible effects on inhabitants. In: Environmental Aspects of Nuclear Power Stations, pp. 247-278. IAEA, Vienna, 1971. O'Brien, K., The cosmic ray field at ground level. In: The Natural Radiation Environment II, pp. 15-54. CONF-720805-P1, 1975. Osmond, J.K., Cowart, J.B. Humphreys, C.L. and Wagner, B.E., Radioelement migration in natural and mined phosphate terrains. Final Report PB-86-223765/XAB. Univ. Tallahassee, Florida State University, USA, 1987. Partridge, J.E., Horton, T.R., Sensintaffar, E.L. and Boysen, G.A., Radiation dose estimates deu to air particulate emissions from selected phosphate industry operations. ORP/EERF-78-1. Office of Radiation Programs, EPA, Montgomery, AL, USA, 1978. Paschoa, A.S., Mafra, O.Y., Cardoso, D.O. and Rocha, A.C.S., Application of solid state nuclear track detectors (SSNTD) to the Brazilian phosphate fertilizer industry to determine uranium concentration. Nucl. Tracks 8 (1984) 469--472. Paul, A.C., Pillai, P.M.B., Komalan Nair, S. and Pillai, K.C., Studies on the leaching of Radium and the emanation of Radon from fertilizer process sludge. J. Environ. Radioact. 1 (1984) 51-65. Petrache, C.A., Marcelo, E. and Santos, G., Notes on the extraction of Uranium from phosphoric acid. Report PAEC (B)-87001. Philippines Atomic Energy Commission, Quezon City, 1987. Pfister, H. and Pauly, H., External radiation exposure due to natural radionuclides in phosphate fertilizers in the Federal Republic of Germany. In: Seminar on the radiological burden of man from natural radioactivity in the countries of the European Communities. CEC report V/2408/80. pp. 447-467, 1980. Pfister, H., Phillipp, G. and Pauly, H., Population dose from natural radionuclides in phosphate fertilizers. Rad. Environ. Biophys. 13 (1976) 147-261. Philipp, G., Pfister, H. and Pauly, H., Occupational radiation exposure from natural radionuclides in phosphate fertilizers and its contribution to the exposure of the population in the FRG (in German). In: Kellermann H.J. (ed.), Radioactivity and Environment, Vol. 12, pp. 890-901. Fachhverband fiar Strahlenschutz, Karlsruhe, 1978. Radiological Impact of Uranium Recovery in the Phosphate Industry. Nuclear Safety Vol. 22, No. 1, 1981. Radke, G., Solar flare dose rates in a near earth polar orbit, in: (J.F. Janni and R.E. Holly (eds.) The Current Experimental Approach to the Radiobiological Problems of Space Flight. Aerosp. Med. 40 (1969) 1495-1503. Ring, R.J., Manufacture of phosphatic fertilizers and recovery of by-product Uranium--a review. Report AAEC/E-355. Australian A.E.C. Res. Establ., Lucas Heights, 1975. Ristagno, C.V., The use of tritium luminous sources for lighting digital wristwatches. In: Radioactivity in Consumer Products. NUREG/CP-0001, pp. 320--322, 1978. Roessler, C.E., Control of radium in phosphate mining, benefication and chemical processing. In: IAEA. The Environmental Behavior of Radium, pp. 269-279. Technical Reports series no. 310. IAEA, Vienna, 1990. Roessler, C.E., Smith, Z.A., Bolch, W.E. and Prince, R.J., Uranium and 226Radium in Florida phosphate materials. Health Phys. 37 (1979) 269-277. Rothbaum, H.P., McGaveston, D.A., Wall, T., Johnston, A.E. and Mattingly, G.E.G., Uranium accumulation in soils from long-continued applications of superphosphate. J. Soil. Sci. 30 (1979) 147-153. Saleh, N.S. and A1-Saleh, K.A., Analysis of Jordanian phosphate using nuclear techniques. Appl. Phys. Commun. 7 (1987) 313-327. Santos, P.L., Gouvea, R.C. and Dutra, I.R., Human occupational radioactive contamination from the use of phosphated fertilizers. Sci. Total Environ. 162 (1995) 19-22. Savun, O.I., Senchuro, I.N., Shavrin, P.I. et al., Distribution of radiation dose in the radiation belts of the earth in the year of maximum solar activity. Kosm. Issled 11(1) (1973) 119-123. Shishkunova, L.V., Grashchenko, S.M. and Strukov, V.N., Entry of Uranium, Thorium and Radium
Technologically Modified Exposure to Natural Radiation
59
isotopes into plants from soils and fertilizers. Sov. Radiochem. 30 (1989) 362-367. Styron, C.E., An assessment of natural radionuclides in the coal-fuel cycle. In: Natural Environment III, p. 1511-1520. CONF-780422 (Vol. 2), 1980. Styron, C.E., Casella, V.R., Farmer, B.M., Hopkins, L.C., Jenkins, P.H., Phillips, C.A. and Robinson, B., Assessment of the radiological impact of coal utilization, Report MLM-2514, UC-90a, 1979. Tomczynska, J., Blaton-Albicka, K., Pensko, J. et al., The results of measurements of the natural radionuclides in coal power plants wastes and light concrete samples. Nukleonika, Warsaw, 1981. Torrey, S. (ed.), Trace Contaminants from Coal. Noyes Data Corporation, Park Ridge, NY, 1978. United Nations Scientific Committee on the Effects of Atomic Radiation: Sources, effects and risks of ionizing radiation. U.N., New York, 1982. United Nations. 1979 Yearbook of World Energy Statistics. New York, 1981. United Nations. Sources and Effects of Ionizing Radiation. United Nations Scientific Committee on the Effects of Atomic Radiation 1977 report to the General Assembly, with annexes. United Nations publication (Sales No. E.77.IX.I). New York, 1977. Van Cleemput, O., De Vriendt, D., Baert, L., Zapata, F. and Van Maercke, H., Radioactivity in agricultural phosphate-containing compounds, Med. Fac. Landbouw Univ. Gent., 58/1 (1993) 11-16. Vucic, N., and Ilic, Z., Extraction and spectrophotometric determination of uranium in phosphate fertilizers. J. Radioanal. Nucl. Chem. Articles 129 (1989) 113-120. Wachsmann, F. and Regulla, D.F., Exposure of aircraft passengers to radiation from transported radioactive goods. Kerntechnik 20 (1978) 318-322. Wallace, R., Measurements of the cosmic radiation dose in subsonic commercial aircraft compared to the city-pair dose calculation. LBL-1505, 1975. Watanabe, A., Lutz, E.J. and Moghissi, A.A., Atmospheric releases from fossil fuel power plants in the United States. Environ. Int. 4 (1980) 357-382. Webb, G.A.M., Wilkins, B.T. and Wrixon, A.D., Assessment of the hazard to the public from antistatic brushes containing Polonium-210 in the form of ceramic microspheres. NRPB-R36, 1975. Wehner, R., Legal and practical aspects of radioactivity in consumer products in the Federal Republic of Germany. In: Radioactivity in Consumer Products, pp. 97-105. NUREG/CP-0001, 1978. Zapata, F. and Axmann, H., Agronomic evaluation test of rock phosphate materials by means of radiosotope techniques. P6dologie XLI-3 (1991) 291-301.
61
CHAPTER 4
Man-made Radioactivity
4.1 I N T R O D U C T I O N The experiment of initiating nuclear transformation artificially was first carried out by Rutherford in 19 19. It was shown that when a particle emitted from 214po w a s absorbed on striking a nitrogen atom, oxygen and proton were produced by the following reaction: J4N((x,p)JTo
or
14N + 4He --~ 170 + JH
(4.1)
All nuclides obtained by Rutherford were the stable nuclides. In 1934 J. Curie first succeeded in artificially producing phosphorus-30 by bombarding aluminium with c~-particles. This was the reaction: 27Al(~,n)3~
or
27A1+ 4He .___)30p + 1n
(4.2)
This was the first man-made radionuclide. From that time on many species of radionuclides were produced by bombardment of elements with charged particles using the various types of accelerators. In addition, practical use of fission energy allowed production of a great amount of artificial radionuclides, not only by neutron irradiation generated with nuclear reactors, but also by processing spent fuel. Nuclear reactions are usually represented as eqs. (4.1) and (4.2), that is, on the left is the symbol for the target nuclides, the first symbol in the parenthesis indicates the bombarding particle (or projectile), the second the emitted particles, and the symbol of the product on the right. In the equation, the left side of the comma shows the system of the reactants, and the right the system of the products. Before and after the nuclear reaction, both the sum of mass number and the sum of atomic number remain unchanged. When a target, as a thin film, is exposed to a bombarding particle (projectile) at the flux [cm 2 s-~] or ~ for t s, the number of the target nuclide, N, decreases by the equation of
Chapter 4
62
dN dt
= - NOr
N = Noe-"*'
(4.3)
where o is the quantity defined by the nuclear reaction in question and the energy of the bombarding particle, and is called cross section which has the dimension of cm 2. Hence, N Ois N at t = 0. When the disintegration constant of the nuclide produced is s-1, the rate of production of the nuclides is expressed as eq. (4.4): dN dt
= -Nor
)vn = N 0 o C e - * ' - n
(4.4)
Solving eq. (4.4) for n, N 0o_______0~ -z, n = )~ _ or [e-"*' - e ]
(4.5)
If )v >> o~) and Ct -- 0, the radioactivity produced, n, is given by eq. (4.5), which is important in practical use. n = No, ( 1 - e -~')
(4.6)
Nuclear fission is the phenomenon that occurs when a heavy nucleus splits into two or more intermediate heavy nuclei. In the field of radiochemistry, the fissions of 233U,235U, 239pu, induced by neutrons, are often treated, and especially that of 235U by thermal neutrons is studied the most. The fission cross sections of several nuclides for thermal neutrons are shown in Table 4.1.
Table 4.1 Fission cross sections for thermal neutron Isotope
o[b] [1 b = 10-24 cm 2]
23eTh 233U 235U 237Np 239pu 24~ 241pu 241Am 242mAm
3.9x 10-5 531.1 583.54 1.9• -2 742.5 3.0x 10-2 1009 3.15 6600
63
Man-made Radioactivity
10'
I
~
10
~
1
1
9
1
I
Neutrons ~ ~ , of 14 MeV
_
Thermal neutrons Z
]0 ~
10~
/ .-.-~ 1 70 80
10 ~
I
90
i
t
!
I
1
100 110 120 130 140 150 MASS NUMBER
Fig. 4.1. Mass number-fission yield curves for the fission of 2-'Uinduced by thermal neutrons and neutrons of 14 MeV. Even after emitting neutrons, all the fission recoils still have too many neutrons compared with the protons, and have a trend toward more stable nuclei by repeating [3disintegration. The series connected with 13-disintegration is called a fission chain. The final member in the series is a stable nucleus without fail. As an illustration, the fission yield for 235U is shown in Fig. 4.1 the fission yield curve for 235U induced by thermal neutrons has a deep valley at the centre compared with that by the 14 MeV neutrons. Since two fragments are produced for one fission, the sum of the yields fragments amounts to 200%. In nuclear reactions, the total momentum of the system is conserved before and after a nuclear reaction, and the product nuclide should generally gain some kinetic energy of recoiling at the emission of ),-ray or a particle. When the energy of incident or the ejected particles is more than 10 KeV, the target atom may gain the recoiling energy exceeding the chemical bond energy (several eV) and consequently the product nuclide is knocked out of the molecule. An atom possessing such a high recoiling energy is called a "hot atom" and the field concerned with this kind of chemical change accompanying nuclear reactions is called "recoil chemistry" or "hot-atom chemistry". The recoil phenomenon is often seen with (n,),) reactions which are known in terms of the Szilard-Chalmers reaction (1934). Although the capture of a thermal neutron
64
Chapter 4
does not provide enough energy to break a chemical bond, the emission of 7 ray following the neutron capture may cause recoiling. The recoiling energy E gained by a target atom with atomic mass M (in amu) is given as: e(eV) =
537E~/M
(4.7)
where Ey (MeV) is the energy of the y ray emitted. For instance 128Iproduced by a (n,7) reaction on ethyl iodide is recoiled and hence, when shaken with water, it is transferred into the water phase in carrier free state (ideally speaking). This is the first example of Szilard-Chalmers reactions which are now extensively studied and used for the production of some isotopes in high specific activity. There are some special methods of chemical synthesis recoiling phenomenon (recoil synthesis). For example anthracene labelled with ~4C is synthesised by irradiating acridine with neutrons through Z4N(n,p)J4C reaction. Similarly 3H labelling is realised with the use of 3He(n,p)3H or 6Li(n,o03H. For this purpose the compound to be labelled is mixed with 4He gas or Li2CO 3 and then irradiated with neutrons. The recoiling phenomenon is, of course, also seen in decay processes. The occurrence of this phenomenon with c~ decay has been well known since the early days and employed as a separation process. It also takes place with 13- decay, providing various nuclear chemical substances. 52
5225MnO s- ~
__
24CrO 4
(4.8)
The spallation reaction is another example of special nuclear reactions, by which many nuclides' relatively small mass number (about 10 to 20, the smaller the number, the higher the yield) in comparison with the target nuclide is produced simultaneously. The example is: 37CI( p, 6 p4 n) 2s 17 12Mg
(4.9)
The actinide elements are the elements with atomic numbers of 90 till 103 being members of a transition series, the first member of which is actinium (Z = 89); fourteen electrons are added successively, beginning formally with thorium (Z = 90) and ending with lawrencium (Z = 103). The straightforward way to obtain light actinides is by neutron irradiation of elements of lower atomic number. For example the production of Pa has been produced by the transmutation of Th with neutron produced in a high-flux nuclear reactor.
23~
_.._)231pa+
(4.10)
The elements Am and Cm also exist in spent fuel. However, large scale production is performed by neutron irradiation of 239pu.
Man-made Radioactivi~
65
The synthesis of transfermium elements beyond Fm is succeeded by nuclear reactions of charged particles with targets of an actinide element of a lower atomic number. 239pu(~,n)242Cm ' 241Am(~,2n)243Bk ' 238U(12C,5n)245Cf
(4.11)
The physical and chemical characteristics of actinides are as follows: 9 In an aqueous solution of pH < 3, four structural types of actinide cations exist, these are M 3§ M 4+, MO + and MO 2+ corresponding respectively to M(III), 2, 2 M(IV), M(V), and M(VI). ~ At a low [H § concentration, actinide ions tend to undergo hydrolysis. For example, uranium (IV) begins to undergo hydrolysis in aqueous solution above pH > 2.9. As the pH increases, U(IV) eventually precipitates as hydroxide, U(OH) 4. The actinide ions of the (IV) state are particularly prone to hydrolysis and polymerisation. 9 Actinide cations have a strong tendency to react with various inorganic and organic ions or agents, forming complex ions.
4.2 I S O T O P E S IN E V E R Y D A Y L I F E
Isotopes of chemical elements represent a tool which can do certain jobs more easily, quickly, simply, and cheaply than competitive methods. Some measurements could not be done at all without the use of isotopes as there are no alternative methods available. Isotopes are ideal tools for use in analysis; a single atom can be detected when using radioactive isotopes, as compared to chemical methods in which the detection limit of an element is enhanced a million times. Stable isotopes can also be detected with great accuracy nowadays; although not quite with the same sensitivity as radiation-emitting isotopes. Most important, especially in biological and medical work, is that radioisotopes can be located during a biological process. The functioning of certain glands can also be checked, by first administering a small amount of a radioisotope and then following the path of this compound in the body simply by measuring the radiation from the outside. For people who may worry about these small amounts of radioactivity, it should be remembered that everyone constantly eats potassium in their food, which is in itself slightly radioactive, and with which animals and humans have lived for a long time. For most of these applications--and there are many--there is no alternative method. Larger sources, which emit penetrating radiation, can be used as a portable X-ray unit to check welds in underground pipelines. Such sources are also used for certain analyses especially suited for work in the field, such as in geology. Very large sources, some 1000 million times stronger than the activities used as tracers, can destroy bacteria or other spoilage organisms in food, can be used for sterilisation of medical sutures or syringes, or can impart specific desirable properties to some materials.
66
Chapter 4
As isotope sources are relatively cheap, the instrumentation is readily available, and the application simple, they find wide application in practically all fields of science and industry. It is not surprising that the importance of the use of these tools, in spite of the growth of other new methods, is steadily increasing.
4.2.1 Food and agriculture In agricultural research and application, isotopes and radiation play a part in so many fields and in so many ways that it is difficult to obtain a proper picture of their enormous importance. In laboratories isotopes are used routinely with an ever-increasing assortment of modern research aids. In emerging biotechnologies, which are used increasingly by scientists, isotopes are a basic tool without which research in molecular biology could not be done. The main agricultural problems that isotopes and radiation are helping to solve are: 9 determination of conditions necessary for optimising fertiliser use and its efficiency for biological nitrogen fixation; 9 breeding of high performance, well adapted and disease resistant agricultural and horticultural crop varieties using radiation induced mutations; 9 eradication or control of insect pests using insects that have been radiation sterilised or genetically altered; 9 improvement of reproduction performance, nutritional status, and health of animals using radioimmunoassay and related techniques, as well as isotopic tracers; 9 reduction of post-harvest losses by suppressing sprouting and contamination using radiation treatment; 9 reduction of food-borne diseases and extension of shelf-life using radiation; and 9 study of the ways to reduce pollution from pesticides and agrochemicals. A good crop needs soil with adequate amounts of nutrients and moisture. Nuclear techniques are ideal tools for measuring the efficiency of fertiliser use by crops and for keeping a watch on the moisture content. In modern agriculture, the use of fertilisers is essential to maximise crop yields; for example, a 50% increase in grain yield of cereals is common in many soils through efficient fertilisation. In order to provide food for the constantly increasing world population, the projected fertiliser consumption in 20 years' time is estimated to be 4-5 times greater than today' s. To reduce the fertiliser requirement to an absolute minimum and thereby save production costs to the farmer and reduce damage to the environment, studies to obtain information on the relative merits of different fertilisation practices--such as methods of fertiliser placement, times of application, and types of fertilisersmare needed. The method used to solve these problems requires the introduction of known quantities of fertiliser labelled with isotopes to the soil at various times and in different positions. Since the plant does not discriminate between elements from the labelled fertiliser and those from native soil, the exact amount of fertiliser nutrients taken up by the plant can be measured.
M a n - m a d e Radioactivit 3,
67
The results of this type of research have been incorporated into agricultural practices for cereals and have increased crop productivity significantly, reduced fertiliser use--and thereby costs--and helped the environment by markedly reducing residual fertiliser in soils. Recommendations based on the results of experiments in this area have been adopted in FAO-organised fertiliser programmes in many countries and great savings have been reported; one country using these techniques claims to have saved as much as US $36 million per year on maize crops alone. Similar natural methods have been adapted to evaluate deposits of cheap rock phosphates as an alternative to expensive, often imported, phosphate fertilisers, and to find the most efficient way to use these fertiliser deposits for maximum plant growth. Although nitrogen constitutes 80% of gases in the atmosphere, few plants can directly make use of it. However, through fixation, plants are able to use the nitrogen in the air. The most important results are obtained from a symbiosis between a plant and a bacterium, which has gained great attention during recent years. Legumes that fix nitrogen can provide high protein for human and animal consumption and also increase nitrogen in soils. The water plant Azolla, for example, can drive 80-90% of its nitrogen by fixation, and is valuable in providing nitrogen to paddy rice crops. In order to obtain maximum benefits from this unique biological process, isotopes and used to find the amount of nitrogen that a plant can fix and how this process can be improved. Isotope techniques are an ideal tool to distinguish nitrogen derived from the atmosphere, soil, and applied fertiliser. Water is the most important limiting factor for crop production in many areas of the world. The efficient use of water in irrigation systems requires continuous monitoring of the moisture content of soil. Neutron moisture gauges are ideal instruments for this purpose and help soil physicists to make the best use of limited water resources. Through these methods, traditional irrigation methods are improved and in some cases up to 40% of the water can be saved. Agricultural production relies heavily on chemical inputs" fertilisers to boost production and pesticides to suppress weeds and control insects. Excessive use of these chemicals harms the environment as well as the food products. Isotopes are ideal tools for studying the behaviour, breakdown, and residues of agrochemicals in soil, water, plants, animals and their products. As a result of their use, it has been possible to devise safer ways to apply agrochemicals and safer formulations which are more effective in controlling pests or promoting growth, as well as being less harmful to health and the environment. For centuries, mankind tried every possible way to improve quantity and quality of crops. Natural evolution results from spontaneous mutation and selection of the fittest mutants. The rate of mutation occurrence can be multiplied by radiation treatment thereby accelerating evolution and the selection of superior crops. Over the last 50 years, a number of plant breeding programmes have included mutation induction with radiation or chemicals to breed improved crops. Physical mutagens like X-rays, gamma rays or fast neutrons are most frequently applied and their use has resulted in the highest number of improved, mutant crops. The
68
Chapter 4
Table 4.2 Economically important mutant varieties (source: IAEA) Crop
Variety
Country
Barley
Trumpf, Triumph Diamant, Krystal Midas Gratiot, Sanilac Arun NIAB78 Lumian No. 1 Star Rugby Wasata, Heiga, Jaran Stellar Calroise 76, M-401 Kashmir Basmati IRAT 13 RD 6 Atomita II Yuanfengzao Ahnsanffae Kalika CO 449, Co 997 Pervenets Casterporziano, Creso Cargidurox, Novosibirskays 67 Sirius Mv 8
Germany, UK CSFR UK USA India Pakistan China USA Poland Canada USA Pakistan Ivory Coast Thailand Indonesia China Korea India India USSR Italy France USSR Germany Hungary
Beans Castor bean Cotton Grapefruit Pea Rapeseed Rice
Sesame Sugar cane Sunflower
number of induced mutant derived crop varieties now exceeds 1500 worldwide with billions of dollars added to farmers' incomes annually. Some of the economically important mutant varieties are shown in Table 4.2. Important desirable properties which can be achieved by radiation include: 9 I m p r o v e d lodging resistance: the desired properties are a reduction in plant height and a stiffer stem, which can withstand rain and storm. 9 C h a n g e d maturing times: early maturing is important to escape frost, pests, etc., or simply to make room in the field for other crops. 9 I n c r e a s e d disease resistance: becoming very important in attempts to decrease the use of chemicals which are used against pests to protect the environment. 9 I n c r e a s e d yields: the yield of many crop varieties has been increased manifold after mutation breeding using nuclear techniques. ~ I m p r o v e d agronomic characters: for example, more winter hardiness, greater tolerance against heat, or generally better adaptability to available soil conditions.
Man-made RadioactiviO,
69
9 I m p r o v e d seed characteristics: improvement of nutritional value (protein or oil
content), baking and melting qualities, or reduction in cooking time. Many of the radiation induced mutants have made a great impact on the income of the region where they took place, in some cases even on the national income. One of the earliest successes concerns peppermint. The only source of peppermint oil in the United States was the Mitcham variety which succumbed to a fungus disease. Crossbreeding methods failed to produce disease-resistant peppermint. Radiation techniques led to induction of resistance which saved the original peppermint taste enjoyed by millions all over the world. Another remarkable success story of applying radiation to obtain economically significant mutants was achieved in Pakistan. There, a new cotton mutant was released by the Pakistan Atomic Energy Commission in 1983. It turned out to be the most productive variety in the country. The cotton production in Pakistan was roughly doubled! It is estimated that the crop value of this mutant during 1988-1989 was more than US $1600 million. However, not always does the success of a mutant show so quickly as in this case; sometimes it takes more than a decade before the usefulness of a new mutant is fully recognised (source: IAEA). In Italy, where nearly everyone eats pasta, 50% of pasta stems from a wheat variety developed through mutation. In China extraordinary results were achieved with this method: a certain type of rice now matures 24 days earlier, another type has a 20 cm shorter culm and in a third mutant a very high protein content was achieved (15.6%). Virtually hundreds, if not thousands, of such benefits have been developed over the last 10 years by artificially produced mutation in China alone where almost a tenth of the total crop acreage is under mutant-derived crops. The list of countries which have released crop varieties developed through induced mutations is impressive. There are more than 40 countries with over 1500 released mutants of which less than 10% are chemically induced and more than 90% are induced by radiation. Insects compete with man for food and fibre and are a threat to animal and human health. In controlling insects with chemicals, we have sometimes created problems of environmental pollution and toxic residues in our food. Also, many insects have developed resistance to insecticides, often resulting in more insecticide being used. Therefore, new approaches to insect control are needed. One way of controlling or eradicating insects without the use of chemicals is the sterile insect technique (SIT). In this approach to insect control, insects are produced in large rearing plants, sexually sterilised using gamma radiation, and released into the native population. When the sterile insects mate with the wild insects, no offspring are produced. This approach is not only environmentally sound, frequently it is the only practical means of insect eradication. Sometimes the native population of the target insects is first reduced by cultural, biological or attractant/chemical methods before sterile insects are released. Then, when sterile insects are released, the ratio of sterile to native insects is high and the probability of a native insect mating with another native insect is low. If the ratio is high enough in an isolated situation, the sect will be eradicated from that area. SIT is
70
Chapter 4
most effective when the sterile insects can be produced in large numbers, and the native population is low and isolated from other infestations. It is an ideal way of eradicating new infestations of insects before they spread over large areas, but also it is effective in area-wide control of established populations. Further, pest-free zones of agricultural production can be maintained through the use of the SIT. SIT must be undertaken on an area-wide basis for an effective programme. Areawide control of key insects without heavy use of insecticides is often the most economically and ecologically sound approach to pest management. This usually involves an integration of several methods of insect control of which the SIT is often a component. The first successful eradication of an insect using the SIT was the screwworm, a devastating pest of domestic animals and wild life, on the island of Curaqao in 1954. Later the screwworm was eradicated from the USA and then Mexico. Texas ranchers alone estimate that the programme has saved them US$100 million annually. In 1998, the "New World screwworm" was reported in North Africa. This is the first report of this insect becoming established outside the Western Hemisphere. The SIT is the logical technology now being applied to eradicate this new introduction. Much of the fruit produced throughout the world is subject to fruit fly infestation. Fruit flies not only damage the fruit, but prevent countries infested with certain fruit flies from exporting their fruit to countries which do not have these flies. The Mediterranean fruit fly (medfly) has been eradicated from Mexico and the Melon fly from most of Okinawa using the SIT. In addition, several fruit fly introductions have been eradicated from the USA using the SIT. Research is being conducted to reduce the cost of sterile medfly production. IAEA has developed a genetic sexing strain so that only male flies are released. This increases the efficacy of the SIT and avoids "stinging" damage to fruit by sterile females. Tsetse flies transmit a disease causing nagana in cattle and sleeping sickness in man. These insects have prevented settlement and development of large areas of Africa. One species of tsetse fly has been eradicated from parts of Nigeria and three species from parts of Burkina Faso using the SIT. Certain groups of insects, such as moths, are seriously damaged by sterilising dosages of irradiation. Scientists have learned that some of these insects can be irradiated at lower doses which will not completely sterilise the insect, but their progeny will be sterile. This inherited, or F-1 sterility, is an effective way of controlling some insects. Infestations of the gypsy moth have been eradicated in several isolated locations in the USA using this technique. Table 4.3 lists previous and current use of the SIT technology (source: IAEA).
4.2.1.1 Food preservation One of the first priorities in the world is to have enough healthy food for everybody. Great trouble is being taken to fertilise the land, develop suitable mutants of basic crop plants, provide a suitable infrastructure adapted to the country and, generally, create the right circumstances for a good harvest. After that we have to do more to make sure that
Man-made Radioactivit3,
71
Table 4.3 Insect pests and the SIT Insect
Previous use
Current use
Screwworm
Curaqao, USA, Mexico, Puerto Rico, US Virgin Islands Italy(e), Peru(e), Mexico, USA (accidental introductions) Japan(e) Rota, Hawaii (e) Netherlands(e) USA/Mexico(e) Switzerland(e)
Guatemala, Belize, Libya
Mediterranean fruit fly Melon fly Oriental fruit fly Orion fly Mexican fruit fly Cherry fruit fly Other fruit flies Pink Bollworm Codling Moth Gypsy Moth Tsetse flies (4 species) Boll Weevil Sheep Blow fly Mosquitos Stable fly Tobacco hornworm
USA(e) Canada(e), USA(e) USA(e) Tanzania(e), Nigeria(e), Burkina Faso(e) USA(e) Australia(e) E1 Salvador(e) St. Croix St. Croix
Guatemala, USA (accidental introductions) Japan Netherlands (control) USA/Mexico (quarantine) Several countries(e) USA (quarantine) Canada (control) USA Nigeria
USA(e) USA(e)
Note: The table shows insect pests for which the SIT or a related genetic control method is being used, has been used, or is being developed. The objective is eradication unless otherwise noted. An (e) indicates an experimental pilot test.
the preciously grown food is preserved and protected against contamination and pests~an especially important priority for the developing world. For thousands of years this problem has been with us and preservation methods have evolved from the earliest days of sun-drying to salting, smoking, canning, freezing, heating, and the addition of chemicals. The latest addition to this list is irradiation--the exposure of foods to carefully controlled amounts of ionising radiation. Although a relatively new commercial process, food irradiation has been studied more thoroughly than any other food technology. More than 40 years of research have shown conclusively that there are no adverse effects from the consumption of irradiated food. In fact for many foods, preservation by irradiation has proved to be by far the best method. Table 4.4 summarises the general applications of food irradiation technology. All necessary rules and regulations to irradiate certain foods have been adopted by the relevant international authorities, but there is still some public reluctance over the acceptance of such foods. This is surely only temporary and in the future food irradiation will certainly develop to become one of the great benefits for mankind, and
72
Chapter 4
Table 4.4 General applications of food irradiation Purpose
Absorbed dose (kGy)
Products
0.05-0.15 0.15-0.50
Potatoes, onions, garlic, ginger root Cereals and pulses, fresh and dried fruit, dried fish and meat, fresh pork Fresh fruits and vegetables
Low dose (up to 1 kGy)
Inhibition of sprouting Incest disinfestation and parasite disinfection Delay of maturation
0.50-1.0
Medium dose (1-10 kGy)
Extension of shelf life Elimination of spoilage and pathogenic microorganisms Improvement of technological properties of food
1.50-3.0 2.0-7.0 2.0-7.0
Fresh fish, strawberries, etc. Fresh and frozen seafood, poultry and meat Grapes (increased juice yield), dehydrated vegetables (reduced cooking time)
High dose (10-50 kGy)
Decontamination of food additives and ingredients Commercial sterilisation (in combination with mild heat)
0-50 30-50
Spices, enzyme preparations, natural gum, etc. Meat, poultry, seafood, prepared food, hospital diet
food preservation by irradiation will be of the greatest importance to food products grown in many countries. What are the benefits of using irradiation? It can kill viable organisms and specific, non-spore forming, pathogenic microorganisms such as salmonella, or it can interfere with physiological processes; for instance it can be used for sprout inhibition of potatoes or for extending the shelf-life of fresh fruit. In short, irradiation of food is an alternative, and in some cases the only, method to: 9 eliminate many health risks in food; 9 enhance the quality of fresh produce; 9 improve the economy of food production and distribution; 9 reduce losses during storage or transportation; and 9 disinfect stored products such as grain, beans, dried fruit, and dried fish. Economically, one of the most important applications results in the extension of shelf-life, which is of utmost value to countries with warm climates like so many of the developing countries. The same is true for the reduction of losses through storage which are very heavy in some areas: some countries report 4 0 - 5 0 % post-harvest losses through infestation of staple foods like grains and yams. Most stored staple foods therefore are fumigated by chemicals when not irradiated. The present status of the worldwide application of food irradiation is shown in Table 4.5. At an international conference held in Geneva in December 1988 on the "Acceptance, Control of and Trade in Irradiated Food" a document was adopted outlining the
Man-made RadioactiviO,
73
Table 4.5 Examples of worldwide approved uses of irradiated foods and commodities Country
Product
Argentina Bangladesh Belgium Brazil Canada Chile China Cuba Denmark Finland France Hungary India Indonesia Israel Japan Korea, Rep. of Netherlands
Spices, spinach, cocoa powder Potatoes, onions, dried fish, pulses, frozen seafood, frog legs Spices, dehydrated vegetables, deep-frozen foods, including seafood Spices, dehydrated vegetables Spices, potatoes, onions Spices, dehydrated vegetables, onions, potatoes, chicken Potatoes, garlic, apples, spices, onions, Chinese sausage, Chinese wine Potatoes, onions, cocoa beans Spices Spices Spices, vegetable seasonings, poultry (frozen deboned chicken), Spices, onions, wine cork Spices, onions, potatoes Spices, tuber and root crops Spices, potatoes, onions, grains Potatoes Garlic powder, potatoes, onions Spices, frozen products, poultry, dehydrated vegetables, rice, egg powder, packaging materials Spices Potatoes, onions, garlic, spices Potatoes, onions, fruit, spices, meat, fish, chicken, processed products, vegetables Potatoes, onions Potatoes, onions, chicken, fruit, spices Onions, fermented pork sausages, potatoes Potatoes, onions, cereals, fresh and dried fruits and vegetables, meat and meat products, poultry, grains Spices, poultry, fruit Spices, cereals, meat, poultry
Norway Pakistan South Africa Spain Syria Thailand USSR USA Yugoslavia
benefits of food irradiation and recommending harmonisation of national procedures to facilitate international trade in such products. At last, therefore, all practical obstacles seem to have been removed which could hinder the rapid development of this most useful application of radiation to mankind in the very near future.
4.2.2 Medical applications Applications of radiation and radionuclides for human health followed rapidly in the wake of the discovery of X-rays by R6ntgen. Techniques which permitted the production of specific radionuclides in useful quantities were developed. Today, hardly a
74
Chapter 4
single major hospital exists in an industrialised country which does not have a department of radiology and a department of nuclear medicine, or which does not use an extensive array of laboratory radiochemical methods for the diagnosis and investigation of a wide variety of diseases. In nuclear medicine, a radionuclide--in a carefully chosen chemical f o r m J i s administered to the patient to trace a specific physiological phenomenon by means of a special detector, often a gamma camera, placed outside the body. The importance of nuclear medicine, which is now a recognised medical speciality by itself, may be seen from the fact than one out of every three patients attending a major hospital in an industrialised country benefits from some type of nuclear medicine procedure. Such procedures may, like an X-ray, provide us with a picture of some particular body organ or part of it. The essential difference is that in nuclear medicine the picture obtained provides a measure of the activity of some specific physiological or biochemical function in the body. Most nuclear procedures are of a diagnostic nature. In some instances, however, radionuclides administered to the patient are valuable therapeutic tools. For example, one in every three persons admitted to U.S. hospitals undergoes a nuclear medical procedure for diagnosis or therapy. Many of these procedures employ radioisotopes. Some of the more frequent uses of medical radioisotopes include diagnosis and treatment of several major diseases, sterilisation of medical products such as tissue grafts, nutrition research, and biomedical research into cellular processes. Radioisotopes play an important role in the diagnosis and treatment of disease: for example, technetium-99m is used in about 36,000 medical procedures each day in the United States. This radioisotope, which is produced from molybdenum-99, allows physicians to diagnose diseases of the brain, lungs, heart and other organs without exploratory surgery. It is also used in bone scans to identify cancer or stress fractures that cannot be seen in X-rays. Germanium-68 is needed to calibrate positron emission tomography equipment, which is used to diagnose some types of cancer. Yttrium-90 is used to treat non-Hodgkin's lymphoma, a type of cancer, and bismuth-213 is being studied as a potential treatment for a form of leukaemia. Alternative treatments, where they exist, generally require painful, costly, often repetitive surgeries. By reducing the need for such surgery, these and other medical radioisotopes save the public approximately $ 1 2 x 109 per year in the U.S. alone. 4.2.2.1 Radiopharmaceuticals In order to be able to trace a specific biological process in the body, or investigate the functioning of a body organ, it is necessary to make a careful choice of both the radionuclide and the chemical form in which it is administered to the patient. Such radionuclide preparations are called radiopharmaceuticals. Today, some 100-300 radiopharmaceuticals are in routine use for diagnosis, most of which are commercially available. The majority of these compounds are organic in nature (see Table 4.6 for details).
Man-made Radioactivity
75
Table 4.6 Radionuclides in clinical use Radio- Half-life nuclide
Decay process
Principal radiation (MeV) e
Production method
Usage
6Li(n,)3H
Whole body water Biochemical research Physiological research
xory
3H
12.26 yr
e-
0.018
11C 14C
20.3 min 5570 yr
e§ e-
0.97 0.155
0.511 -
l~ JC 14N(n,p)14C
13N
10.0 min
e§
1.20
0.511
12C(d,n)13N
150
2.05 min
J8F 24Na 32p 365
110 min 2.58 yr 14.45 d 87 d
0.511 0.511 1.37 -
42K
310-000 yr 12.5 h
1.74 0.63 1.39 1.71 0.167 0.71
14N(d,n)15O 160(ot,pn) 18F 23Na(n,y) 24Na
36C1
e§ e § EC eeee-, EC e-
43K
22 h
e-
2.0 3.6 0.83
45Ca
165 d 4.53 d
e e
475c
3.43
e-
5~Cr
27.8 d
EC
52Fe
8.3 h 2.7 yr 45d
+ e , EC EC e-
5SCo
267 d 71d
EC + e , EC
0.49
65Zn
245 d
-4e , EC
0.33
67Ga
78h
EC
75Se
120 d
EC
47Ca
+
55Fe 59Fe 57Co
0.25 0.69 2.00 0.44 0.60
0.81 0.27 0.46
Breath tests Physiological research Physiological research
31p(n,7)32 P 35Cl(n,p)35S
Cancer research Exchangeable sodium Therapy of polycythaemia Drug research
35C1(n,7)3 6C1
Physiological research
1.53
41K(n,7)42 K
Exchangeable potassium
0.37 0.61
4~
Exchangeable potassium
K
44Ca(n,7 ) 45Ca 46Ca(n,qt) 47Ca
Calcium kinetics Calcium kinetics
0.322
5~
0.511
52Cr(o~,4n )52Fe 54Fe(n,y) 55Fe 58Fe(n,y) 59Fe
Red cell labelling Glomerular filtration rate Bone marrow imaging Ferrokinetics Ferrokinetics
6~ 7Co 58N(n,p)58Co
Vitamin B12 absorption Vitamin B12 absorption
64Zn(n,7) 65Zn
Physiological research
65Cu(o~,2 n)67Ga
Location of neoplasms and abscesses Imaging of the pancreas and adrenal glands
m
1.31 0.16
0.006 1.10
51Cr
1.29
0.122 0.511 0.81 0.511 1.11 0.18 0.30 0.14 0.27
74Se(n,7)7 5Se
continued
76
Chapter 4
Table 4.6 (continuation) Radio- Half-life nuclide
Decay process
Principal radiation (MeV) e
81Rb
4.5 h
e +, EC
81mKr 77Br 82Br
13.5 s 58 h 35 h
IT e +, EC e-
90y
99Mo
64.4 h 67 h
ee-
99mTc
6h
IT
lln
2.8 d
ll3Sn 113mln 123I
$
+
2.27 0.45
79Br(ot,2n )81Rb
Radionuclide generator Lung function studies
0.190 0.520 0.55 0.62 0.78
m
75As(o~,2n )77Br 81Br(n,qt) 82Br
Extracellular water Extracellular water
89y(n,7)9~ Y
Treatment of arthritic joints Radionuclide generator
0.74
98Mo(n,7 ) 99M0 U(nf) ----)99M0
EC
0.141 0.17 0.25
l~
118 d
EC
0.26
112Sn(n,y )ll3Sn
104 m 13.3 m
IT EC
0.39 0.16
121Sb(,2 n)123I
+
Jl
0.44
Usage
xory 0.511
D
Production method
1.23
2 n) II lln
127I(p,5n ) 123Xe ._.) 123I 125I
60 d
EC
131I
8.1 d
e-
127Xe
36 d
EC
133Xe 137Cs 198Au
5.3 d 30 yr 65 h
eee-
2~
74 h
EC
0.61
0.34 0.51 0.96
0.035
124Xe(n,7 )125Xe 1251
0.36
!3~ )131Te 1311 U(nf)131Te ---) 1311
133Cs(p,2p5n)127Xe
0.17 0.20 0.38 0.081 0.662 0.41
U(nf) ----)133Xe U(t/t/c) ---)137 Cs 197Au(n,7) ~ 198Au
0.07
203Tl(p,3n) 201Pb ---) 201T1
Organ imaging (table 6.1) White cell labelling Imaging of cerebrospinal fluid Radionuclide generator Cardiac output Thyroid studies Renal studies Radioimmunoassay Plasma volume Effective renal plasma flow Deep vein thrombosis Thyroid studies Renal studies Treatment of thyrotoxicosis Treatment of thyroid cancer Lung function studies
Lung function studies Calibration source Treatment of intrapleural or intraperitoneal neoplasms Myocardial imaging
77
Man-made Radioactivity
To minimise the already small radiation dose to the patient through the use of diagnostic radiopharmaceuticals, more and more short-lived--or very short-livedw radioisotopes are being used. These short-lived radioisotopes decay to stable elements within minutes or hours. Radiopharmaceuticals of short-lived isotopes have to be produced at the hospital where they are to be used. This is often done by "milking" the desired isotope from a longer-lived radioactive parent. This is a relatively simple procedure, but it often must be followed by some rapid chemical procedures to convert it into the requisite radiopharmaceutical. This technique is used routinely in hospitals for diagnostic investigations of the functioning of the liver, brain, lung, heart or kidney. Short-lived radionuclides such as indium-111, gallium-67, gallium-68, thallium-201, and the most commonly used technetium-99m, find wide applications. New applications and radiopharmaceuticals are being developed to extend the range of procedures available to doctors. However, it must always be remembered that any in vivo nuclear medicine procedures involve a small radiation dose to the patient. As an illustration, we shall briefly discuss the production of 123I(after Witenboer et al., 1986). For commercial production of iodine- 123, two routes are currently used, viz. the direct reaction 124Te(p,2n)123I and the indirect reaction 127I(p,5n)123Xe ~ 123I. The iodine-123 produced is contaminated with other radioisotopes of iodine, the main contaminant in the first route is 124I, formed by the 124Te(p,n)124I reaction, and in the second route 125I,formed by the 127I(p,3n)~25Xe ---) 125Ireaction. Production of sizeable quantities of iodine-123 of higher purity is possible via proton irradiation of highly enriched xenon-124. The reactions leading to iodine-123 are: 124
Xe(p,2n) 1 23Cs
124Xe(p,pn)123Xe
(5.9 min) --~ 123Xe
(2.1 h) ---) 1231 (Q =-15.5 MeV)
(4.12a)
(Q =-10.3 MeV)
(4.12b)
(Q = -6.8 MeV)
(4.12c)
(2.1 h) -"-) 123I
124Xe(p,2p) 123I Reactions leading to other radioisotopes of iodine are: 124~t r
,,
- 120
Aetp,om)
I
124Xe(p,~)1211
(2.1 h) ---) 121Te
'24Xe(p,3He)122I 124Xe(p,p2n)'22Xe 124Xe(p,qt)125Cs
(20 h) ~ 1221 (45 min) --~ 125Xe (17 h) ~ 125I
(Q = -6.5 MeV)
(4.13a)
(Q = +3.8 MeV)
(4.13b)
(Q = -9.0 MeV)
(4.13c)
(Q =-18.7 MeV)
(4.13d)
(Q = +3.9 MeV)
(4.13e)
The cross section for the interfering 124Xe(p,~) reaction is relatively low, in the energy range of 27-20 MeV, so that only a minute 1251impurity is to be expected. Since
78
Chapter 4
iodine-120, iodine-121 and iodine-122 have short half-lives (1.35 h, 2.1 h, and 3.5 min respectively), no substantial contamination with these radioisotopes will be present at calibration time. Highly enriched xenon-124 is expensive (US $ 150,000 per litre STP) due to its natural abundance of only 0.096%. This calls for an effective gas target and a reliable gas handling system. In the past considerable experience has been obtained in handling enriched krypton-82 gas for the production of rubidium-81 for krypton-81m generators. Since June 1984 this technology is being applied and further improved for the production of iodine-123 via proton bombardment of the enriched xenon-124.
4.2.2.2 Diagnostic methods in cardiology Radionuclides play an important role in cardiological diagnosis. When a doctor examines the pulse of a patient, he is trying to gauge the blood flow, judge the condition of the blood vessel, and indirectly evaluate the force of the pumping action of the heart. A circulating radioactive tracer, like a small spy, can relay the same kind of information from within, such as what volume it occupies after dilution as a blood pool in the heart, and how this volume changes when the heart contracts. With the help of a computer, such information is obtained quantitatively and sequentially in relation to time. Such intelligence forms the heart of nuclear cardiology, one of the most useful applications of modem nuclear medicine. When a patient sees a doctor because of heart trouble, the doctor has many options depending on his suspicions. One rather elaborate way to diagnose is to inject a radiotechnetium compound into the blood stream, followed by an analytical method known as single photon emission computed tomography (SPECT). A rotating gamma camera measures the radioactivity at short intervals providing, with the help of a computer, a reconstructed picture, which enables the physician to determine how much of the heart muscle is deprived of blood. If the blood flow to the heart, as well as the metabolism of the muscle, are to be assessed, then another new method can be very useful. The positrons emitted from some radionuclides which have been incorporated in organic compounds are measured by positron emission tomography (PET). The positrons are produced when certain short-lived isotopes decay and, through interaction, produce very strong gamma rays (511 keV) which go off in almost exactly opposite directions. These can be detected easily by a special device using detectors placed on opposite sides of the patient. During the last few years, a much smaller and more sensitive detector has been developed which will make this method even more useful in the future. As a result of such measurements, one can show the distribution of the tracers, or rather the compounds containing these tracers, indicating how metabolically active these tissues are. Molybdenum-99 is a radioactive isotope that decays to form technetium-99m, an isotope used in about 36,000 medical procedures each day in the United States. Technetium-99m allows physicians to diagnose many conditions in the brain, lungs,
Man-made Radioactivi~
79
heart, and other organs without the use of dangerous and expensive exploratory surgery. For example, technetium-99m imaging is used to diagnose poor blood flow in the lungs and heart. Alternative diagnostic methods include an arteriogram, a procedure in which an imaging device is inserted into a large vein, and cardiac catheterization, which requires inserting a tube into the heart. These alternative methods cause the patient some discomfort and require a recovery period. Because technetium-99m imaging is not a surgical procedure, costs for surgical facilities and personnel, as well as medication to ease pain and promote healing, can be avoided. Technetium-99m is also used in bone scans to identify the spread of cancer to the skeletal system or to detect stress fractures that cannot be seen in X-rays. There are many other usable positron emitters, like rubidium-82, which are used to measure the blood flow to the heart muscle. There are other techniques as well, some using non-radioactive compounds by making use of the known X-ray computed tomography method. More recently, even magnetic resonance imaging methods are being applied for certain diagnostic work. Ultrasound techniques also are being tried for certain heart assessments. These examples illustrate that with sophisticated radiation-emitting methods, it is possible to make diagnoses which would have been impossible not so long ago. Roughly three percent of the population of Europe, some six million people, suffer from coronary artery disease. A routine procedure could involve many of them in tomographic tests using a radiopharmaceutical. Nuclear imaging is used more and more widely, such as for brain disease diagnosis. Cerebrovascular diseases occur at roughly the same rate as cardiac troubles. In these cases, organic radiochemicals are labelled with fluorine, oxygen, nitrogen or carbon radionuclides for imaging. Tumours may be located with similar methods, using either simple radiopharmaceuticals or complex radionuclide-labelled antibodies. As an example, let us discuss in some detail thallium-201 which is used as an agent for myocardial imaging studies. There is an ever-increasing use of thallium-201 (Z~ in nuclear medicine in the last two decades (Pennel et al., 1992). The 2~ given intravenously as thallous chloride is used in myocardial perfusion scintigraphy, because its rapid clearance from the circulating blood into the myocardial tissue reflects, reasonably well, the myocardial perfusion. Myocardial perfusion scintigraphy has gained worldwide acceptance as a non-invasive approach to the evaluation of patients with suspected coronary heart disease (Steien and Aaseth, 1995). The isotope, 2~~ is a cyclotron-produced radioactive compound, decaying to mercury-201 (2mHg) with a physical half-life of 73 h, the decay being accompanied by emission of gammaphotons of 135 and 167 keV, but the main emission is X-rays of 67-82 keV. After intravenous administration of a tracer dose of thallous-201 chloride, the cation disappears quickly from the circulation, with a biological half-life in the blood of less than 1 min, as the 2~ is rapidly taken up by different tissues, especially heart and skeletal muscle (Kazantzis, 1986; Pennal et al., 1992). In apparently healthy individuals subjected to standardised physical exercise on a bicycle ergometer before intravenous administration of 2~ chloride (80 MBq), it was found that 3.9% of the
Chapter 4
80 Table 4.7
Thallium-201 activity in various organs, given as percentage body burden (mean and range), at different time intervals after an intravenous injection of 80 MBq [2~ thallous chloride (after Steien and Aaseth, 1995) Organ
Heart Brain Thyroid Liver Kidneys Lower extremities
Time interval after 2~
injection
30s
4h
24h
3.9 (3.6--4.1) 1.4 (1.0-1.7) 0.8 (0.7-1.0) 3.5 (2.9-4.3) 6.5 (5.2-9.0) 42 (40-46)
2.2 (2.0-2.4) 1.4 (1.1-1.7) 1.2 (0.8-1.3) 3.4 (2.5-4.4) 6.0 (4.8-8.3) 38 (35-42)
1.8 (1.3-2.3) 2.1 (1.8-2.9) 1.1 (0.9-1.6) 3.6 (2.5-5.1) 6.0 (4.6-8.1) 31 (28-33)
dose was rapidly taken up by the heart (Table 4.7). The washout rate from the heart was relatively low, with 2.2% of the body burden being retained after 4 h and 1.8% after 24 h. In the thyroid gland, as in the brain, the uptake was rather small, and the 2~ deposits in these tissues were not subjected to apparent washout/redistribution during the observation period (Table 4.7). The lower extremities with their considerable amount of muscles showed the highest 2~ uptake (42%), and a significant washout was observed during the subsequent 24 h period. The hydrated thallous ion is similar in size to the hydrated potassium ion, and early literature reported that the uptake of T1 cations in muscle cells made use of the specific uptake mechanism developed for potassium. However, later studies, taking account of the complexity of potassium transport, and the different types of potassium channels, have found some differences between the cellular TI uptake and the potassium uptake. Thus, digoxin that inhibits the Na/K ATP-ase enzyme system as well as the potassium ion-transport, did not affect the 2~ transport. Furthermore, once inside myocardial or other cells, 2~ shows a low washout rate compared with potassium, probably owing to its interactions with intracellular constituents. The crucial physiological factor that interferes with the 2~ uptake by heart muscle cells, in vivo, is local hypoxia. Thus, hypoxia induced by physical stress in patients with coronary heart disease, can be scintigraphically visualised at 2~ uptake defects, provided that the imaging is performed soon after the isotope injection. After a redistribution period of 3-4 h, the 2~ uptake is accomplished even in poorly perfused muscle cells, indicating that the 2~ distribution after a 3 h equilibration period will reflect the viable mass of the myocardium. 4.2.2.3 Radionuclides in the treatment of disease
There are relatively few situations in which the administration of a radiopharmaceutical to the patient can be used for treatment of disease. The oldest and best known of these applications is the treatment of overactivity of the thyroid gland and of some types of
Man-made Radioactivi~'
81
thyroid cancer, by giving the patient a carefully calculated amount of iodine- 131. Other examples are the use of strontium-89 to palliate pain provoked by bone metastases of prostatic, mammary and possibly other carcinomas; or the treatment of phaeochromocytoma and other tumours of the cromoffin tissue with iodine-131 labelled metaiodobenzyl-guanidine. Much hope for the future lies in the development of tumour-specific antibodies which could be used to target radionuclides to tumours and thereby destroy them. Teletherapy is radiation treatment where the radiation source is not in direct contact with the tumour to be treated. The radiation used for the treatment can be of different types and energies and originate from different sources. Gamma-emitting radioactive sources such as cobalt-60 are often used, because they are convenient, need virtually no maintenance, and are almost ideal gamma emitters. Many of these sources are in use for cancer treatment. Brachytherapy is a treatment where the radiation source is in direct contact with the tumour. This method is used widely for a number of special medical cases. As cancer of the cervix is quite a common disease in many developing countries, brachiotherapy has become the method of choice for treatment because many patients can be treated relatively cheaply and effectively. One of the first big projects of this kind in a developing country was organised in Egypt with the co-operation of the World Health Organisation (WHO) and the IAEA. This method, however, is only applicable when the tumour has not spread more than a few centimetres. Fortunately, this is the case with many patients. Should the tumour be larger however, the more costly teletherapy must be applied. The usefulness of brachytherapy for cancer treatment can be assessed when one realises that roughly one quarter of all cancer cases in countries like Nigeria are suitable for such treatment. With a relatively inexpensive and uncomplicated application of radiobrachytherapy, one can not only treat but, in especially early cases, also cure many patients. In the last two decades, scientists have developed homing materials (monoclonal antibodies, peptides) that attach themselves to various types of cancer cells. Methods of linking radioactive isotopes to these homing materials have also been discovered, resulting in so-called "smart bullets" that can be delivered directly to the locations of cancer cells. Scientists have demonstrated this procedure by attaching yttrium-90, a beta-emitter, to a monoclonal antibody as a potential treatment for non-Hodgkin's lymphoma, a type of cancer, with very positive results. Researchers are now assessing yttrium-90 for use in treating many types of cancers. Together, these cancers are diagnosed in an estimated 210,000 people each year in the U.S. alone. Since alpha particles have a smaller range than beta particles, by using them the radiation is delivered to cancer cells without damaging surrounding healthy tissue. The successful use of alpha-emitters for cancer therapy depends on the identification of a homing material for each type of cancer to be treated. Chemical processes that can attach an alpha-emitter to the desired homing material must also be found. If this therapy proves successful, specific alpha-emitters must be produced at a large rate that allows for full-scale treatment of affected populations.
82
Chapter 4
Three alpha-emitters, bismuth-213, astatine-211, and radium-223, have been shown to have the properties needed for cancer therapy. All three have been successfully linked to a homing material. Pre-clinical trial results have been promising. Clinical human trials for treatment of a type of leukaemia began at New York City's Memorial Sloan Ketting Cancer Center in October 1996. The University of California at Los Angeles is also studying bismuth-213 for lung cancer therapy. The National Cancer Institute is conducting studies to determine the value of this therapy in treating brain cancer. Pre-clinical trials Using astatine-211 for brain cancer therapy have been initiated at Duke University Medical Center. Studies using radium-223 are under way at Pacific Northwest National Laboratory, Idaho State University, Washington State University, the New Jersey School of Medicine, and UCLA. If the use of alpha-emitters for cancer therapy proves successful, it has been estimated that some 30-50,000 cancer patients could be treated in the United States each year.
4.2.3 Industrial applications Many beneficial applications of radiation and radioisotopes in industry are well established. Use of radioisotopes and radiation in modern industry is of great importance for process development and improvement, measurement and automation, and quality control. Today, almost every branch of industry uses radioisotopes and radiation in some form. The use of radioisotope thickness gauges is a prerequisite for the complete automation of high speed production lines such as for steel-plate or paper. Tracer experiments give exact information on the condition of expensive processing equipment and increase its usable life. The use of isotopes has grown rapidly in virtually all industries. For dams, aircraft, bridges, and piping, isotope use has become critical to ensuring structural integrity. As an example, let us mention that radioisotopes are the only tool available today for scanning the interior structure of a jet engine or an oil pipeline to detect flaws prior to failure. Several radioisotopes are used to ensure safety in industry and transportation. For example, iridium-192 is used to verify the structural integrity of aircraft, ships, bridges, and other structures, for weld inspection, and other purposes. Californium-252 is used to gauge the moisture content of soil in road construction and the building industries. Various isotope applications are used to monitor the quality of materials and structures. Isotopic tracer techniques measure wear, corrosion, moisture, leakage, and many other factors. Neutron radiography creates images of materials that are not as dense as those captured in X-ray photos. This method is used chiefly to check uranium fuel in nuclear reactors for flaws, to find cracks in the inner plastic or aluminium parts of airplanes, or to detect tiny fissures in gas turbine blades. Californium-252 is used for neutron radiography and neutron activation analyses. Some of the more common industrial uses of isotopes to ensure safety include" 9 wear and corrosion analysis;
Man-made Radioactivi~'
9 9 9 9
83
leak, flaw, and malfunction investigations; elimination of static electricity; light sources for space and other remote locations and emergency lighting; and smoke detectors.
4.2.3.1 R a d i o i s o t o p e s as tracers
The fact that minute amounts of radioactive substance can be measured readily and precisely makes radioisotopes an important tool for investigations in which transport of material is involved and exact information about spatial and temporal distribution of the material is required. A wide range of different industries use tracer techniques including: 9 coal, 9 oil, gas and petrochemical; 9 cement, glass, building materials; 9 ore processing; 9 pulp and paper, iron and steel; 9 non-ferrous metals; and 9 automotive. The main areas where radioisotope tracers may be used are: 9 p r o c e s s i n v e s t i g a t i o n s - - r e s i d e n c e time, flow rate, velocity, modelling, parameter estimation; 9 m i x i n g m m i x i n g time, mixer optimisation, mixer performance; m a i n t e n a n c e - - l e a k detection, investigation of malfunctions, material transport; 9 w e a r a n d c o r r o s i o n m e n g i n e wear corrosion of process equipment, lubrication studies. 9 In the processing industries, one of the major applications of radioisotope tracers is for residence time investigations in which important parameters for plant optimisation, modelling, and automation are obtained. Once optimum performance of the plant has been reached, tracer experiments may be carried out to indicate deviations from optimum conditions. Often the reasons for malfunction are found, like unwanted by-pass streams, or obstruction of vessels and pipes which can cause changes in flow-rate or the appearance of dead zones. Often the necessity for a shut-down can be tested and vital information for repair work to be done can be obtained prior to shut-down. Typical examples are reported from the petrochemical industry for the optimisation of fractionating columns. Mixing is a very important step in some processes. It consumes time and energy and expensive equipment is necessary. Optimisation of mixing processes, therefore, is an important goal that can be reached by the application of tracers. The study of wear on machine parts, which were labelled by radioisotopes, is an important stage in the development work of the automotive industry. The design of a new motor necessitates hundreds of wear tests to be carried out. These tests can be made by using the radioisotope tracer technique. The surface activation technique, in 9
84
Chapter 4
which only a thin layer of the part under investigation is activated by bombardment with ions from an accelerator, guarantees extremely high sensitivity and uses only small amounts of radioactive material. Impressive figures are available concerning savings in the automotive industry due to the use of radioisotope tracers for wear studies. Reports say that in the development of a new engine the costs for testing a new cylinder liner amount to about US $360 000 for each liner when using conventional wear measuring methods. By using radioisotope tracer techniques, the costs are cut below US 50 000. For a series of measurements on 10 linear modifications, which are usually made during the development process, the savings made by applying radioisotope techniques would be around US$ 3,100,000. Similarly, the savings can be calculated for tests on bearing cups. For a series of tests on 20 beating-cup modifications, the costs amount to US$ 3,500,000. When applying radio-isotope techniques, the same results can be obtained for only US $ 400 000 resulting in a saving of US $ 3,100,000 (after IAEA Report). In addition to savings, there are further technical advantages of great importance. When using radioisotopes, the entire test can be run without dismantling the engine which allows more accurate results to be obtained. A very important factor in development is time. The results from the test series using radioisotopes are usually available within six months; the conventional tests may take up to five years. In general, tracer techniques are used throughout industry to improve the efficiency of the processes, to save time, energy and raw material, to reduce down-time of equipment, and to facilitate development work.
4.2.3.2 Radioisotope instruments The greatest impact of radioisotopes in industry has resulted from the use of radioisotope instruments. Due to the nature of the ionising radiation emitted from radioisotopes, a few unique advantages are provided with this technique: 9 Because radiation has the ability to penetrate matter, measurements can be made without direct physical contact of the sensor with the material being measured. 9 On-line measurements on moving material can be made; measurement is nondestructive. 9 The stability of the source is excellent and little maintenance is required. 9 Excellent cost/benefit ratios can be achieved. Radioisotope instruments became available for all kinds of measurements just when the trend towards automation in industry was strong. Radioisotope instruments can perform certain measurements such as mass per unit area which cannot be made by other equipment. For other measurements, like level or distance, there are now other competing methods available. Radioisotope gauges for measuring mass per unit area (sometimes also called "thickness gauges") are unequalled in their performance and are used in almost every kind of industry in which sheet material is produced. In the paper industry, not only the mass per unit area of the paper sheet itself is measured by radioisotope gauges, the
Man-made Radioactivity
85
production of the felt, which is used to support the still very wet pulp in the first stages of paper production, relies heavily on the use of radioisotope gauges to guarantee its extreme uniformity, as well. The latter is of vital importance for the paper machines operating at high speed. Similarly, the production of steel plate at the speed of modern rolling mills could not be done without accurate measurement of thickness at every moment of the production and automatic control of the rolling stands. In the plastics industry, radioisotope gauges are used to improve the uniformity of the product, and hence savings can be made in raw material and in energy needed for production. Microprocessor technology had a great impact on the development of radioisotope instruments. Linearisation of complex calibration curves, compensation for the decay of the radioisotope, and performance of important calibration checks can be handled easily by the microprocessor. In this way, radioisotope instruments of modern design added yet another dimension of reliability and sophistication to their proven excellence. Density gauges based on the absorption of gamma radiation are used wherever the automatic determination and control of the density of liquids, solids, or slurries is important. The oil industry relies heavily on such instruments. Other applications are in the handling of slurries in mineral processing or even in the food industry. One of the earliest users of radioisotope instruments was the tobacco industry, where density gauges make sure the right amount of tobacco is packed into each cigarette. The coal industry has benefited greatly through the applications of nuclear techniques. Nucleonic gauges and on-stream analysers are now regularly employed for monitoring and controlling the ash and moisture content in coal and coke. Nuclear techniques make possible the on-line determinations of sulphur and nitrogen (the causes for acid rain) in coal; both of these are important for pollution control. Hundreds of millions of tonnes of coal are analysed annually by this method, a process which has become routine in the coal industry. Radiation from radioisotope sources can be used to excite characteristic X-rays in samples upon which the beam of radiation is directed. Detection and analysis of these X-rays yield information about the composition of the sample. This opens the field of analytical applications of X-ray fluorescence analysis. The most frequent applications are in the ore processing and the metal coating industries. In ore processing, a sample stream of the slurry of ground ore is fed to a measuring head containing the radioisotope source and the X-ray detector. The exact composition of the slurry can be determined and the operation of the plant controlled to give optimum performance. Great savings can be achieved by better utilisation of raw ore, energy, and chemicals used for the process. Although the cost of such an instrument, including installation, is high it can be recovered usually within one year of operation. In metal coating, such as galvanising or tin-coating of steel plate, the exact amount of coating must be applied. A surplus of material is extremely expensive; undercoating results in complaints and early corrosion. Through the use of radioisotope gauges,
86
Chapter 4
coating processes can be controlled to meet tight limits and thus up to 10% of material (zinc, tin) can be saved. At the same time, the reject rate due to undercoated strip is reduced. In the production of sheets and plates cut to a certain length, special steps are taken to measure the exact length of material when it passes the gauge. Digital counting techniques are preferred for this type of measurement because the timing of the measurement can be made to fit exactly the desired stretch of material. Level measurements can be made by installing a source and a detector on opposite sides of a tank or silo. When filled, the material absorbs the radiation otherwise sensed by the detector. This technique is most useful where circumstances such as pressure, heat, or the presence of toxic, corrosive, or abrasive substances make access to the tank and installation of conventional gauges difficult or impossible. Level gauging using movable source detector combinations is a useful tool for the inspection of process equipment such as chemical reactors. Checking catalyst levels in chemical reactors or monitoring the operation of large fractionating columns in refineries are two applications widely used. Again, savings can amount to impressive figures if one considers that down-time costs due to production losses of a distillation column in a petroleum refinery can be in the order of US$ 300 000 per day (IAEA Report). Yet another extremely useful application of radioisotopes which can save considerable costs and prevent severe damage is in quality control during the construction of pre-stressed concrete bridges. The strength of these bridges is based on bracing cables which run through encasing tubes in the bottom section of the bridge girder. If the bracing cables do not lie in a straight line, considerable damage to the building may result when the necessary stress is applied to the cables. Parts of the concrete slab may be caused to fly off due to the unexpected forces, representing not only a severe hazard, but also necessitating a complete reworking of the structure. A radioisotope source, which is inserted into the encasing tubes and pulled through before the bracing cables are pulled in, is used to determine the exact position of the tubes. If any deviation from their target position is observed, corrective measures may be taken before damage to the building occurs. Such deviations can originate when the encasing tubes are detached from their fastenings by the force of the concrete cast into the sheathing. Neutron moisture gauges are especially well suited for measuring moisture in bulk material such as sand. Their use in the production of glass and concrete continues to grow. Portable instruments are indispensable for checking thicknesses of bituminous material in the construction of roads and dams. A gamma density measurement completes the important information about the quality of the construction. A novel, routine use of neutron sources is in the rapid detection of hidden explosives. Instruments have been developed that can detect small amounts of explosives by measuring gamma rays emitted when neutrons are captured by nitrogen atoms which are present in explosives. Nuclear techniques such as nuclear bore-hole logging and radiometric in-situ analysis play an increasingly important role in the exploration for oil, gas, and metalliferous minerals.
Man-made Radioactivity
87
Radiography using x- or gamma-rays is well established and is a routinely used technique of non-destructive quality control. It is applied for checking welds, castings, assembled machinery (such as jet engines), and in ceramics. Radioisotopes as a source of radiation offer the advantage that they do not require electrical power so that they can be used readily in the field. Different sources are available as well, ranging from low to high energy. The small size of radioisotope sources allows inspection of parts or machinery which could not be examined by X-ray tubes. The most frequent application of gamma radiography is checking the welds in pipelines. This is done most conveniently by putting the source inside the centre of the pipe and attaching the film to the outside of the weld. For checking long pipelines, sophisticated, self-propelled crawlers which travel in the pipe are used. These devices can be positioned exactly at the desired position from the outside. At a command the exposure is made. Then the crawler is instructed to move on to the next weld. Practically all new gas- or oil-pipeline systems are checked with this type of equipment. 4.2.3.3 Radiation in manufacturing Radiation can induce certain desired chemical reactions. It can, for example, be used in the making of plastic, or to graft plastic to other materials. Some polymers whose cross-linkage is induced by radiation can be tailored to shrink when heated--a desirable property in some packaging applications. The wood and printing industries make extensive use of electron-beam radiation to cure surface coatings. The rate of production of wire and cable insulated with radiation cross-linked polyvinylchloride is increasing steadily. Such insulation has better resistance to heat and chemical attack and increased cut-through resistance, and is more compact. The products are used in the automobile industry, telecommunications, the aerospace industry, and in home electrical appliances. Other important products include radiation cross-linked foamed polyethylene which is used for thermal insulation, floor mats, crash padding, floating jackets, and wood/plastic composites cured by gamma irradiation. These have been used successfully for flooring in places such as department stores, airports, hotels, and churches where their excellent abrasion resistance, the beauty of the natural grain, and low maintenance costs are important. This latter technique is also being used in the conservation of objects made of stone and wood of interest to our cultural heritage. The vulcanisation of rubber sheet by radiation--instead of using sulphur in the manufacture of tyres--is being used commercially by several tyre companies. A "super-absorbent" material manufactured by radiation grafting techniques has come onto the market recently. The material is capable of absorbing and holding large amounts of liquid. Products manufactured from it include disposable diapers, tampons, and air-freshener elements. Radiation is beginning to be used to decompose septic or poisonous waste. Some cities irradiate human waste products. Radiation replaces the otherwise necessary addition of chemicals such as chlorine, itself a poison.
88
Chapter 4
Radiation processing has great potential in a new area of application known as radiation immobilisation of bioactive materials such as drugs, enzymes, antigens, and antibodies on polymeric materials. Such immobilisation assures better stability and longer shelf-life for the sensitive biological molecules and offers the possibility of producing slow and sustained drug delivery systems for prolonged controlled therapy of many diseases.
4.3 MANUFACTURING OF RADIOISOTOPES Radioisotopes, as well as stable isotopes, can be produced by accelerators, mainly cyclotrons, reactors or by devices constructed for isotope separation. In this chapter we shall mention some of the most important locations in several countries which are in the business of manufacturing and selling isotopes. 4.3.1 U.S.A. Although there are several private isotope production facilities we shall mention here only isotope production and distribution carried out by the U.S. Department of Energy (DOE). The U.S. Department of Energy's national laboratories offer unique isotope production and separation facilities and processes, such as reactors, associated hot cells, accelerators, and calutrons. The location of these laboratories is shown in Fig. 4.2.
Fig. 4.2. United States Department of Energy isotope origins.
Man-made RadioactiviO,
89
The production, acquisition and distribution of isotopes, and performance of related services, continue long-standing activities conducted by the United States Department of Energy and its predecessor agencies. Materials in inventory or produced in nuclear reactors, charged particle accelerators and separated stable isotopes, DoE offers for sale. The isotopes are mostly in intermediate forms suitable for incorporation in diverse pharmaceuticals, generator kits, irradiation targets, radiation sources, or other finished products. The 85-megawatt High Flux Isotope Reactor at Oak Ridge National Laboratory provides the world' s highest steady-state neutron fluxes. The neutron currents from the four horizontal beam tubes are also very high. The reactor operates about 43 weeks per year, and is used primarily to produce transuranic isotopes. Built-in experimental irradiation facilities also provide versatility, significant experimental capabilities, and the capability of producing a wide variety of isotopes. Products produced at this facility include californium-252, used primarily for cancer therapy, and iridium-192, used for industrial radiography. Sandia National Laboratories' Annular Core Research Reactor (ACRR) is a 2 megawatt, pool-type research reactor that is used to produce isotopes for medical applications. The ACRR and Sandia's nearby hot cell facility, along with Los Alamos National Laboratory' s (LANL) chemistry and Metallurgy Research Facility, have been chosen for US domestic production of molybdenum-99 and related medical isotopes. In addition, the US DoE processes byproducts from nuclear operations to obtain isotopes. For example, Pacific Northwest National Laboratory obtains yttrium-90 from strontium-90, a waste product. Researchers throughout the U.S. are now assessing the effectiveness of yttrium-90 in treating prostate and many other types of cancers. The Isotope Production Facility at LANL operates about 22 weeks per year. This accelerator facility produces radioisotopes using either the primary proton beam or neutrons from the beam stop of the Los Alamos Neutron Science Center (LANSCE), a halfmile-long accelerator that delivers medium energy protons. The unique characteristics of the LANSCE accelerator include a high energy, high beam current that allows production of higher quality radioisotopes, as well as exotic radioisotopes that cannot be produced in other facilities. Three major products produced at the site are germanium-68, a calibration source for positron emission tomography (PET) scanners; strontium-82, the parent of rubidium-82, used in cardiac PET imaging; and sodium-22, a positron-emitter used in neurologic research. The Brookhaven Linear Isotope Producer (BLIP) at Brookhaven National Laboratory uses a linear accelerator that injects 200 MeV protons into the 33 GeV Alternating Gradient Synchrotron. The BLIP facility operates about 16 weeks per year and produces radioisotopes such as strontium-82, germanium-68, copper-67, and others that are used in medical diagnostic applications. The electromagnetic calutrons at Oak Ridge National Laboratory separate isotopes with the same atomic number, but different mass, to produce enriched stable isotopes. During this process, mixed isotope material is vaporised (heated) and then ionised. The ionised particles are accelerated, and their trajectories are bent by a magnetic field. The
90
Chapter 4
Fig. 4.3. The U.S. Departmentof Energy isotope sales by production category. lighter particles separate from the heavier particles as they travel in an arc and are deposited on collectors, from which they are removed, chemically purified, and stored. The Oak Ridge National Laboratory's calutrons operate as needed to maintain an appropriate inventory of enriched stable isotopes. Many of these isotopes, such as strontium-88, thallium-203, and zinc-68, are required to produce other isotopes used to help diagnose cancer and heart disease and provide cancer therapy. Only one bank of calutrons is expected to operate during fiscal years 1998 through 2000. The contact point for information in this field is: 9 U.S. Department of Energy Office of Isotope Production and Distribution, Room A430 GTN, Washington, DC 20585, USA. Telephone: (301)903-5161; Fax: (301)903-5434; Telex: (710)828-0475. Two separate Government operations which supply products are: 9 The New Brunswick Laboratory Reference Materials Sales, Bldg. 350, 9800 South Cass Avenue, Argonne, IL 60439, USA. Telephone: (708) 972-2767; Fax: (708) 972-6252. 9 National Institute of Standards and Technology Radioactivity Standards Reference, Sales Office, Bldg. 245, Rm C114, Gaithersburg, MD 20899, USA. Telephone: (301) 975-5531; Fax: (301) 926-7416 It might be of interest to present a complete list of isotopes produced by DoE facilities; this is shown in Table 4.8. It is also interesting to see the distribution of isotope sales by production category: around 60% of radioisotopes are for medical use (as shown in Fig. 4.3.). Finally, let us mention that a variety of anodised, electroplated, deposited and polysurface disc and large area planar alpha and beta standards are available from: 9 Isotope Products Laboratories, 1800 N. Keystone Street, Burbank, CA 91504, USA. Telephone: (818) 843-7000 Fax: (818) 843-6168.
Man-made Radioactivio,
91
Table 4.8 Isotopes produced by DoE facilities Element
Radioisotope
Actinium-227
Ac-227
Aluminium
A1-26
Americium
Am-241 Am-243
Stable isotope
Natural target
Sb-121 and -123
Antimony
Ar-36 to 40
Argon Arsenic
As-72 As-73 As-74
Astatine
At-211
Barium
Ba-133g -133m
Berkelium
Bk-249
Beryllium
Be-7
Bismuth
Bi-205 -206 -207
Ba-130 to -138
Natural target
Natural target
Boron Br-76 -77 -80m
Br-79 and -81
Cadmium
Cd-109
Cd-106 to -116
Caldium
Ca-45
Ca-40 to -48
Californium
Cf-249 Cf-252
Carbon
C-14
C-12 and C-13 Natural target
Cerium
Ce-141
Ce-136 to-142
Caesium
Cs-137
Natural target
Chlorine
C1-36
C1-35 and-37
Chromium
Cr-51
Cr-50 to -54
Cobalt
Co-60
Natural target
Copper
Cu-64 Cu-67
Cu-63 and-65
Bromine
tp
pp
C1-35 and-37
Chlorine
continued
Chapter 4
92
Table 4.8 (continuation) Element
Radioisotope
Curium
Cm-244 Cm-248
Dyprosium
Dy-165
Stable isotope
Dy-156 to-164
Erbium
Er-162 to -170
Europium
Eu-151 and -153
Fluorine
F-18
Natural target
Gadolinium
Gd-153
Gd- 152 to - 160
Gallium
Ga-67
Ga-69 and -71
Germanium
Ge-68
Ge-70 to -76
Hf-172
Hf- 174 to - 180
Hydrogen
Deuterium
Tritium
Indium
In-114m
In-113 and -115
Iodine
1-124 1-125
Gold Hafnium
Natural target
Helium-3
He3-Rg & Pg, He3995
1-129 Iridium
Ir-192
Ir-191 and -193
Iron
Fe-52
Fe-54 to-58
Fe-55
Fe-59 Krypton
Kr-85P Kr-85E
Lanthanum Lead
Kr-78 to-86 La- 138 and - 139
Pb-203
Lithium
Pb-204 to -208 Li-6 and-7 Natural target
Lutetium
Lu- 175 and - 176
Magnesium
Mg-28
Mg-24 t o - 2 6
Manganese
Mn-54
Natural target
Mercury
Hg-203
Hg- 196 to -204
Molybdenum
Mo-92 to - 100
Neodymium
Nd-142 to-150
Neon
Ne-20 to -22
continued
Man-made Radioactivi~.
93
Table 4.8 (continuation)
Element
Radioisotope
Neptunium
Np-236 Np-237
Nickel
Ni-63
Ni-58 t o - 6 4
Niobium
Nb-95
Natural target
Nitrogen Osmium
Stable isotope
N-14 a n d - 1 5 Os- 194
Oxygen
Os- 184 to - 192 O-16 to -18
Palladium
Pd-103
Phosphorus
P-33
Platinum
Pt-195
Plutonium
Pu-237 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242
Polonium
Po-210
Pd-102 to -110 Pt- 190 to-- 198
Potassium
K-39 to -41
Praseodymium
Natural target
Promethium
Pm-147
Radium
Ra-224
Rhenium
Re-186
Re- 185 and - 187
Re-188 Rubidium
Rb-83
Rb-85 a n d - 8 7
Ruthenium
Ru-97
Ru-96 to - 104
Samarium
Sm-145 Sm-153
Sm-144 t o - 1 5 4
Scandium
Sc-47
Natural target
Selenium
Se-72 Se-75
Se-74 t o - 8 2
Silicon
Si-32
Si-28 t o - 3 0
Na-22M Na-22S
Natural target
Silver Sodium
Ag-107 a n d - 1 0 9
continued
Chapter 4
94
Table 4.8 (continuation)
Element
Radioisotope
Stable isotope
Strontium
Sr-82
Sr-84 to-88
Sr-85 Sr-89 Sr-90 Sulfur
S-35
S-34 and-36 S-32 t o - 3 6 Ta- 180 and - 181
Tantalum
Ta-182
Technetium
Tc-95 Tc-95m Tc-96 Tc-99
Tellurium
Te-127
Terbium
Te- 120 to - 130 Natural target
Thallium
T1-204
Thorium
Th-229
Thulium
Tm-170
Natural target
Tin
Sn-117m Sn-119m
Sn-112 to -124
Titanium
Ti-44
Ti-46 t o - 5 0
Tungsten
W-188
W- 180 to
Uranium
U-233 U-234 U-235 U-236 U-238
Fissile target
TI-203 and-205
- 186
Vanadium
V-48 V-49
V-50 to -51
Xenon
Xe-127
Xe- 124 to - 136
Xe-133 Ytterbium
Yb- 169
Yttrium
Y-88 Y-90
Zinc
Zn-62 Zn-65
Zn-64 t o - 7 0
Zirconium
Zr-88
Zr-90 to -96
Yb-168 to-176
Man-made Radioactivity
95
4.3.2 France
The Bureau National de M6trologie (BNM) has designated The "Laboratoire de Messure des Rayonnements Ionisants" (LMRI) as "Approved Calibration Center". It is a laboratory of the "Commisariat a l'Energie Atomique (CEA)" implanted in the Nuclear Research Center of Saclay. It belongs to the "D6partement des Applications et de la MEtrologie des Rayonnements Ionisants" (DAMRI) whose laboratories are specialised in radioactivity for research and industrial applications of radionuclides. In addition, the LMRI elaborates and distributes radioactivity standards and references, and provides calibrations, measurements and testings in radioactivity and dosimetry, for measuring instruments and ionising radiation sources. These services are intended for research, industry and medicine. The certified values of the produced standards are linked to the national standards by the Primary Laboratory of the "Bureau National de MEtrologie". This traceability is achieved through the utilisation of measuring instruments which are periodically calibrated standards provided by the Primary Laboratory. At international level, traceability is established for a certain number of radionuclides, with national laboratories with which the Primary Laboratory performs direct comparisons or indirect comparisons through the International Reference System of the "Bureau International des Poids et Mesures". An official calibration certificate containing all necessary information is provided with each standard. Because of some physical or chemical phenomena, such as, for example, adsorption in wall containers, the quality of a standard deteriorates with time. In addition, due to the uncertainty about the half-life, it is recommended not to use a standard beyond a certain space of time. This time is twice as long as the half life for radionuclides with "short half life" (T~/2 < 1 year) and 1 year for those with "long half life" (T1/2 -> 1 year). According to official regulations, solid standard sources are generally submitted to classical tests of non-contamination by wiping or by immersion, as required. However, in view of the meteorological quality required, some standards being brittle and of low level activity, these tests are not performed in order to avoid any alteration of the standards or of their accuracies. On the other hand, the sealed sources are submitted to strict tests. The radionuclides which can be fabricated together with their characteristics (lifetime energies of emitting radiation) are shown in Table 4.9. Radioactive standard solutions are usually supplied in sealed glass ampoules. However, for high activities, standards are supplied in capped glass vials for easier handling. For safety, large volume standard solutions for environmental survey are delivered in plastic vials. Multigamma standard solutions differ from other solutions by the fact that they are characterised in terms of photon emission flux per unit mass in 4rt sr (expressed in s-~ g-~). The energies of gamma-rays given as reference are also certified. For calibration, either in energy or in efficiency of NaI(T1) or Ge(Li) detectors, the following compositions are proposed:
96
Chapter 4
Table 4.9 Radionuclides and radiation energies (MeV)
Radionuclide
T1/2
1l~
2.50X102 days
l l0Ag
(x
~max
it
2.235 2.892
0.658 0.706 0.764 0.885 0.937 1.384 1.505 1.808 0.060
26A1 241Am
7.16x 105 years 4.33x102 years
195Au
1.83x102 days
198Au
2.70 days
133Ba
1.05•
7Be 2~
5.32x101 years 3.28x101 years
82Br
3.53•
hours
0.265 0.445
14C
45Ca
5.73• 103 years 1.63• 102 days
0.156 0.257 0.257
l~176
4.63•
139Ce 14ICe
1.174 5.443 5.486
XK
0.030 0.099 0.129 0.412 0.676 0.081 0.161 0.223 0.276 0.302 0.356 0.384 0.477 0.570 1.063 1.770 O.554 0.619 0.698 0.777 0.827 1.044 1.317 1.475
0.067
days
0.088
1.38• days 3.25x101 days
0.166 0.145
0.022 0.026 0.033 0.036
0.285 0.961
! years
0.435 0.580
0.072 0.031 0.035
0.075
Man-made RadioactiviO'
97
Radionuclide
T1/2
~max
144Ce+144pr
2.85x 102 days
0.185 0.238 0.318 2.996
252Cf
9.67• 102 days
36C1 244Cm
3.01 x 105 years 1.81xlO 1 years
57Co
2.72x 102 days
58Co
7.08x101 days
0.475
60Co
1.93x103 days
0.318
5~Cr
134Cs
2.77x101 days 7.55x102 days
137Cs+137mBa
3.02x101 years
169Er
9.40 days
J52Eu
1.35x 101 years
6.075 6.118
XK 0.080 0.134 0.697 1.489 2.186 0.043 0.100
0.036
0.014 0.122 0.136 0.811 0.864 1.675 1.173 1.333 0.320 0.563 0.569 0.604 0.796 0.802 1.168 1.365 0.662
0.006
0.709 5.666 5.763 5.805
0.089 0.415 0.658
0.511 1.173 0.343 0.352 0.387 0.698 1.475
0.006
0.005 0.032
0.032
0.008 0.122 0.245 0.296 0.344 0.411 0.444 0.689 0.779 0.867 0.964 1.086 1.112 1.213 1.299 1.408 continued
Chapter 4
98
Table 4.9 (continuation) Radionuclide
TI/2
55Fe
9.79x 10 2 days
59Fe
4.45x101 days
67Ga
3.26 days
3H 2O3Hg
1.23x101 years 4.66x101 days
0.018
166mHo
1.20x 10 3 years
1231
XK
~max
0.006 0.007
0.273 0.466
0.192 0.335 1.099 1.291 0.091 0.185 0.300 0.394
0.008
0.212
0.279
0.072
0.032 0.072 1.314
0.081 0.184 0.280 0.411 0.529 0.712 0.810 0.830
0.049
1.32x101 hours
0.159 0.529
0.027
125I
5.99x 101 days 1.57x 10 7 years
0.051
0.035 0.039
0.027
129I 1311
8.02 days
0.334 0.606
111in
2.80 days
192ir
7.38• 101 days
0.256 0.536 0.672
0.296 0.308 0.316 0.468 0.604 0.612
40K
1.26• 109 years
1.312
1.469
42K
1.24• 101 hours
1.996 3.521
1.524
85Kr
1.07•
0.173 0.687
0.514
0.080 0.284 0.364 0.637 0.722 0.171 0.245
years
0.030
0.023
Man-made Radioactivi~
99
Radionuclide
T1/2
[~max
176Lu
3.79• 101~years
0.589
54Mn 99Mo+99mTc
3.12• 2.75 days
0.436 0.848 1.214
22Na
9.50x 102 days
0.545
24Na
1.50x 101 hours
1.390
63Ni 237Np+233pa
1.00x 102 years 2.14x 106 years
32p 21opb 147pm 21Opo 238pu
1.43x 101 days 2.22x 101 years 9.58x 102 days 1.38x 102 days 8.77x101 years
239pu
2.41 x 104 years
l~176
3.93x 101 days
0.113 0.226
0.040 0.497 0.610
3.73x 102 days
0.979 2.407 3.029 3.541
0.512 0.622 1.050
35S
8.74x101 days
0.167
125Sb+125mTe
1.01xl03 days
0.095 0.125 0.303 0.446 0.622
l~
+l~
days
4.766 4.771 4.788
0.066 0.156 0.174 0.232 0.260 0.572
XI< 0.088 0.202 0.307 0.835 0.140 0.181 0.739 0.778 0.511 1.275 1.368 2.754
0.056
0.030 0.087 0.143 0.195 0.312 0.340
0.098
0.005
1.710 0.047 0.225 0.803
5.305 5.456 5.499 5.105 5.143 5.156
0.176 0.428 0.463 0.600 0.636
0.027
continued
l O0
Chapter 4
Table 4.9 (continuation) Radionuclide
TI/2
75Se
1.20• 102 days
113Sn+113mIn
1.15• 102 days
85Sr
6.49•
XK
~max
days
89Sr
5.06• 101 days
9~176
2.82• 101 years
99Tc 99mTc 228Th
2.14• 105 years 6.01 hours 6.99• 102 days
2~
3.04 days
2~ 232U
1.38• 103 days 6.98• years
233U
1.59• 105 years
235U
7.04• 108 years
127Xe
3.64•
133Xe 88y
5.41 days 1.07• 102 days
0.346 0.755
9Oy
2.67 days
2.284
0.121 0.136 0.265 0.280 0.401 0.255 0.392 0.514
0.010
0.140 0.084 0.132 0.216 0.135 0.167
0.018
0.024 0.013 0.015
1.492 0.546 2.284 0.294 5.340 5.423
0.071
0.763 5.263 5.320 4.824
0.058 4.783
4.218 4.365 4.400 4.556 4.599
days
0.042 0.054 0.097 0.146 0.164 0.291 0.317 0.109 0.143 0.163 0.185 0.205 O.057 0.145 0.172 0.203 0.375 0.081 0.898 1.836
0.029
0.013 0.090 0.105
0.028 0.029 0.032 0.033 0.031 0.014
Man-made Radioactivity
101
Radionuclide
TI/2
ot
~max
Y
XK
169yb
3.20• 101 days
0.050
days
0.330
0.063 0.109 0.130 0.177 0.198 1.115
65Zn
2.44•
95Zr+95Nb
6.40x101days
0.366 0.399
0.008 0.009
0.724 0.756 0.766
9 europium-152 in the 100 to 1500 keV energy range, (Fig. 4.4b) 9 barium-133 in the 30 to 400 keV energy range, (Fig. 4.4a) 9 the mixed radionuclide (241Am, l~ 57C0, 139Ce,5~Cr, ll3Sn, 85Sr, 137C8,6~
88y)
in the 60 to 1836 keV energy range (Fig. 4.4c). Compared to the above standard solutions, this mixture provides a simpler spectrum, but varies greatly in time because of difference between the half-lives of the radionuclides. Solutions are generally supplied in sealed ampoules. However, solutions with high activity concentrations are supplied in capped vials, in order to allow easy handling. For environmental survey, multigamma standard solutions are characterised by: 9 low activity concentrations in large volumes, approximating the experimental conditions for the monitoring of radioactive effluents; 9 a packaging allowing the use of the standard without opening the container, for the direct calibration of NaI(T1) and Ge(Li) detectors. The standard sources have been designed in order to allow the calibration of all the classical detectors of or, [3, e-, 7, n, X radiation (ionisation chambers, Geiger-Mtiller or proportional counters, scintillation or solid-state counters, etc.). They are classified as: alpha sources, electron sources, beta sources, gamma sources, neutron sources, X-ray sources, heat flux sources, and sources for radiation protection dose meters. Other solid sources are supplied as references and standards for biology and medicine including: iodine-125 mock standard, sources and accessories for gammacameras, and gamma reference sources for dosimeters. Alpha sources as standard sources are characterised either in terms of activity (Bq) or in terms of emission flux in 2rt sr (s-l). The radionuclide is electroplated, either on a polished stainless steel disc 25 or 30 mm in diameter and 0.5 mm thick, or polished platinum disc 22 mm in diameter and 0.1 mm thick. The contribution of the sources to the FWHM of a spectrometer is about 1 keV, the total FWHM being thus for a commercial spectrometer less than 15 keV. All these sources can be used for energy calibration of efficiency calibration for all detectors and c~ measuring devices (see Table 4.10 for the list of alpha sources).
Chapter 4
102
a) o~
i -
i~
r
L
t~
eq
oO e~
. . . . . . . . . . . .
..;,
i
...... _
L_ . . . . . . . . .
i iI ....................
t
0
I~
il
i
100
200
.......
300
keV
b)
F eq
:
i
t _
i
i~
eq
!I L i
.... _
~
~
eq
......
"'IL~
J-
i
b _
500
_
J
1500
1000
keV
..,..
t~
e~
oc
i
k,, ' i il 0
500
1000
1500
keV
Fig. 4 . 4 . G a m m a ray e n e r g y spectra. (a) B a r i u m - 1 3 3 ; (b) E u r o p i u m - 1 5 2 ;
(c) m i x t u r e s o f r a d i o n u c l i d e s .
Man-made Radioactivit3,
103
Electron sources for spectrometry as standard sources are characterised in terms of electron emission flux in 4rt sr solid angle, expressed in s-1. They are point sources (~3<4 mm) very thin and deposited on a metallised mylar film (1 mg x cm -2 thick); they are mounted on a metallic ring which provides rigidity and easy handling. The contribution of the source to the F W H M of a spectrometer is less than 1 keV, the total F W H M being thus for a semiconductor detector, less than 3 keV (for the list of the sources available see Table 4.11). Sources for calibration of beta detectors as standard sources are characterised in terms of emerging [~ emission flux in 4rt sr solid angle, expressed in s-~. Intended for the Table 4.10 Alpha sources Radionuclide
Half-life
c~ energies (MeV)
241Am
4.33x102 years
5.443 5.486
244Cm
1.81 x 101 years
5.666 5.763 5.805
21~
1.38x102 days
5.305
238pu
8.77x101 years
5.456 5.499
239pu
2.41 x 104 years
5.105 5.143 5.156
233U
1.59•
4.783 4.824
years
Table 4.11 Electron sources for spectrometry Radionuclide
Half-life
Radiation energies (MeV)
l~176
4.63• 102 days
0.062 0.084
139Ce
1.38• 102 days
0.127 0.165 0.165
137Cs+137mBa
3.02• 101 years
0.624 0.656
ll3Sn+ll3mln
1.15•
0.364 0.388
days
104
Chapter 4
Table 4.12 Sources for calibration of beta detectors Radionuclide
Half-life
Radiation energies ~max(MeV)
14C 144Ce+l~Pr
5.73x103 years
0.156
2.85x 102 days
36C1
3.01x105 years
0.185 0.238 0.318 2.996 0.709
6~
1.93x 103 days
0.318
134Cs
7.55x102 days
137Cs+ 137mBa
3,02x 101 years
22Na
9.50x 102 days
0.089 0.415 0.658 0.511 1.173 0.545
147pm
9.58x 102 days
0.225
89Sr
5.06x 101 days
1.492
9~176
2.82x 101 years
2~
1.38x103 days
0.546 2.284 0.763
calibration in efficiency of [3 detectors and counting systems, these sources are hot sealed between two thin plastic foils and gold-coated. For rigidity and ease of handling, the foils are mounted on a metallic ring. This ring can be removed if necessary so that the source can be used with or without the ring holder to calibrate all 13 detectors, including windowless 2rt or 4rt counters (see Table 4.12). Two categories of gamma standard sources are provided: 1. Sources for activity calibration: These sources enable a direct calibration of y-ray spectrometers, for radionuclides for which standards are available. Moreover, with the kits of y-ray sources, the efficiency/energy curve can be plotted; in this case, knowledge of decay scheme parameters of the radionuclides involved is needed (y branching ratio, internal conversion coefficient, etc.). 2. Sources for efficiency calibration: These y-ray emission standard sources, and the associated kits, enable a direct and accurate plot of the efficiency/energy curve of y-ray spectrometers, without use of the decay parameters of radionuclides. The multigamma standard sources enable a direct and rapid plot of the efficiency/energy curve of y-ray spectrometers, without use of the decay parameters of radionuclides (Fig. 4.4c). These two methods differ mainly by their accuracies. For calibration at high energies, a 6.13 MeV special source is proposed.
Man-made RadioactiviO'
105
Gamma-ray sources for activity calibration as standard sources are characterised in terms of activity, expressed in kBq. They can be used for the calibration of all gamma-ray detectors and spectrometers NaI(T1) or Ge(Li) (see Table 4.13 for the list of radionuclides available and their characteristics). Four types of sources are available: 1. Gamma-ray point sources: The point-source is hot sealed between two thin plastic foils (overall thickness: 24 mg x cm-2). The source is mounted in a plastic ring which provides rigidity and easy handling. 2. Gamma-ray large sources: The activity of the source is uniformly distributed on the surface of a disc 50 mm in diameter. It is hot sealed between two thin plastic foils mounted between two plastic discs (overall thickness: 460 mg x cm -2) for rigidity and easy handling. 3. Gamma-ray plastic sources: The semi-point source is deposited in the leak-proof cavity of a rigid plastic holder. 4. Gamma-ray cylindrical sources: The point source is mounted at the end of a plastic rod (diameter 9 mm). A disc can be fitted to this rod to facilitate its use in automatic samples changer. These sources are particularly suited for the calibration of welltype NaI(T1) scintillation counters. Gamma-ray sources for efficiency measurements as standard sources are characterised in terms of photon emission flux in 4re sr, expressed in s-~, for each specified gammaray. The activity of the source is indicated. When an activity standard is used to determine the efficiency of a y-ray spectrometer as a function of photon energy, certain decay scheme parameters are required (gamma branching ratio, internal conversion coefficient, etc.). In this case, the calibration uncertainty is the combination of the uncertainty on the activity of the standard and of the uncertainties on the parameters of the decay scheme. The X-ray sources have been designed to allow direct calibration efficiency/energy without knowledge of the decay schemes. For the list of radionuclides and their properties see Table 4.14. Beyond 3.5 MeV, no radionuclide is usable as a reference. The usual methods of calibration used at low energies (sets of different sources or multigamma sources) cannot be applied. Two methods are presently used for such calibrations: 1. A semi-empirical formula allowing the extrapolation of the calibration curve towards higher energies. 2. Calibration by means of high energy gamma rays from (n,y) or (p,y) reactions produced in accelerators or nuclear reactors. These methods are time consuming, expensive and often inaccurate. The 6.13 MeV reference is a composite source of 238pu and ~3C; it emits, by de-excitation of ~60, a 6.13 MeV gamma ray:
13C(~,rt)160"
~
gamma 6.13 MeV.
The spectrum contains the 6.13 MeV peak accompanied by two escape peaks (5.62 and 5.11 MeV). These three peaks are entirely separate from the 238pu gamma ray spectrum.
106
Chapter 4
Table 4.13 Gamma-ray sources for activity calibration Radionuclide
Half- life
"~energies (MeV)
241Am
4.33• years 1.05• 101 years
0.060
133Ba
l~176
4.63• 102 days
0.088
145Ce
3.25• days 2.85• 102 days
0.145
144Ce+144pr
57Co
2.72•
days
0.014 0.122 0.136
6~
1.93• 103 days
1.173 1.333
51Cr
2.77• 3.02•
days years
0.320
537Cs+137mBa 2~ 131I
4.66• 101 days 8.02 days
54Mn 22Na
3.12x 102 days 9.90• 102 days
0.279 0.080 0.284 0.364 0.637 0.722 O.835
l~176
3.93• l05 days
l~176
3.73• 102 days
85Sr 88y
6.49• days 1.07• 102 days
65Zn
2.44•
days
0.081 0.161 0.223 0.276 0.302 0.356 0.384
0.080 0.134 0.697 1.489 2.186
0.662
0.511 1.275 0.040 0.497 0.610 0.512 0.622 1.050 0.514 0.898 1.836 1.115
107
Man-made Radioactivi~
Table 4.14 Gamma-ray sources for efficiency measurements Radionuclide 241Am
Half-life
3' energies (MeV)
4.33x102 years
0.060
4.63x 10e days
0.088
139Ce
1.38x102 days
0.166
57Co
2.72x 102 days
0.014 0.122 0.136
51Cr
137Cs+137mBa
7.55x 102 days 3.02x101 years
0.320 0.662
54Mn
3.12x102 days
0.835
ll3Sn+ll3mln
1.15x102 days
85Sr
6.49x101 days
0.255 0.392 0.514
65Zn
2.44x102 days
1.115
l~
l~
The 6.13 MeV gamma-ray is emitted without D6ppler effect; the sharpness of the ray allows us to test the resolving power of Ge high energy detectors. The half-life of the source is similar to that of 238pu, i.e. about 87 years.
4.3.3 Germany A number of interesting sources is manufactured and distributed by "DuPont": 9 DuPont Nemours (Deutschland) GmbH, Postfach 401240, D-6072 Dreieich, W. Germany, Telephone: (06103) 803-0, Telex: 4-17993 NEN D, Fax: (06103) 87897; 9 E.I. DuPont de Nemours & Co. (Inc.), 331 Treble Cove Road, North Billericay, MA 01862, USA. Toll-free 800-225-1572 (Telephone: 617-482-9595) Telex: 6817017, Fax: 617-663-7315 9 DuPont Canada, Inc., P.O. Box 660, Station A, Montreal, Quebec H3C 2V1, Canada, Telephone: 514-397-2748, Telex: 05-267687, Fax: 514-397-2720. Their products include: radiopharmaceuticals, radioimmunoassay kits for medical diagnosis, radiolabelled and liquid scintillation chemicals for research, and radioactive sources used in nuclear medicine, research and industry. Some of their sources and uses are listed below: 9 63Ni: The physical and chemical properties of 63Ni uniquely suit it to applications where a source of safe localised ionisation is required. [Half-life: 96 years; Beta energy (maximum): 66 keV, (average): 17 keV]. The long half-life decay of 63Ni
Chapter 4
108
to stable 63Cuprovides a constant ionisation level, while the pure low energy beta decay minimises the need for external shielding. 9 85Kr: The 10.7-year half-life and 0.69 MeV E max beta particle make Kr-85 ideal for weight and thickness gauging in manufacturing process control for paper, plastic film and rubber sheet. X-ray sources: X-ray fluorescence analysis is a technique for elemental identification and quantification. X-ray fluorescence sources provide stable outputs for energy, direction and intensity. Unlike competing technologies, radioisotope sources do not require external power supplied. X-ray fluorescence sources are useful either for single- or multi-element analysis. Characteristic X-ray line spectra of the elements are excited in a specimen by the source X-rays. A source is selected whose X-ray energy is slightly above the X-ray energy threshold, or "absorption edge", of the element being analysed (see Table 4.15). In energy dispersive analysis, the X-rays from the sample interact with a semi-conductor detector that gives pulses directly proportional to the energy of each X-ray. The detected count rate of the characteristic X-ray pulses is proportional to the weight fraction of the element(s) in the sample. Several sources with different X-ray photon energies can be used to analyse several elements from sodium to uranium. Proper selection of source and secondary exciter targets provide optimum sensitivities. Wherever possible, calibrated sources produced by DuPont are directly calibrated against standards certified by the National Bureau of Standards (NBA) using instruments in DuPont laboratories. DuPont participates in a Measurement Assurance Program organised by the Atomic Industrial Forum and NBA. 9
Table 4.15 X-ray fluorescence sources/exciter systems Source nuclide
Half-life
Excitation mode
Photon emission and energies
Abundance Element X-rays excited usefully
Fe-55
2.7 y
Direct
Mn K X-rays; 5.9 and 6.5 keV
28%
Na-V K X-rays Zn-Ce L X-rays
Cd- 109
462.6 d
Direct
Ag K X-rays; 22 and 25 keV
102%
Ca-Mo k X-rays W-U L X-rays
Am-241
432.2 y
Direct Secondary*
Gamma rays; 59.6 keV Mo target 17.4; 19.6 keV K X-rays
36%
Sm-Tm K X-rays V-Tc K X-rays Pr-Pu L X-rays
Co-57
271.7 d
Direct
Gamma rays; 14, 122, 136 keV
10%, 86%, 11%
Ta-U K X-rays
*Secondary mode target energy selection can be made from virtually any element which can be formed into a target.
Man-made RadioactiviO,
109
Table 4.16 Calibrated beta reference sources. Individual beta reference sources calibrated with an accuracy of _+3 to 5% (at 99% confidence). Nuclide
Half-life
Nominal activity
Principal energies (Emax 13)
Bismuth-210
22 y (Lead-210 parent)
740 Bq (0.02 ~tCi)
1.16 MeV 13
Calcium-45
164 d
0.37 MBq (10 ~tCi)
0.25 MeV 13
Carbon- 14
5730 y
5550 Bq (0.15 ~Ci)
0.156 MeV [3
Caesium- 137
30 y
1480 Bq (0.04 ~tCi) 0.37 MBq (8 ~tCi)
0.52; 1.1 MeV 13 0.662 MeV 3t
Chlorine-36
3x105 y
740 Bq (0.02 ~tCi) 740 Bq (0.02 ~tCi) 0.074 MBq (2 ~tCi)
0.714 MeV [3
Cobalt-60
5.25 y
1480 Bq (0.04 ~tCi) 0.030 MBq (0.8 ~tCi)
0.31 MeV 13 1.17; 1.33 MeV y
Nickel-63
96 y
0.185 MBq (5 ~tCi)
0.066 MeV [3
Phosphorus-32
14d
0.37 MBq (10 ~Ci)
1.71 MeV [3
Promethium- 147
2.6y
3700 Bq (0.1 ~tfi) 3700 Bq (0.1 ~tCi)
0.224 MeV 13
Strontium-90 Yttrium-90
28.5 y
740 Bq (0.02 ~tCi) 740 Bq (0.02 ~tCi) 3700 Bq (0.1 ~tCi)
0.546 (2.27) MeV 13
Sulfur-35
87 d
0.37 MBq (10 ~tCi)
0.167 MeV
Technetium-99
2x 105 y
1480 Bq (0.04 ~tCi) 1480 Bq (0.04 ~Ci)
0.292 MeV [3
E a c h c a l i b r a t e d r e f e r e n c e s o u r c e is supplied with a calibration certificate listing: 9 the s o u r c e nuclide, activity and date o f calibration; 9 the radiation e m i t t e d by the source (half-life and principal radiations with abundance); 9 a d e s c r i p t i o n o f the p h y s i c a l f o r m o f the source; 9 a d e s c r i p t i o n o f the m e t h o d o f calibration; 9 the r a d i o a c t i v e impurities, if any; 9 A n analysis o f the r a n d o m and s y s t e m a t i c errors a s s o c i a t e d with the calibration measurement. T h e list o f c a l i b r a t e d b e t a r e f e r e n c e s o u r c e s is s h o w n in T a b l e 4.16, while calibrated g a m m a r e f e r e n c e s o u r c e s are s h o w n in T a b l e 4.17.
110
Chapter 4
Table 4.17 Calibrated Gamma Reference Sources Nuclide
Half-life
Nominal activity
Principal energies (Ema x [~)
Americium-241
432.2 y
0.185 MBq (5 ~tCi)
0.060 MeV y
Barium-133
10.5 y
3700 Bq (0.1 ~tCi) 0.259 MBq (7 ~tCi) 0.037 MBq (1 ~tCi)
0.080, 0.302, 0.356 MeV y (others at 0.276, 0.054, 0.161 MeV)
Cadmium- 109
462.6 d
0.037 MBq (1 ~tCi) 0.296 MBq (8 ~tCi) 0.296 MBq (8 ~tCi)
0.088 MeV y 0.023 MeV X-ray
Caesium-137
30 y
3700 Bq (0.1 ~tCi) 3700 Bq (0.1 ~tCi) 0.037 MBq (1 I.tCi) 0.259 MBq (7 ~tCi) 0.037 MBq (1 ~tCi) 0.037 MBq (10 ~tCi) 0.296 MBq (8 ~tCi)
0.662 MeV y 0.031 MeV X-ray
Chromium-51
27.7 d
0.37 MBq (10 ~tCi)
0.320 MeV y 0.0049 MeV X-ray
Cobalt-57
271.7 d
3700 Bq (0.1 ~tCi) 3700 Bq (0.1 ~tCi) 0.259 MBq (7 ~tCi) 0.037 MBq (1 laCi) 0.37 MBq (10 ~tCi)
0.122, 0.136, 0.01 MeV 7 0.006 MeV X-ray
Cobalt-60
5.27 y
3700 Bq (0.1 [aCi) 0.030 MBq (0.8 ~tCi) 0.259 MBq (7 ~tCi) 0.030 MBq (0.8 ~tCi) 0.030 MBq (0.8 ~tCi)
1.173, 1.333 MeV 7
Gadolinium- 153
242 d
3700 Bq (0.1 ~tCi)
0.97, 0.103 MeV 7 0.043 MeV X-ray
Iodine- 125
59.6 d
0.37 MBq (10 ~tCi)
0.035 MeV y 0.027 MeV X-ray
Iodine-129
1.6 x 107 y
2960 Bq (0.08 ~tCi) 2960 Bq (0.08 ~tCi)
0.039 MeV y 0.029 MeV X-ray
Iodine- 131
8.021 d
0.296 MBq (10 ~tCi)
0.364, 0.637 MeV y
Manganese-54
312.14 d
5500 Bq (0.15 ~tCi) 0.259 MBq (7 ~tCi) 0.037 MBq (l~tCi)
0.836 MeV y 0.0054 MeV X-ray
Sodium-22
2.6 y
3700 Bq (0.1 ~tCi) 0.259 MBq (7 ~tCi) 0.037 MBq (1 ~tCi)
0.511, 1.27 MeV y
Man-made Radioactivity
111
4.3.4 United Kingdom AEA Technology is the commercial division of the United Kingdom Atomic Energy Authority: Their address is: 9 AEA Technology, 220 Harwell, Didcot, Oxfordshire OX11 0RA, U.K. Telephone: (+44) 235 434212; Fax: (+44) 235 434522. Their 9 9 9 9 9 9
production line includes" Solid alpha and beta sources High purity tracer solutions Bulk dispensed radionuclides Plutonium and uranium s t a n d a r d s - certified nuclear reference materials Standard sources for non-destructive plutonium assay Californium fission fragment sources.
Some 9 9 9 9 9 9
of the special sources the company has supplied include: 36C1 and 9~ check sources for contamination-in-air monitors Very high resolution 241Am source, <10 keV FWHM 20 Bq natural uranium sources on 100 x 160 mm plates 241Am source of specified activity in epoxy resin High-intensity 238pu sources for diagnostic probes Mixed-actinide sources (23Su + 238pu + 239pu + 244Cm + 252Cf) for alpha energy calibration.
Their stock of alpha radiation sources are listed in Table 4.1 8. Table 4.18 Alpha Radiation Source Isotope(s)
Principal alpha energy (MeV)
Nominal activity (kBq)
e39pu, e41Am, 242Cm
5.157, 5.486, 5.805
0.06-0.08, 0.5-0.7, 5-8, 4-6
241Am, ea3Am, 244Cm
5.486, 5.274, 5.805
0.003-0.2, 0.08-0.9, 0.001-0.010
237Np, 241Am, 244Cm
4.788, 5.486, 5.805
0.06-0.4, 0.1-0.6, 0.5-1.4
233U, 238pu, 239pu, 244Cm, 25~
4.824,5.499, 5.157, 5.805, 6.118 0.1-0.2, 0.1-1.0, 2.0-5.0
natu, 239pu, 241Am
4.196, 4.775, 5.157, 5 . 4 8 6
0.002-0.005
239p
5.157
0.8, 0.4-5.5
24~
5.168
0.6-9.3
241Am
5.486
0.7-0.8, 0.04-66, 0.3-1.5, 0.1-2.8
244Cm
5.805
1.5-3.5
1 12
Chapter 4
The high purity tracer solutions specially formulated for environmental analysis and equilibrium studies include a wide range of radionuclides. The list of radioisotopes includes: 93mNb
2~opb
208p0
209p0
210po
223Ra
226Ra 232Th
227Ac 23~Pa
227Th 233pa
228Th 232U
229Th
23~
233U
234U
235U
236U
237U
238U
natU
235Np
237Np
236pu
238pu
239ptl
240pt1
241pu
ea2pu
241Am
243Am
242Cm
244Cm
2s~
All of them are in carrier-free solutions and sealed in glass ampoules for highest purity with certified isotopic composition and each sample individually documented. The list of plutonium and uranium standards used as certified nuclear reference materials in safeguards-related measurements is shown in Table 4.19. Another institution of interest in the U.K. is: 9 Centre for Ionising Radiation Metrology, National Physical Laboratory, Teddington, Middlesex TW 1 1 0LW, U.K. Telephone: (+44) 18 1 977 3222; Fax: (+44) 18 1 943 6 16 1, E-mail:
[email protected]. NPL radioactivity standards include: 9 Gamma-ray emitting standards: 7Be, ~SF, 24Na, 42K, 43K, 465c, 47Ca, 475c, 51Cr, 54Mn' 55Fe, S6Co' 56Mn ' 57C0 ' 58C0 ' 59Fe ' 6~ ' 64Cu ' 65Zn ' 67Cu ' 67Ga ' 68Ga ' 75Se ' 82Br' 85Sr' 86Rb ' 87mSr' 88y, 95Nb ' 99M0 ' 99mTc ' 106Ru ' 109Cd ' ~ I n , 1J3In, Jl3Sn, 1231, 1241, 125i, 1255b ' 129I, 131I, 1321, 133Ba ' 134Cs ' 137Cs ' 139Ce' 141Ce ' 144Ce ' 152Eu ' 153Gd ' 1535m ' 154Eu' 155Eu' 16~ ' 169yb ' 17~ 182Ta ' 192ir' 197Hg ' 198Au ' 199Au ' 2~ ' 2~ ' 2~ ' 232U, 233pa, 237U, 237Np, 238pu, Z39Np, 241Am, 242r~ t'u, and 243Am"
9 Pure beta-ray emitting standards: ~4C,32p, 355, 89Sr ' 90Sr+90y, 99Tc ' 147pm ' and 2~ 9 Gamma-ray reference sources: 22Na, 54Mn, 56C0, 57C0, 58C0, 6~ 75Se, 88y, 109Cd ' ll3Sn, 133Ba, 137Cs, 139Ce, 152Eu and 169yb. 9 Gas standards: Tritium (3H), Radon(222Rn), 85Kr and ~33X. Standards of radioactive surface contamination: 9 beta-emitters: 3H, 14C, 147pm, 36C1, 9~176 9 alpha-emitters: 237Np, 238U, 241Am photon-emitters: 55Fe. 9 NPL secondary standard radionuclide calibrator standards: 7Be, 18F, 22Na, 42K, 465c ' 47Ca ' 475c ' 51Cr ' 54Mn ' 57C0 ' 58C0 ' 59Fe ' 6~ ' 65Zn ' 6VGa ' 755e ' 82Br, 85Kr' 85Kr, 86Rb ' 87mSr ' 88y, 99Mo ' 99Tc ' 106Ru ' m9Cd ' ~ I n , 1135n, 123I, 124I, 125I, 131I, 133Ba, 133Xe, 9
134Cs ' 137Cs ' 139Ce ' 141Ce ' 144Ce ' ~5:Eu ' 153Gd ' 153Sm ' 154Eu' 16~ ' 169yb ' 192ir' 197Hg ' 198Au ' 199-/A_u, ZOlT1' 203pb' 203Hg ' 233pa ' 237Np ' 239Np ' 241_ /-km.
9 Pure beta-emitters: 32p, 89Sr and 90y; Brachytherapy sources: 192Irwire and hairpin sources. 9 Environmental standards: 241Am, 243Am, laG, 134Cs, 137Cs, 152EH, 129I, 242pu, 9~
99Tc, 232U,mixed,
gas and natural matrix" milk powder
(137Cs& 134Cs),222Rn(gas).
Man-made Radioactivi~.
113
Table 4.19
Certified nuclear reference materials (Characterised and certified for analysis and safeguards) Application
Material
Composition, certified values
Elemental Analysis
Uranium dioxide pellets Plutonium-gallium alloy
Isotopic Analysis
Uranyl nitrate solution ( 1:1:1) Uranyl nitrate solution (1:1:1) Plutonium nitrate solution (1:1 : 1) Plutonium nitrate solution (1:1:1) Plutonium nitrate solution (6% 24~ Plutonium nitrate solution (3:3:3:1) Plutonium nitrate solution (34% 24~
(88.136_+0.006)wt% U (98.075-+0.017) wt% Pu 233U:235U (0.9991,+0.0004) 238U:235U (0.9994,+0.0006) 233U'235U (0.9991,+0.0004) 238U'235U (0.9994,+0.0006) 24~ (0.9999,+0.0006) 242pu:239pu (0.9999_+0.0010) 24~ (0.9999-+0.0006) 242pu:239pu (0.9999-+0.0010) 24~ (0.0638_+0.0007)
Burnup Analysis
U+Pu+Nd nitrate solution U+Pu+Nd nitrate solution Nd isotopic mixtures
24~ (0.9662,+0.0011) 242pu:239pu (1.0253,+0.0019) 244pu:239pu (0.3358,+0.0008) 24~ 239pu (0.54 37,+0.0003 )
99.2% U, 0.5% Pu, 0.3% Nd In preparation 67% U, 30% Pu, 3% Nd In preparation 142Nd/143Nd (0.087966,+0.000097) 144Nd/la3Nd (1.512985,+0.000670) 145Nd/143Nd (0.861148-4-0.000476) 146Nd/143Nd (0.766473,+0.000385) 148Nd/la3Nd (0.432400,+0.000337) 15~ (0.195368,+0.000190)
4.3.5 Russia Information from Russia is limited and is based on personal communications. The address below is of interest for the many radioisotopes they produce, but especially highly enriched actinide isotopes as shown in Table 4.20 (see Vesnovskii and Polynov, 1992; Vesnovskii et al., 1992; Vesnovskii et al., 1993): 9 Dr. S.P. Vesnovskii, Radiochem. Dept., Russian Federal Nuclear Center, 607200 Arzamas-16 Nizhny Novgorod Region, Russia, Fax: 831-30-54565. Electromagnetically enriched radioactive Cm, Am and Pu isotopes are available also through the following enterprise: 9 All-Union Foreign Economic Association "Techsnabexport" Staromonetniy Per., 26, 109180 Moscow, Russia, Tel.: 233-48-46, Telex: 411328 TSESU, Fax: 233-18-59.
114
Chapter 4
Table 4.20 Highly enriched actinide isotopes u-233 u-235 U-238 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Am-241 Am-242m Am-243 Cm-243 Cm-244 Cm-245 Cm-246 Cm-247 Cm-248
99.947% 99.993% 99.997% 99.6% 99.3% 100% 93% 99.98% >99.9% 73% 99.998% 93.6% -100% 99% 80-95% 73-85% 97%
The isotopes of Cm, Am and Pu are supplied as: 9 oxides, nitrates, chlorides in glass ampoules; 9 thin layers on solid backings of all sorts. Highly enriched curium isotopes: Cm-243: 93.3-99.99%; Cm-244: 99.9%; Cm-245: 98.4-99.998%; Cm-246: 99.5-99.8%; Cm-247: 70.3-90.2%; Cm-248: 95.8-97.0, americium isotopes: Am-242m: 66.8-73.6%; Am-243: 99.2-99.94%, and plutonium isotopes: Pu-238: 99.6; Pu-239: 99.9977%; Pu-240: 99.9-100%; Pu-241: 99.699.998%; Pu-242: 97.8-99.96%, are available for scientific and applied research in physics, chemistry, geology, medicine, biology and other fields.
4.3.6 Others Many reactors and accelerators around the globe are involved in the business of radioisotope production. Most of them satisfy only local needs; many would be glad to find customers outside their region. As a matter of interest, we mention: 9 Institute of Physics & Nuclear Engineering, Bucharest, M~gurele, P.O. Box MG6, R-76900 Romania. Telephone: 401 780-59-40, Fax: 401 312-11-45, Telex: 11350. They produce 192Ir, a radioactive source used in brachiotherapy. The source in the shape of a wire (~ = 1 mm, h = 40 mm) is packaged in a stainless steel cylinder and is used for the treatment of malignant tumours. In addition, they produce sets of standard radioactive sources and solutions, whose activities are certified with an expanded uncertainty of only a few percent. The description of different sets and their characteristics are shown in Table 4.21.
0
0
~, i ,
r r~
o
0
0 r~
~
0
r r~
0 . ,...~
r
<
~ ,i~
~
8~
0
.>_ ~.
"~ = .= .~
iR
E
.=
,., ~
r~
~
,-~
~.-~
E E
E
<
cq
+1
QQ
C',I
m
c,l
m
...-
E
[-
+
,ll,
~
+1
E
VI
"-~
C'-I
r.~
+1
[.~l]
VI
Q~|
:
E
,.~
"~.~ .~ ~ ~ . ~ ; -..~ ~ . z 9
.,~
~'~= q.)
~5
Man-made Radioactivi O,
C'.I
[-,
r.~ .ill.
t.0
o~
r.~
E E
u
r,j
.-,
E E C~
Q|
r r.~
E E r,3 r,D
~
•
o
l~
o
r.~
E
EE~
E E
~E
r.,j
II
~Q
II
QQ
+1 vI
o ~
,..
#
-~
.~
,i~
_
•
,,,,i.
x
%
r
e... c~
r.~ "I., ~
._o
0
. ,,.i4
0 r.~
r.~
115
116
Chapter 4
REFERENCES Even, G.W., Tracer techniques in hydrology. Appl. Radiation Isotope 34, 1 (1983) 451-475. Holmshaw, R. (ed.), Physics of Industrial Radiology. Heywood Books, London, 1966, 498 pp. International Atomic Energy Agency, Isotopes in Everyday Life. IAEA, Vienna, 1993. International Commission on Radiological Protection. Recommendations of the International Commission on Radiological Protection (adopted 17 Jan, 1977). ICRP Publication 26. Annals of the ICRP, Sowby, F.D. (ed.), Pergamon Press, Oxford, 1977. Kazantzis, G., In: Friberg, L., Nordberg, G.F. and Vouk, V.B. (eds.), Handbook of the Toxicology of Metals. Elsevier, Amsterdam, 1986, Ch. 22, pp. 549-567. Pennal, D.J., Underwood, R., Costa, D.C. and Ell, P.J., Thallium Myocardial Perfusion Tomography in Clinical Cardiology. Springer-Verlag, London, 1992. Radiological Research and Radiotherapy, Vols. I and II. Proc. Int. Syrup. on Radiological Research needed for the improvement of Radiotherapy. Vienna 22-26 Nov. 1976. IAEA Vienna, 1977. Stein, S. and Aaseth, J., Thallium-201 as an agent for myocardial imaging studies. Analyst 120 (1995) 779. The Application of Radioactive Tracers in the Water Industry, in Four Essay Reviews on Applications of Radiation Measurement in the Water Industry 1984, HMSO, in this series. Vesnovskii, S.P., Polynov, V.N., Nikitin, E.A. and Vjachin, U.N., Quality and availability of actinide isotopes from All-Russian Scientific Research Institute of Experimental Physics in Arzamas-16. Nucl. Instrum. Meth. Phys. Res. A334 (1993) 37. Vesnovskii, S.P., Polynov, V.N. and Danilin, L.D., Highly enriched isotope sample of uranium and transuranium elements for scientific investigation. Nucl. Instrum. Meth. Phys. Res. A312 (1992) 1. Vesnovskii, S.P. and Polynov, V.N., Highly enriched isotopes of uranium and transuranium elements for scientific investigation. Nucl. Instrum. Meth. Phys. Res. A312 (1992) 9. Witsenboer, A.J., de Goeij, J.J.M. and Reiffers, S., Production of '23I via proton irradiation of 99.8% enriched '"4Xe. J. Labelled Compounds Radiopharm. 23 (1986) 1284.
117
CHAPTER 5
Measurements of Radioactivity
5.1 RADIATION INTERACTION WITH MATTER The detection of radiation is based on the interactions of the various types of ionizing radiations with matter. The differences between the interactions and the penetrating abilities of the various radiations are very relevant to radiation detection and measurement---e.g, they partly explain the variety of detector types and designs. Radiations can be grouped into directly ionizing radiations and indirectly ionizing radiations. Directly ionizing radiations include all charged particles such as alpha particles and heavier ions and beta particles. All charged particle radiations lose energy interaction with the orbital electrons or nuclei of atoms in the materials they traverse. There are two main processes involving the orbital electrons: 1. Atomic or molecular excitation, with the emission of light resulting from subsequent de-excitation. 2. Ionization, which involves the ejection of an orbital electron, resulting in the creation of an ion pair. Indirectly ionizing radiations include some types of electromagnetic radiations and neutrons. These radiations interact with matter by giving rise to secondary radiation which is ionizing. Indirectly ionizing radiations lose energy by collisions with electrons, or atomic nuclei, and the charged particles thus set in motion interact in turn with the orbital electrons or nuclei.
5.1.1 Heavy charged particles Heavy charged particles lose energy by Coulomb interaction with the electrons and the nuclei of the absorbing materials. The collision of heavy charged particles with free and bound electrons results in the ionization or excitation of the absorbing atom, whereas the interaction with nuclei leads only to a Rutherford scattering between two types of nuclei. Thus the energy spent by the particle in electric collisions results in the creation of electron-hole pairs, whereas the energy spent in nuclear collisions is lost to the detection process.
118
Chapter 5
The interactions of charged particles with matter are generally expressed in terms of the energy loss per unit path length (also known as stopping power) and the total range of the particle. The range is most important in detector selection, although some applications require the stopping power for particle identification. For heavy charged particles in which the path is a straight line, the theoretical energy loss while undergoing collisions with electrons is given by:
dz
mv"
/
where m,e = mass, charge of electron; Z~ = atomic number of moving particle; Z 2 = atomic number of stopping material; v - ion velocity; N = number of atoms per unit volume; and I = average ionization potential of stopping atom. The logarithmic term varies slowly with energy so that it is approximately correct to write the energy loss as Ed ---~ = dx
Constant
E
(5.2a)
or as dE dx
=
Z~ Constant
E
(5.2b)
to identify the charged particle atomic number. The range is obtained by integrating the stopping power dE/dx over the energy: e~ dE range = J Emax( d E / d x )
(5.3)
Nuclear collisions can become an important part of the energy loss process, especially in the case of heavy ions and fission fragments. The theory describing this process is too complicated for a brief summary. Finally, it should be mentioned that channelling effects (the steering of charged particles in open regions in the lattice) could reduce the specific ionization loss. 5.1.1.1 Energy loss due to ionization
A charged particle moving through matter loses energy as a consequence of collisions with atomic electrons. Let us first consider the effect of a fast charged particle on an unbound electron. As the particle passes by, the electron "sees" the rapidly changing electric field as an impulse
Measurements of Radioactivit3,
119
(5.4)
Ap = FAt = -b2
where Z i is the atomic number of the incident charged particle, v is the centre-of-mass velocity, and b is the distance of closest approach, called the impact parameter. The energy transferred to the electron is then (AP) 2 2 Z 2 e 4 ( 1 ) E(b)= ~ = me v2 ~-y
(5.5)
In order to find the total loss of the kinetic energy by the charged particle to the atomic electrons in the medium, the number of electrons, dn, affected by the charged particle must be known. It follows
dE = E(b)dn
~ meV-
-'~
(5.6)
NZ, 2rcdbdx)
where N is the number density of the target atoms, and Z t is the atomic number of the target material. The above expression includes the estimate of the number of electrons within a cylinder having its axis along the path of the charged particle. The energy loss per distance traversed is then
47tZ i2Z ,e4 N b~ db --~ = me V2 b dE
x
bm m
4rr,Z2Zt e 4 N me V 2
bmax log~ b min
(5.7)
We shall now discuss the values of bmax and bmin. For bmax we shall consider electrons b bound in atomic orbits. In order to perturb their orbit the duration of perturbation,--, V 1
must be smaller than the periods, x, of the system under consideration: b/v < x = - This v determines bmax as"
V bmax "-(V Z
(5.8)
where (v) is the appropriate average of the electron frequencies. As a result, the electron binding energy becomes important~it impedes energy absorption by the bound electron. Taking the relativistic corrections into account, the duration of the perturbation is shortened by a factor ~ = (1 - [32)-1/2, therefore
bmax = '~v [ (v)
(5.9)
120
Chapter 5
The lower limit on b, bmi n, is determined by quantum mechanics since the electron can be localized with respect to the heavy ion only to the accuracy of its de Broglie wavelength: h bmi n > -
p
h
(5.10)
= ~
"y~VneV
By taking these values for bmin and bmax, o n e obtains" dE
dx
=
4 rtZ i2Z t e 4 N
mev 2
m e V 2 ~[,2
log ~
(5.11)
h(v)
The quantity 2rth( ) is a special average of the excitation and ionization potentials of the atoms in the stopping material. 2rth ( v ) = t
(5.12t
A more complete theory along these lines was developed by Bethe (1930) using Born' s approximation. He obtained the following expression: dE --~-=
42Z4I(Wme'21 ] nL___/L_.:e mev 2
In
[32
I
-
-C k
(5.13)
where C k is a correction term for the "nonparticipating" K-shell electrons that are strongly bound and are partially screened by outer electrons. The mean ionizationexcitation potential, L can be experimentally determined. Its relation to Z is given by the semi-empirical formula (Segre, 1965) I = 9.1 Z(1 + 1.9 Z -2/3) eV
(5.14)
The general form for the energy loss formula allows us to draw some conclusions. Energy loss of a charged particle is proportional to its mass, square of its charge and inversely proportional to its energy dE
Z2
~ O C
dx
v2
Z~M = ~
2E
(5.15)
Therefore, simultaneous measurement of kinetic energy, E, and energy loss, dE/dx, allows the determination of the mass of the charged particle since dE .E,,,:Z2M dx
-
(5.16)
Measurements of Radioactivity
121
1000
500 ~ Al~~',~
E "~t~
~
2~176 l
100~ c
50 20
> t
p
lO~
2 0.1
0.2
0.5
1.0
2
5
10
20
50
100 200
PROTON ENERGY / MeV Fig. 5.1. The rate of energy loss for protons in C, A1, F e and Pb.
10'~i-
1o
9 ,...1
10' z
10
1
10
100
---
ENERGY / MeV Fig. 5.2. Energy versus energy loss curves for hydrogen and helium isotopes in silicon.
Figure 5.1 shows the rate of energy loss for protons in some materials (C, A1, Fe, Pb) commonly used as targets or stopping foils; the unit of dE/dx is keV/(mg/cm2). Figure 5.2 shows energy loss curves for hydrogen and helium isotopes in silicon. These additional curves are given for convenience because silicon is a coinmon detector material, but, in general, these curves can be obtained from the proton curves by using the following relations:
Chapter 5
122
f
dx (deuteron) E= -~- (proton)
f
(5.17a)
2
dE
- ~ (triton) E= - ~
(pr~
(5.17b)
_~E 3
- - ~- ( H e 3
) E=
4-~- (proton) !E
(5.17c)
3
dE (a )l E= 4 -~dE (proton)I
(5.17d)
!e 4
The factor 4 that appears in eqs. (5.17c) and (5.17d) is valid only when the equilibrium charge of the helium ion is essentially 2. The specific ionization loss measures the amount of energy lost by the particle per unit-length of its track; the range indicates how deeply the particle penetrates the absorbing material. Silicon and germanium are the two most common materials in semiconductor industry and especially in radiation detector manufacturing. Therefore we present here the stopping power of Ge and Si for p, d and u-particles as a function of energy, as shown in Fig. 5.3
~E
lOO0 2C ~0 500 1C~0 5 ~0
>
3: 9
200 100
2 ~0
5(
130
O~
5O
20 Z
lO
20
9
5
10 5 2
1 Si Ge
i 0I 0.05 .1 0'.2 0'.5 ~ ~ ~
. 0.1 . . 0.2 . 0.05
. . 1 . 2. 0.5
1'0 A
. 5 . 10. 20.
5'0~00200 5~0 . 50 100 200
500
ENERGY / MeV Fig. 5.3. Stopping power vs. energy for protons, deuterons, and alpha particles in Si and Ge.
Measurements of Radioactivity
123
5.1.1.2 Ranges o f charged particles
The range of a charged particle of incident energy E i in a material in which its rate of energy loss is dE/dx is given by dE
(5.18)
R(Ei) = J
o dE / d x
If dE/dx is known for 0 < E _<E~, then the range can easily be calculated. As an example Fig. 5.4 shows the range (~tm) of protons in Fe for proton energies up to 3 MeV, while the range as a function of the energy in silicon and germanium for alpha particles, protons and deuterons is shown in Fig. 5.5. Unfortunately, stopping cross sections have not been measured for very low energies nor can they be calculated with reliability. Therefore, computed range-energy relations are subject to considerable uncertainty at low energies. On the other hand, range differences from, say, 1 MeV to E i can be calculated with confidence. The following curves give such range differences, i.e., dE Rdif ( E i ) "- J IMeVdE / dx
(5.19)
The total range is given by Rdif(Ei) + R(1 MeV). Table 5.1 lists some estimates of R(1 MeV) for the materials considered here. With better estimates (or actual measure40
E =i. Z
I 0
1.0
I
I
2.0
3.0
PROTON ENERGY / MeV Fig. 5.4. Range of protons in Fe.
Chapter 5
124
loo_ I I
I I t
I I f
50-20 - -
p(Ge)~.
10 - 5--
d (Ge)
1--
d (Si)\
0.5-0.2-0.1--
elO
z
_
p (Si)~
2 --
krd
I 1 [
0.05 - -
0.02
"\c~(Si)
0.0I - 0.005 - 0.002 0.001 0.0005
0.0002 l, I [ 0.05 0.1 0.2 0.5 l
I 2
l I l ] l I 5 10 20 50 100 200 500
ENERGY / MeV Fig. 5.5. Proton, deuteron and alpha particle ranges in Si and Ge.
Table 5.1
Approximate ranges for 1 MeV Material
Unit
Particle p
d
t
He 3
He 4
C
(mg/cm 2)
2.7
1.9
1.7
0.55
0.59
A1
(mm)
0.0146
0.011
0.010
0.0035
0.0037
Si
(mm)
0.0170
0.013
0.012
0.0041
0.0043
Fe
(mm)
0.0075
0.0061
0.0057
0.0019
0.0020
Ge
(ram)
0.0130
0.0108
0.0100
0.0033
0.0034
Pb
(mm)
0.0116
0.0095
0.0087
0.0027
0.0028
NaI
(mm)
0.0218
0.014
0.011
0.0029
0.0025
ments) this table can be corrected. Figure 5.6 shows range differences ofp, d, t, 3He and 4He for carbon. All of the range differences in Table 5.1 (except for carbon) are given in mm. The density of carbon, however, depends on its manufacture and it is recommended that the density be measured for the sample used before converting the range differences in mg/cm 2 to range differences in mm. The following equations are frequently useful for obtaining ranges for particles in terms of proton ranges:
125
Measurements of Radioactivi O,
5~176 I tJ~d
200~ 100
3He
50
2o z
10
II 1
I
I
I
I
I
I
I
I
1
I
I
I
I
2
5
10
2
5
102
2
5
103
2
5
10 '
2
RANGE DIFFERENCE / 1 MeV mg/cm 2 Fig. 5.6. Range differences for carbon of 1 MeV particles.
2Rp (89E)
(5.20a)
R, (E) = 3Rp (89E)
(5.20b)
R~(E)--Rp ( 88 + 0.25 mg/cm 2
(5.20c)
RHe3 (E)
(5.20d)
R e ( e ) --
= ~R~(3E) 3
where the subscripts p, d, t, t~, and He 3 refer to protons, deuterons, tritons, or-particles, and He 3 ions, respectively. Of course, when range differences are considered, the approximate additive factor for helium ion ranges does not contribute. When considering the range of charged particles it is useful to look at the property of the beam of charged particles. Figure 5.7 shows the range curve for a beam of particles penetrating to a given depth.
5.1.1.3 Rutherford scattering In addition to interacting with atomic electrons, charged particles interact with nuclei when passing through matter. When approaching the nucleus the charged particle feels a potential
Chapter5
126
Fig. 5.7. Range curve showing the number of particles in a beam penetrating to a given depth.
ZiZte2
U(r) - ~ + r
~(( + 1)]~2
2mr 2
(5.21)
The first term is due to the Coulomb repulsion, while the second term is the result of angular momentum of relative motion. The approaching charged particle will undergo scattering on the above potential. As early as 1906, Rutherford detected anomalously large alpha particle scattering by thin sheets of mica, gold and some other materials. He has explained this scattering in one of the most cited papers ever (Rutherford, 1911). Rutherford's most significant assumption was that the scattering centre of the target atom was the atomic charge concentrated into a nucleus o f - 10-~ cm. Assuming a Coulomb potential (U(r) = ZiZ, e2/r) between the alpha particle and the target nuclei, the differential scattering cross section is classically derived as:
dtJ(ZiZte2) dr=
2--~
1 sin4(~
(5.22)
where Zie is the electronic charge of the impinging particle (in this case an alpha particle), Zte is that of the target nucleus,/~ is the reduced mass of the two particles, v is the velocity of the centre of mass, and 0 is the centre-of-mass scattering angle. 5.1.2 Electrons
The interaction of electrons with matter is similar to the interaction of heavy particles, with the following differences: 1. Nuclear collisions are not part of the interaction because of the very light electron mass. 2. At energies higher than a few MeV, radioactive processes (bremsstrahlung) must be considered in addition to the inelastic electron collision.
Measurements of Radioactivity
127
10' > 10~ Ge x
1
10 2 10~
-
I
10'
I
102
,I
103
......
104
RANGE / l.tm Fig. 5.8. Zero transmission range vs. energy for electrons in Si and Ge.
3. Again because of their light mass, electrons are so intensely scattered that their trajectory in the material is a jagged line; therefore, the concept of range as previously used cannot be applied. Rather, the concept of zero-transmission range is introduced. This is done by means of absorption experiments, which permit definition of the absorber thickness resulting in zero-electron transmission at a given energy. Figure 5.8 shows the zero-transmission range as a function of energy in silicon and germanium.
5.1.3 Gamma and X-Rays The interaction of ionizing electromagnetic radiation with matter is different from the processes previously mentioned, and the concept of ranges and specific ionization loss cannot be applied. Only the three most important absorption processes are considered: the photoelectric effect, the Compton effect, and the pair-production effect. The corpuscular description of electromagnetic radiation is the most appropriate for these effects, as one photon in a well-collimated beam of N Ophotons disappears at each interaction. The attenuation of the photon beam can be described by a simple exponential law N = N Oexp (-gx)
(5.23)
where N is the remaining photons in the beam after traversing distance x, and the absorption coefficient Is is the sum of three terms due to the three above-mentioned processes. These processes are strongly dependent upon the energy of effect, such as Rayleigh scattering. Thomson scattering, and others are much less important and can be ignored in detection processes (see Fig. 5.9).
Chapter 5
128 1.0
(a)
0.8 Z ,~
0.6
~
0.4
ton
1.0 ~"~NN
photoelelctric
(b)
' p a i ~r" ~
0.8 0.6 0.4
-
0.20.01
0.1
1
lO
lO0
h~/MeV Fig. 5.9. Relative contributions of various photon interactions to the total attenuation coefficient for (a) carbon and (b) lead.
In the photoelectric interaction, the photon ejects a bound electron from an atom. All of the photon energy, hv, is given to the atom, which ejects the electron with an energy h v - E ~ , where E 1 is the binding energy of the electron. The excited atom then releases energy E 1by decaying to its ground state. In this process, the atom releases one or more photons (and possibly an electron, called an auger electron). The cross section of the photoelectric effect increases rapidly with the atomic number Z and decreases with increasing energy. The Compton effect is essentially an elastic collision between a photon and an electron; during this interaction, the photon gives a fraction of its energy to the electrons, and its frequency v is therefore decreased. The cross section for this effect decreases with increasing energy, but the decrease is less rapid than for the photoelectric effect. The energy of electron and the scattered photon are given by: Ey = E0[ 1 + (E0/mc 2) (1 - c o s 0 ) ] -~
E 0=E 0-Ev=E
(E ~ ] 0 I+(E 0/mc2)(1-cos0)
(5.24)
(5.25)
where E 0 is the incident photon energy; Ey is the scattered photon energy; E e is the electron energy; m = electron mass; c = velocity of light; and 0 = angle between incident and scattered gamma ray directions.
Measurements of Radioactivity
Z 9
129
E~0.5 MeV
b-. r.~ r.g3 o
Ev--1.0 MeV
Z
)
E~=2.0 MeV
I 0.5
..... I ~ 1
1.0
1.5
....
2.0
ELECTRON ENERGY / MeV Fig. 5.10. Compton scattered electron energy distribution.
The maximum energy loss by the photon is for a head-on collision (0 = 180 ~ and is equal to: E o -Ev =
Eo 1 + mc 2 /
2E ~
(5.26)
The fractional energy loss of low-energy photons is small since the scattering is nearly elastic, but becomes appreciable at higher energies. The probability of scattering to a particular angle is a complicated function of energy and angle, but can be generally described as becoming increasingly peaked at small angles as the photon energy increases. The total cross section depends upon the number of electrons available, or as the atomic number Z of the material. The energy distribution of Compton electrons for several gamma-ray energies is shown in Fig. 5.10. In the pair-production effect, a high-energy photon near a nucleus gives up its energy to produce an electron-positron pair. The photon energy goes into the rest-mass energy and the kinetic energy of the electron-positron pair. The minimum energy necessary for this effect is set by elementary relativistic considerations at the value of 1.022 MeV, an amount equivalent to two electron rest masses. The cross section P for pair production increases with energy. Up to energies of 10 MeV, the P/Z ratio remains constant with energy. At higher energies the cross section starts to decrease. Figure 5.11 summarizes values of the linear absorption coefficients of the above-mentioned effects as a function of gamma-ray energy for silicon and germanium.
Chapter 5
130
101 m
-E Ge(PE) [-. Z
m m
10 ~ Z
c) ~d
9 Z 9 [..
-
e(PP) =
10-' = m
=
m m
/
x
9 <
= Si(PP)
-
10 .2 =
z
<
Z
Z
10
-3
_
10 .2
10-'
10 ~
10'
10 2
E N E R G Y / MeV
Fig. 5 . 1 1 . Linear absorption coefficients vs. gamma-ray energy for Si and G e ( P E = p h o t o e l e c t r i c , C = Compton, P P = pair production).
In the pulse-height distributions of Compton interactions of y-rays in scintillation detectors there are two prominent features usually present: (1) the Compton edge, which corresponds to the maximum energy that can be transferred to an electron by the "/-ray, and (2) the backscatter peak, which corresponds to the absorption of a photon which has been scattered through 180 ~ in the material surrounding the detector. The energy of the Compton edge is given by: 22
Ec=E ~I(| + moC /
E~)
(5.27)
where Ey is the energy of the incident "f-ray. The energy of the backscatter peak is given by: 2
E b = E v - E c = moC2/(2 + moC/E~,).
(5.28)
The quantities E c and E b are shown in Fig. 5.12 and Fig. 5.13 as functions of E~ for both low- (<0.7 MeV) and high-energy (0.5-4 MeV) regions. 5.1.3.1 A b s o r p t i o n coefficients
The probability of interacting with matter in one of these three processes can be expressed as a cross section or as an absorption coefficient. The absorption coefficient contains the cross section and is therefore more practical in calculating absorption fractions (Hubbell, 1982).
Measurements of Radioactivity >
131 0,5
0,3 0.2 Back-scatter peak
0,1 0
0,07
/
0,05 0.03
Compton edge
0.02
Z
0,01 0
0,1
I 0,2
I 0.3
I 0,4
I 0.5
I 0.6
i 0,7
INCIDENT y-RAY ENERGY / MeV Fig. 5.12. Energy of Compton edge and backscattered peak for gammas of energy <0.7 MeV. ;>
A~rd
9t~
1.0"
0.7
z
0.5 i
~ 0r..) ~
0.3-
Z
Compton edge
2--
back-scatter peak
0.2-
0.1 0.5
1.0
2
3
INCIDENT ),-RAYENERGY / MeV Fig. 5.13. Energy of Compton edge and backscattered peak for high energy gamma rays (0.5-4 MeV).
The attenuation coefficient for a beam of gamma rays is related to the number of gamma rays removed from the beam, either by absorption or scattering. For the Compton effect, the absorption cross section is determined by the energy absorbed by the electron, which is the total collision energy minus the average scattered photon
132
Chapter 5
energy. For all three processes, the total attenuation coefficient/t is the sum of the three partial attenuation coefficients:
(5.29)
~total = ~photoelectric "{" ~Compton -t" ~pair production
The attenuation coefficients themselves are defined in terms of thickness of material or surface weight of material. This is just using a thickness x (cm) or a surface weight p x (g/cm 2) where 9 is the density (in g/cm3). The number of primary photons n removed from a beam of n photons is dn =
n
- ~tclx
(5.30)
which integrates to:
(5.3~)
n = no e-~'a~
for an initial beam intensity n 0. For surface density px the equation is:
-It
n = noe~
(5.32)
(px) 10"
Iodine K edge
101
NaI
10~ ca)
E
Total
10-' m. 10.2 [-
Photoelectric
Pair Production
10.3
| 0 .4
0.1
1.0
-i 10.0
GAMMA RAY ENERGY / MeV Fig. 5.14. Absorption coefficients for NaI.
Measurements of Radioactivity
133
L edges 102
101
K edge
e~t) eq
E ::l.
Lead
1 00
Copper 10-I
10-2
I 10 2
.....
I 10
__1 2
10 2
GAMMA RAY ENERGY / MeV Fig. 5.15. Absorption coefficients for Pb and Cu.
tx is known as the linear attenuation coefficient and ~ p as the mass attenuation coefficient. Graphs of the values for NaI are shown in Fig. 5.14, showing the total Wp value and Compton, photoelectric, and pair production components. Figure 5.15 shows total ~ p values for lead and copper. The K and L edges in the photoelectric absorption are at the energy where the photon can eject a K or L shell electron, thus providing an additional absorption mechanism. 5.1.4 Neutrons The interaction of neutrons with matter is quite different from that of either charged particles or gamma rays. Depending on their energy, neutrons interact with matter by various processes. 1. Elastic scattering: The neutron shares its initial kinetic energy with the nucleus, which suffers recoil only and is not left in an excited state. The smaller the mass of the nucleus, the greater the fraction of the kinetic energy taken by it. The average fraction of the neutron energy transferred per collision to a medium of atomic
134
Chapter 5
weight A is given by 2A/(I+A) 2. A 2-MeV neutron gets thermalized in about 18 collisions in water and in about 420 collisions in lead. 2. Inelastic scattering: Inelastic scattering is possible only with fast neutrons: the scattered neutron carries less energy than the incident neutron and the nucleus goes into an excited state. The excited nucleus either emits a gamma ray or remains in a metastable state. 3. Capture: The incident neutron is captured by the target nucleus forming a compound nucleus which may be excited and emit gamma radiation. This reaction is probably the most common, since thermal neutrons can induce this reaction in nearly all nuclides. The excitation energy of the target nucleus may be emitted in a single photon or in several. Every such capture results in energy emission amounting to about 6 to 10 MeV. Hence, materials in which neutron capture is allowed to take place for purposes of attenuation are so chosen that, as a result of the capture, charged particles or photons are emitted that can be easily absorbed. Cadmium and boron are commonly used for capturing thermal neutrons. 4. Particle emission: In this type of reaction the interaction of the incident neutron with the target nucleus may lead to the emission of particles such as protons, neutrons and alphas. Since the charged particles will have to overcome the Coulomb barrier before escaping the nucleus, this type of reaction is most probable for light nuclides and fast neutrons. 5. Fission: In this process the compound nucleus splits into two fission fragments with the emission of one or more neutrons. Fission reactions take place with thermal neutrons in 235U, 239pu and 233Uand with fast neutrons in many heavy nuclides. Essentially, the absorption of neutrons occurs in two distinct stages. Fast neutrons are slowed down by elastic and inelastic scattering processes with nuclei, particularly light nuclei like carbon and hydrogen. The slowed-down neutrons are then captured, as the capture cross-section for low-energy neutrons is high for most elements. It should be noted that neutron capture and certain nuclear reactions are the only interactions that can make the receiving medium radioactive, o~, 13, 7 and X-rays cannot make a medium radioactive. They can ionize a medium, but that does not make the medium radioactive. An ionized stable atom is not radioactive, because ionization alters only the electron structure of an atom, not the nuclear structure, which determines whether an atom is radioactive or stable.
5.1.5 Penetrating powers of ionizing radiations Ionizing radiation gives up some or all its energy for each interaction it undergoes. This means that radiation gradually loses its energy as it passes through a medium, and that at best, only some of the incident amount of radiation can pass entirely through any medium. The radiation that does not emerge from the medium is not trapped inside it--the radiation energy has been transferred to the medium, resulting in ionization and excitation. Some or all of the radiation in a beam entering a medium may cease to exist,
Measurements of Radioactivity
135
some may exit from the medium with diminished energy and some may emerge without having undergone any interactions, depending on the penetrating power of the specific radiation. In general, the penetrating power of ionizing radiation is determined by: 9 the type of radiation, 9 the energy of the radiation, 9 the medium the radiation passes through. The penetrating power of ct particles: 9 ct particles are the heaviest of the four most important ionizing radiations and they move the slowest. 9 An t~ has two positive charges. 9 An ct therefore has a high chance for interactions, rapidly loses its energy and so only travels for short distances, especially in dense media. 9 It is possible to stop ~ particles completelymafter having lost all its energy, an t~ captures two free electrons and becomes an ordinary stable 4He atom and is then no longer an ionizing particle. The penetrating power of [3 particles: 9 [3 particles are much lighter than t~ and travel much faster. 9 A [3 has only one negative charge. 9 A [3 therefore has a much smaller chance for interactions and loses its energy slower than an ct, and so travels longer distances in media. 9 It is possible to stop [3 particles completely--after having lost all its energy, a may be captured by a positive ion, becoming an ordinary orbital electron, and so ceases to be an ionizing particle. The penetrating power of X- and y radiation: 9 Photons travel at the speed of light and are not electrically charged. 9 A photon therefore has a much smaller chance for interactions than either o~or ~. 9 G a m m a rays give up their energy only a little at a time and it is almost impossible to stop all photons in a beam completely; the beam is only weakened or attenuated by a medium. The penetrating power of neutrons: 9 Neutrons have a very small chance for interactions, because they are not electrically charged. 9 Neutrons travel long distances through dense media and shorter distances in less dense media. Media containing lots of hydrogen atoms slow down neutrons the quickest. When a neutron collides with an atom that has about the same mass as a neutron (like a hydrogen atom), it loses the most energy (think of a marble colliding with another marble, as opposed to a marble colliding with a soccer ball). 9 Slower neutrons travel shorter distances in materials that have a greater chance of capturing (absorbing) them. 9 Eventually a neutron is either absorbed by a nucleus or it decays into a proton and an electron; these are eventually neutralized much like c~ and [3. It is very difficult to completely stop neutrons.
Chapter 5
136
Table 5.2 Ranges of ionizing radiations with (maximum energy of 1 MeV) Radiation
Electrical charge
Range in air
Range in tissue
ct
+2
6 mm
0.008 mm
13
-1
3m
4mm
3I
0
>500 m
>65 cm
n
0
>500 m
>65 cm
Data taken from Radiological Health Handbook (1970) and from Introduction to Health Physics, Cember (1983).
Alpha particles
J Beta particles
--0
~....0. ' ~ 9
--e
9
----p=e
Gamma rays
Paper
Aluminium
Lead
Fig. 5.16. Penetrating powers of ionizing radiations: a-particles can be stopped by paper, 13-particles can be stopped by aluminum, ),-radiation is weakened by lead, but is never greatly totally blocked; neutrons will pass through lead, but will be stopped by thick wax or concrete.
Fig. 5.17. Penetrating powers of ionizing radiation in tissue: y and n radiation penetrates easily; ]3-particles penetrate a few centimetres at max. o~-particles do not penetrate the dead layer of the skin.
The penetrating abilities of the various radiations are compared in Table 5.2 and Figs. 5.16 and 5.17. The ranges given in Table 5.2 are approximate, and are only valid for the specified energy, since radiation ranges are energy-dependent.
Measurements of Radioactivity
137
5.2 R A D I A T I O N D E T E C T O R S Ionizing radiation cannot be detected by any of the human senses. Nor can we perceive the cause of radiation, namely radioactivity ("glowing in the dark" is, unfortunately, a myth). Man has to use special instruments to detect radiation. These instruments are based on the interactions of ionizing radiation with matter, in particular 9 ionization in gases, and ~ ionization and excitation in certain solids. A variety of instruments based on these principles exists for the assessment of radiological hazards in the environment. The operating principles and practical applications of these instruments are discussed below. Measurement techniques often also require calculations using the instrument reading, in order to obtain the desired quantity.
5.2.1 Charged particle detection Charged particle detection is a process in which the particles interact directly with the material in the region of ionization, and the path of the particle is clearly defined. The efficiency for detecting a particle once it has entered the detector is nearly unity, so that efficiencies for standard geometries are generally calculated from the solid angle alone. For a small detector at sufficiently large distance from a point source (source-todetector distance greater than about three times detector diameter), the solid angle is A]x 2 steradians, where A is the detector area and x is the source-to-detector distance. At closer distances the following formula should be used where O~ is the maximum angle subtended by a circular detector /21
solid angle =2rt I sin0d0 = 2rt(1 - cos0~)
(5.33)
0
For extended sources at close distances experimental calibration should be made. Many types of detectors, such as Geiger-Mtiller counters, proportional counters and scintillation detectors, are used for charged particle detection. The selection is made on the basis of resolution and range of particle in the gas or scintillator. In some cases, the particles are not completely stopped within the detector for an energy measurement, but deposit only a portion of their energy. This is related to the relative ionization of the particle and can be used to identify different kinds of particles. Semiconductor detectors are also used for charged particle detection, but differ from those used for gamma rays in being very thin and operating at room temperature. They have very good resolution and are used for total energy, or partial energy loss, known as stopping power or dE/dx measurements. Silicon-charged particle detectors have a p-i-n structure in which a depletion region is formed by applying reverse bias, with the resultant electric field collecting the electron-hole pairs produced by an incident-charged particle. The resistivity of the silicon must be high enough to allow a large enough depletion region at moderate bias
138
Chapter 5
voltages. A traditional example of this type of detector is the gold-silicon detector. In this detector, the n-type silicon has a Schottky barrier contact as the positive contact, and deposited aluminum is used at the back of the detector as the ohmic contact. A modem version of the charged particle detector is called PIPS, an acronym for Passivated Implanted Planar Silicon. This detector employs implanted rather than surface barrier contacts and is therefore more rugged and reliable than the Silicon Surface Barrier (SSB) detector it replaces. At the junction there is a repulsion of majority carrier (electrons in the n-type and holes in p-type) so that a depleted region exists. An applied reverse bias widens this depleted region which is the sensitive detector volume, and can be extended to the limit of breakdown voltage. Detectors are generally available with depletion depths of 100 and 1000 gm, with the cost approximately proportional to the depletion depth. Detectors are specified in terms of surface area and alpha or beta particle resolution as well as depletion depth. The resolution depends largely upon detector size, being best for small area detectors. Alpha resolution of 12 to 35 keV and beta resolution of 6 to 30 keV are typical. Areas of 25 to 3000 mm 2 are available as standard, with larger detectors available in various geometries for custom applications. Additionally, PIPS detectors are available with fully depleted silicon wafers so that a dE/dx energy loss measurement can be made, with the particle exiting into another detector to measure the remaining energy.
5.2.2 Gamma and X-ray detection The kinds of detectors commonly used for gamma and X-ray detection can be categorized as: a. Gas-filled detectors, b. Scintillation detectors, c. Semiconductor detectors. Gas-filled detectors are used for X-rays or low energy gamma rays. These include ionization chambers, proportional counters and Geiger-Mtiller counters. Scintillation detectors are used in conjunction with a photomultiplier tube to convert the scintillation light pulse into an electric pulse. Solid crystal scintillators such as CsI or NaI are commonly used, as well as plastics and various liquids. Semiconductor detectors, made from single crystals of very pure germanium or silicon, are the highest performance detector type. The superior resolution of these detectors has revolutionized data-gathering for X-ray and gamma-ray measurements. The comparison of the pulse resolving ability of the three types of X-ray detectors: scintillator, gas proportional and Si(Li) is shown in Fig. 5.18. The choice of a particular detector type for an application depends upon the gamma energy range of interest and the application's resolution and efficiency requirements. The detector must have sufficient material to absorb a large fraction of the gamma ray energy. Thus, a gas-filled proportional counter is suitable for 14.4 keV gamma rays or for X-rays, but would not "see" 1 MeV gamma rays because the probability of
139
Measurements of Radioactivity
Gas proportional
....
I
100
260
3()0
CHANNEL NUMBER Fig. 5.18. The pulse resolving ability of three types of X-ray detection: scintillator, gas proportional and Si(Li).
absorbing the gamma ray energy is too low. Further, the higher gamma ray energies are more effectively absorbed by higher Z materials. Other considerations are count-rate capability, resolution, and if timing applications are involved, pulse rise time. The efficiency of a detector is a measure of how many pulses occur for a given number of gamma rays. Various kinds of efficiency definitions are in common use for gamma ray detectors: 1. Absolute efficiency: The ratio of the number of counts produced by the detector to the number of gamma rays emitted by the source (in all directions). 2. Intrinsic efficiency: The ratio of the number of pulses produced by the detector to the number of gamma rays striking the detector. 3. Relative efficiency: Efficiency of one detector relative to another; commonly that of a germanium detector relative to a 3 in diameter by 3 in long NaI crystal, each at 25 cm from a point source, and specified at 1.33 MeV only. 4. Full-energy peak (or photopeak) efficiency: The efficiency for producing fullenergy peak pulses only, rather than a pulse of any size for the gamma ray. An example of a full-energy peak efficiency curve for a germanium detector is shown in Fig. 5.19.
5.2.2.1 Gas-filled detectors A gas-filled detector is basically a metal chamber filled with gas and containing a positively biased anode wire. A photon passing through the gas produces free electrons and positive ions by the interactions previously described. The electrons are attracted to the anode wire and collected to produce an electric pulse (see Fig. 5.20).
140
Chapter 5 10 .3
8~
57Co
t~
6.
4 --
2O3Hg 113Sn ~N~
2 G) Z
10" ~
85Sr 137Cs 88y
6
~~
4
88y
2 m 10"
I I II lliltliltl
1 Ii III1
4 6 4 6 8 100 2 ENERGY / k e V
Fig. 5.19. Efficiency
1000
I ! i
2
calibration for Ge detector.
!!i.Positiveelectrode
9 Negative electrode
Fig. 5.20. Basic
" """
Meter
I
[ _ ,i,l_
Fill gas
Battery
Y
elements of gas-filled detector.
At low anode voltages, the electrons may recombine with the ions. Recombination may also occur for a high density of ions. At a sufficiently high voltage nearly all electrons are collected, and the detector is known as an ionization chamber. At higher voltages the electrons are accelerated toward the anode at energies high enough to ionize other atoms, thus creating a larger number of electrons. The detector is known as a proportional counter. At higher voltages the electron multiplication is even greater, and the number of electrons collected is independent of the initial ionization. This detector is the Geiger-Miiller counter, in which the large output pulse is the same for all photons. At still higher voltages continuous discharge occurs. The different voltage regions are indicated schematically in Fig. 5.21. The actual voltages can vary widely from one detector to the next, depending upon the detector geometry and the gas type and pressure. The ionization chamber is the simplest form of gas-filled detector. It consists of a chamber provided with two electrodes coupled to an electric potential. Gas ions created by the radiation are attracted to the respective electrodes of opposite charge, causing an
Measurements o f Radioactivi~. ,
141
1012
elSje O rlrge 7
10 l~ - -
Mueller
Continuous
/
10 8
106 _
/
Ionization
104
102
L..----''~. Proportional ~/~ , ~l-----i c~ -"
I
I
I
I
[
I
I
I
500
I
'
.
I
I ,,1
1000
ANODE VOLTAGE / V Fig. 5.21. Gas detector output vs. anode voltage.
electric current (called the ion current) to flow between the electrodes. The ion current is electrically amplified and is measured with a micro-ampmeter calibrated to read in dose rate units. The electric current is related to the dose rate as follows: the electric current is proportional to the ion current, which is a measure of the rate of ionization in the gas filled chamber, which in turn is a measure of the rate at which the gas in the chamber absorbs energy from the radiation passing through it, i.e. of the dose rate. Ions created inside the chamber walls may enter the chamber causing an incorrect reading. This can be avoided by manufacturing the walls from a material with similar ionization properties to the gas. For an air-filled chamber, such chamber walls are called air-equivalent walls or simply air walls. If the detector is required to respond to beta radiation, the chamber must have thin walls or a thin window to allow the betas to enter (normally structural materials will stop the betas before they can enter the detector chamber). The working principle of an ionization chamber is illustrated in Fig. 5.22. Ionizing particle Amplifier
mA meter Gas
" Fig. 5.22. W o r k i n g principle o f an ionization chamber.
142
Chapter 5
High voltage
iFig. 5.23. Gas amplification.
Proportional counter: If the voltage applied to a gas-filled detector is increased beyond a certain point, it ceases to function like an ionization chamber, and an effect known as gas amplification occurs. Gas amplification occurs when electrons produced by the radiation through ionization of the gas in the chamber are accelerated by the applied voltage to such an extent that they gain enough kinetic energy to cause further ionization, resulting in a cascade of ionization that causes an electric pulse. The original amount of ionization caused by the radiation is thus amplified to create a much larger electric signal. The principle of gas amplification is illustrated in Fig. 5.23. If the applied voltage is not too high, the size of the output pulse is proportional to the amount of energy deposited in the detector by the incoming radiation particle/ photon. (This is why the detector is called a proportional counter.) It is called a counter because the number of output pulses are counted by a counting system. The read-out can be either a total number of counts or a count rate (in cpm or cps). The count rate is a measure of the rate at which individual radiation particles/ photons cause pulses in the detector. For this reason, proportional counters can be used to determine the amount of radioactivity in a sample. The proportionality of the detector can be used to distinguish between different radiation energies of types of radiation, based on pulse size. A specific type of proportional counter that is used for accurate counting of ~ and [3 activity on smear samples, is the gas flow proportional counter. In some types of gas flow proportional counters, the sample is put inside the detector for greater sensit i v i t y - t h e r e is no structural material to absorb the radiation before it can be detected. Figure 5.24 shows a diagram of the detector of such a gas flow proportional counter. Other gas flow proportional counters have very thin Mylar windows sealing off the detector's gas chamber. During counting, samples are positioned very close to the window. Geiger-Miiller counter: If a very high electric potential is applied across the electrodes of a gas-filled detector, the gas amplification achieves a maximum value;
Measurements of Radioactivity
143
Fig. 5.24. Diagram of windowless gas flow proportional counter used for oc and 13counting. The sample is inserted into the detector on a sliding tray.
this means that all pulses are amplified to the same maximum pulse size. All output pulses then have the same size and are no longer proportional to the energy deposited by radiation; an output pulse is only a sign of some radiation particle/photon having been detected--it cannot be said what type or energy. This type of detector is called a Geiger-Mtiller counter (also generally known as a Geiger counter or GM tube). A GM tube is usually of the general construction shown in Fig. 5.20. If the detector is to be sensitive to 13radiation, it is fitted with a thin end-window (see Fig. 5.25) that allows the [~ particles to penetrate into the detector's sensitive volume. GM tubes are widely used because they are quite sensitive to radiation, but simple and rugged in construction. Gas-filled detectors do not respond instantaneously; their meters take some time to reach the eventual reading. This is caused by the time it takes to collect all the electric charges and the characteristics of the electric circuit. The time constant is used to quantify the response time. The time constant is the time an instrument takes to indicate 63 % of its eventual reading. An instrument with a large time constant responds slowly, and one with a small time constant responds quickly. The significance of the response time is that in practice one must wait for the instrument reading to reach its full response and stabilize before taking a reading; this is especially important for portable instruments. A rule of thumb is that one must wait for at least three times the time constant before taking a precise reading. A time constant for an ionization chamber is in the order of 10 s; for Geiger counters it can vary from a few seconds to more than 20 s.
Fig. 5.25. A typical thin end-window GM tube for [3 detection.
144
Chapter 5
The resolving or dead time of a detector is the minimum period of time that must elapse after the detector has detected a particle/photon, before it is able to detect the next particle/photon. Dead time is caused by the time the detector takes to collect all the charges created by an ionization avalanche, and recover for the start of the next impulse. While collecting the charge from one pulse, the detector is "dead", i.e. it cannot register another pulse. A particle/photon that arrives in the detection volume while the detector is still "dead" will not be detected. The higher the activity/dose rate, the more particles/photons will be "missed". In other words, detector dead time causes the instrument to increasingly under-respond at high dose rates/source activities. Dead time is particularly significant for Geiger counters; they have the largest dead times (up to 200 ~ts) because they have the most charge to collect (maximum gas multiplication, maximum size ionization avalanche), which takes time. At very high dose rates, a Geiger counter with large dead time may become completely "paralyzed"; its reading drops fight down and stays there, even if the dose rate or activity is further increased. Geiger counters can be designed to electronically compensate for counts lost due to dead time. However, many are not. It is important to know whether a particular Geiger counter is dead time compensated, so that if not, its drawback can be kept in mind. 5.2.2.2 Scintillation detector
Scintillation means the production of small flashes of light. Some crystals, e.g. sodium iodide (NaI) convert the ionization and excitation produced by radiation into a light pulse or scintillation. The amount of light that is produced is proportional to the energy deposited by the radiation particle/photon. The small light pulse is converted into an electric pulse by an electric component called a photomultiplier tube. The size of the amplified electric pulse is proportional to the energy deposited by the radiation photon/particle (for details see Birks (1964) and Knoll (1979)). The basic components and operating principle of a scintillation detector are illustrated in Fig. 5.26. The proportionality of the output pulses of a scintillation counter to the deposited radiation energy enables the output to be used 9 to discriminate between different types and energies of radiation, e.g. with portable instruments, and 9 as input pulses for a gamma spectrometer. There are also liquid scintillation counters that use a scintillation liquid instead of a crystal. The sample to be counted is dissolved in the scintillating liquid. This method is suitable for t~ and ]3 counting. The properties of a scintillation material required for good detectors are transparency, availability in large size, and large light output proportional to gamma ray energy. Relatively few materials have good properties for detectors. Thallium activated NaI and CsI crystals are commonly used, as well as a wide variety of plastics. Both NaI and CsI require an activator such as Thallium for proper operation. NaI is the dominant
Measurements of Radioactivity
145 Gamma ray in ]
,i
[H~ inI ISi a
~ IIreamp owor supi ery
Fig. 5.26. Basic components and operating principle of a scintillator detector.
material for gamma detection because it provides good gamma ray resolution and is economical. However, plastics have much faster pulse light decay and find use in timing applications, even though they often offer little or no energy resolution. The actual process by which light is produced is very complex. The high Z of iodine in NaI gives good efficiency for gamma ray detection. Energy dependence of absorption coefficient for NaI in show in Fig. 5.14. A small amount of T1 is added in order to activate the crystal, so that the designation is usually NaI(T1) for the crystal. The best resolution achievable is about 8.0% for the 662 keV gamma ray from 137Csfor 2 in diameter by 2 in long crystal, and is slightly better for larger sizes. Typical
Chapter 5
146 1200-800 0.662MeV 400
k..
o 9
20
1200 -
800
400
0
-
40
60
80
'
1~)0 '
120
'
1'~0
0.511MeV 9
22
--
l
-
/
eV
Na
..o
20
40
1.27iM ,-~.,,.
.
60
80
100
120
140
160
CHANNEL NUMBER Fig. 5.27. NaI(T1) spectra for '37Cs and 22Na gamma sources.
spectra are shown in Fig. 5.27 for 137Cs and 22Na gamma sources. NaI is slightly non-linear (about 5%) at low energies (below 200 keV) because of light output variations with gamma energy. The light decay time constant in NaI is about 0.25 ~ts. Typical charge sensitive preamplifiers translate this into an output pulse rise time of about 0.5 ~ts. Fast coincidence measurements cannot achieve the very short resolving times that are possible with plastic, especially at low gamma ray energies. Many configurations of NaI detectors are commercially available, ranging from very thin crystals for X-ray measurements to large crystals with multiple phototubes. Crystals built with a well to allow nearly spherical (4~) geometry counting of weak samples are also a standard configuration. Many types of plastic scintillators are commercially available and find applications in fast timing, charged particle or neutron detection, as well as in cases where the rugged nature of the plastic (compared to NaI), or very large detector sizes, are appropriate. Sub-nanosecond rise times are achieved with plastic detectors coupled to fast photomultiplier tubes, and these assemblies are ideal for fast timing work. 5.2.2.3 Semiconductor detector
Semiconductors are materials that do not normally conduct electricity because their crystals do not contain enough free charged particles to carry the current, but that do become conducting when atoms in the crystal become ionized (Knoll, 1989; Bertolini and Coche, 1968; Goulding and Pene, 1974).
Measurements of Radioactivity
147
When a relatively small voltage (25-300 volt) is applied across the crystal, and it is exposed to ionizing radiation, the electric field sweeps the free charged particles formed by the radiation out of the crystal. This creates an electric pulse in the external circuit. The size of the pulse is proportional to the radiation energy deposited in the semiconductor. The number of pulses per size range can be counted, and the count rate can be used to determine the activity of the radiation source. Like scintillation detectors, semiconductor detectors are usually used in gamma spectrometer set-ups to identify radionuclides and determine their activities in a sample. A semiconductor detector is much more expensive and somewhat more troublesome to operate than a scintillation detector, but it can distinguish much better between different radiation energies and is better for nuclide identification. The group IV elements silicon and germanium are by far the most widely used semiconductors, although some compound semiconductor materials are finding use in special applications as development work on them continues. Semiconductor detectors have a P-I-N diode structure in which the intrinsic (I) region is created by depletion of charge carriers when a reverse bias is applied across the diode. When photons interact within the depletion region, charge carriers (holes and electrons) are freed and are swept to their respective collecting electrode by the electric field. The resultant charge is integrated by a charge sensitive preamplifier and converted to a voltage pulse with amplitude proportional to the original photon energy. Since the depletion depth is inversely proportional to net electrical impurity concentration, and since counting efficiency is also dependent on the purity of the material, large volumes of very pure material are needed to ensure high counting efficiency for high energy photons. Prior to the mid-1970s, the required purity levels of Si and Ge could be achieved only by counter-doping P-type crystals with the N-type impurity, lithium, in a process known as lithium-ion drifting. Although this process is still widely used in the production of Si(Li) X-ray detectors, it is no longer required for germanium detectors since sufficiently pure crystals have been available since 1976. The band gap figures in Table 5.3 signify the temperature sensitivity of the materials and the practical ways in which these materials can be used as detectors. Just as Ge transistors have much lower maximum operating temperatures than Si devices, so do Ge detectors. As a practical matter both Ge and Si photon detectors must be cooled in order to reduce the thermal charge carrier generation (noise) to an acceptable level. This requirement is quite aside from the lithium precipitation problem which made the old Ge(Li), and to some degree Si(Li) detectors, perishable at room temperature. The most common medium for detector cooling is liquid nitrogen, however, recent advances in electrical cooling systems have made electrically refrigerated cryostats a viable alternative for many detector applications. In liquid nitrogen (LN 2) cooled detectors, the detector element (and in some cases preamplifier components), are housed in a clean vacuum chamber which is attached to or inserted in a LN 2 Dewar (see Fig. 5.28 as an illustration of the often used
Chapter 5
148
Table 5.3 Shows some of the key characteristics of various semiconductors as detector materials Element vs. Band Gap Material
Z
Band gap (eV)
Energy/e-h pair (eV)
Si
14
1.12
3.61
Ge
32
0.74
2.98
CdTe
48-52
1.47
4.43
Hgl 2
80-53
2.13
6.5
GaAs
31-33
1.43
5.2
configuration). The detector is in thermal contact with the liquid nitrogen which cools it to around 77~ At these temperatures, reverse leakage currents are in the range of 10-9 to 10-~2 amperes. In electrically refrigerated detectors, both closed-cycle freon and helium refrigeration systems have been developed to eliminate the need for liquid nitrogen. Besides the obvious advantage of being able to operate where liquid nitrogen is unavailable or supply is uncertain, refrigerated detectors are ideal for applications requiring long-term unattended operation, or applications such as undersea operation, where it is impractical to vent LN 2 gas from a conventional cryostat to its surroundings.
Fig. 5.28. Vertical dipstick cryostat.
Measurements of Radioactivity
149
There are various types and configurations of semiconductor detectors. One type, the surface barrier detector, is especially useful for alpha spectrometry. This type of detector uses a semiconductor wafer, and the sensitive volume of the detector is the surface layer of the wafer. It detects a radiation very effectively, because the ~ particles are not absorbed before they reach the sensitive volume (the usual problem with other detectors). The energy lost by ionizing radiation in semiconductor detectors ultimately results in the creation of electron-hole pairs. Details of the processes through which incoming radiation creates electron-hole pairs are not well known, but the average energy necessary to create an electron-hole pair in a given semiconductor at a given temperature is independent of the type and the energy of the ionizing radiation. The values of are: 3.62 eV in silicon at room temperature; 3.72 eV in silicon at 80 K, and 2.95 eV in germanium at 80 K. Since the forbidden bandgap value is 1.115 eV for silicon at room temperature and is 0.73 eV for germanium at 80 K, it is clear that not all the energy of the ionizing radiation is spent in breaking covalent bonds. Some of it is ultimately released to the lattice in the form of phonons. The importance of these interactions for detectors is in how they relate to incident gamma ray energy deposited in the crystal. An ideal detector converts all of the energy of the gamma ray into an electric pulse that is directly proportional to the gamma ray energy, i.e.: linear. For gamma rays the Compton scattering often results in only a fraction of the energy being deposited because the gamma ray can scatter and then escape from the crystal without further interaction. The full-energy peak can be produced by a photoelectric absorption, or one or more Compton scatterings followed by photoelectric absorption. If pair production occurs, the positron slows down in the material and then annihilates, producing two 511 keV gamma rays. Each of these may escape from the detector totally, or leave part of their energy by Compton scattering. If one or both totally escapes, the deposited energy is the full energy minus 511, or 1022 keV, leading to designation of these peaks as "single escape" and "double escape" peaks. An idealized spectrum from a source of gamma rays at 2 MeV is shown in Fig. 5.29. The constant value of E for different types of radiation and for different energies contributes to the versatility and flexibility of semiconductor detectors for use in nuclear spectroscopy. The low value of ~ compared with the average energy necessary to create an electron-ion pair in a gas (typically 15 to 30 eV) results in the superior spectroscopic performance of semiconductor detectors. If all of the energy lost by ionizing radiation in a semiconductor were spent breaking covalent bonds in the detector's sensitive volume, no fluctuations would occur in the number of electron-hole pairs produced by ionizing radiation of a given energy. At the other extreme, if that energy entering the semiconductor detector that is partitioned between breaking covalent bonds and lattice vibrations or phonon production were completely uncorrelated, Poisson statistics would apply. The variance in the number of electron-hole pairs n would then be
2 = n. In fact, neither of these suppositions simulates reality. As the incoming ionizing radiation gives
Chapter 5
150
Double Escape Peak
Full Energy Peak Single Escape
,,
0.5
1.5
2.0
DETECTED ENERGY /MeV Fig. 5.29. Idealized gamma ray spectrum.
up energy, a large shower of hot electrons is created. After many generations, the energy of these hot electrons gets close to the ionization energy necessary to create an electron-hole pair in the semiconductor detector, so that there are several possible competing mechanisms for energy loss. Thus the Fano factor F is introduced to modify the more familiar Poisson relation for this case. The equation for the variance can be written as 2
=
F < n > 2 = Ft/.
In the case where there are no fluctuations in the number of electron-hole pairs, F would be zero; in the case where Poisson statistics apply, F would be equal to 1. Since the energy necessary to create electron-hole pairs in semiconductor detectors is much smaller than that of the incoming ionizing radiation, it can be concluded that F is closer to zero than to 1. The true value of F for silicon and germanium is still unknown; the conflicting theories on the subject do not lead to experimentally distinguishable results. However, by assuming a value of 0.1 for F in both silicon and germanium, satisfactory agreement with measured results is found in most cases. By assuming a value of 0.1 for the Fano factor, the following formula gives the germanium detector resolution at LN 2 temperature: AE = 1.27x/E-
(5.34)
with E measured in eV. AE must be summed in quadrature with the FWHM keV noise AN in order to obtain the measured energy resolution AEs:
aes-
+(AN) 2
(5.35)
The value of K for silicon at room temperature is of little interest because, in such conditions, other factors than fundamental statistics dominate energy resolution values.
Measurements of Radioactivi~. ,
151
i
Z
A
I~ O
I(t) CD1..
RD
,w
O
Fig. 5.30. Equivalent circuit of semiconductor detector: I(t) = current generator; Co = capacitance of the depletion region; Z = series impedance; RDis the resistance of the depletion region. These simple formulas show that, as expected from the better statistics due to the lower value of ~, when the energy resolution is dominated by the detector contribution, germanium detectors have an advantage over silicon detectors. The equivalent circuit of a semiconductor detector operated as a spectrometer is shown in Fig. 5.30. In most cases, effects of high resistance of the reverse-biased junction are negligible. If a zero-electric-field radiation-insensitive region is present in the detector, its impedance (a parallel RC combination) appears in series with the circuit and is indicated in Fig. 5.30 by the impedance Z. The impedance also accounts for any resistance (or resistance-capacitance combination) appearing in series with the contacts. When semiconductor detectors are used as spectrometers, they are invariably connected to a charge-sensitive (integrating) preamplifier with a high dynamic input capacitance. The charge-sensitive preamplifier integrates on its feedback capacitance the current signal delivered by the detector and feeds the resulting voltage signal to the filter amplifier (main amplifier). The time behaviour of the current signal at the input of the charge-sensitive preamplifier is determined by the current signal' s shape and by the effect of the equivalent circuit shown in Fig. 5.30. The effect of the equivalent circuit is usually either negligible or easily calculated, whereas detailed considerations on the charge collection process in the detector are needed to calculate the induced current signal l(t). The current delivered by the signal generator l(t) is induced on the contacts of the detector by the motion of the charge carriers created by the ionizing radiation. Therefore, the first problem in determining l(t) is calculating the motion of the charge carriers in the detector's electric field. When this problem is solved, the induced charge can be calculated by electrostatic considerations. The charge carriers created by the ionizing radiation drift to the contacts of opposite polarity, following the lines of force of the electric field established by the applied voltage. In the case of heavily ionizing particles such as fission fragments, the drift process does not begin immediately due to the creation of the charge cloud. The electric field E(r) in the detector can be calculated from known quantities: applied bias voltage, detector geometry, and resistivity of the bulk material. Once the
152
Chapter 5
electric field is known, the motion of a charge carrier created at a given point r 0 or the detector volume can be calculated by using the values for the drift velocity Vd as a function of the electric field E. Thus the differential equation dF
- - = V d [E(r)]
dt
(5.36)
can be written for every charge carrier and can be solved if the initial positions r 0 are known only when the charge carriers are created along a well-defined track (heavy charged particles). In the case of beta, x, and gamma radiation, the only information on r 0 values is of a statistical nature. The integration of eq. (5.36) leads to r(t) for every charge carrier created. The charge induced by every carrier can then be calculated by electrostatic considerations. For instance, in the case of a detector with parallel contacts and a field E ( x ) across a distance W, the charge induced by a carrier q moving along a length Ax in the direction of the field is given by Ax Aq = q ~ W
(5.37)
independently of the shape of E ( x ) . Equations (5.36) and (5.37) (or the appropriate induction equation) yield the contribution to l ( t ) of every single charge carrier and, by integration over all the created charge carriers, the total l(t) function. The rise time Tt of the pulse generated by a semiconductor detector can be measured at the output of a charge-sensitive preamplifier. If the preamplifier is sufficiency fast, T, is determined by the following factors: 1. The charge collection time TR, 2. The rise time of the detector equivalent circuit, in most cases a negligible quantity, 3. The plasma time In most cases T R is the dominant factor. Although a precise calculation of TR can be quite complex, the order of magnitude of TR can be easily obtained by the following formulas: T R = W• 10-v s
(5.38a)
for silicon detectors at room temperature, and T R = Wx 10-8 s
(5.38b)
for germanium detectors at LN 2 temperature. In these formulas, W is the thickness of the depletion region measured in mm. For silicon detectors and for planar HPGe detectors, the value of W is provided with each detector. For coaxial Ge detectors, W is the radius of the cylinder. The formulas given above are indicative only of orders of magnitude and do not give exact values.
Measurements of Radioactivity
153
The previous discussion did not consider trapping effects, which result in a loss of charge to the collection process and consequent distortion of the shape of the peak as observed with a multichannel analyzer. Trapping of a charge carrier in a semiconductor occurs when the carrier is captured by an impurity or imperfection centre and is temporarily lost to any charge transport process. In semiconductor detectors, it is useful to introduce the quantity ~+ (mean free drift time): T~+-" ( N tcy gth) -I
(5.39)
where N, = density of trapping centres, c = trapping cross section, and V,h = thermal velocity. Note that x§ does not ordinarily coincide with the classical lifetime in photoconductivity theories. This is because in photoconductivity the traps are generally filled, while in a depleted detector, the traps are generally empty. The trapped charge carrier can be re-emitted in the relevant band and take part again in the charge transport process. The average time spent by a carrier in a trap is called the mean detrapping time "co and is strongly temperature dependent: o = C exp - - ~
(5.40)
where C = constant, E~ = activation energy of the trap, k = Boltzmann's constant, and T = absolute temperature. If the mean detrapping time is of the same order of magnitude as, or larger than, the electronic clipping-time constants, the charge carrier is lost to the charge collection process or is collected with significantly reduced efficiency. The result is poor energy resolution and peak tailing. On the other hand, if the mean detrapping time is orders of magnitude shorter than the charge collection time due to drift of the carriers, then the trap has no effect on the charge collection process. For this reason, normally used dopants such as Li, P, B, and Ga, which are shallow donors or acceptors, do not act as traps. It can be shown that, to first-order approximation, the efficiency of collection of a charge carrier subjected to trapping with a mean free drift time x+ is given by 11= 1 -
~
(5.41)
where 1"1is the collected fraction of the created charge. In a modern germanium gamma-ray spectrometer the charge collection efficiency is of the order of 0.999, and as TRis of the order of 10-v s, then x+, according to eq. (5.41), is of the order of 10-4 s. As typical values of V,h and ~ a r e 10 7 cm s-1 and 10 -13 cm 2 respectively, the maximum concentration of trapping centres permissible in the detector is of the order of 10 l~ cm -3 corresponding to approximately 1 for every 1012 atoms of germanium.
154
Chapter 5
(i) Germanium detector Germanium detectors are semiconductor diodes having a P-I-N structure in which the intrinsic (I) region is sensitive to ionizing radiation, particularly X-rays and gammarays. Under reverse bias, an electric field extends across the intrinsic or depleted region. When photons interact with the material within the depleted volume of a detector, charge carriers (holes and electrons) are produced and are swept by the electric field to the P and N electrodes. This charge, which is in proportion to the energy deposited in the detector by the incoming photon, is converted into a voltage pulse by an integral charge-sensitive preamplifier. Because germanium has relatively low band gap, these detectors must be cooled in order to reduce the thermal generation of charge carriers (thus reverse leakage current) to an acceptable level. Otherwise, leakage current induced noise destroys the energy resolution of the detector. Liquid nitrogen, which has a temperature of 77~ is the common cooling medium for such detectors. The detector is mounted in a vacuum chamber which is attached to or inserted into an LN 2 Dewar. The sensitive detector surfaces are thus protected from moisture and condensible contaminants. A modern Ge detector is a suitably shaped cylinder of highly purified germanium; it is rather hard to imagine that on 10 l~ atoms of germanium we have only one atom of an impurity. The modern metallurgical methods, zone refining, allow us to obtain such extreme purity of material. There are very few laboratories where germanium can be refined to this purity, and a single crystal can be grown; there are only three that produce such crystal for commercial uses. And even these three have difficulties in producing really big crystals. The technology of making a gamma ray detector from a piece of pure germanium is in principle simple but in practice very demanding. The Ge crystal must be 9 properly shaped, 9 equipped with suitable electrical contacts, 9 mounted in a cryostat (two important criteria: it must have excellent contact with the cold finger, and it should be mounted in such a way as to have minimum microphonics), 9 connected to the first stage of the preamplifier (which is usually placed inside the cryostat). After a good vacuum has been created in the cryostat, and the detector has been cooled by placing the assembly in liquid nitrogen, the characteristics of the detectors are measured. If the detector shows the resolution of 1.68 keV, it will obtain a high price tag. With the resolution of 2.2 keV, it will cost only half as much. And if it has resolution of 3 keV, it will be thrown away. The production of the detector is only to a smaller extent science. Mainly, it is good workmanship, applying modern technologies~and hoping for good results. There are different types of germanium detectors. Their geometry and construction features are illustrated in Fig. 5.31. The "classical" coaxial Ge detector is made of p-type germanium, and is used for spectroscopy of gamma rays. It covers the energy range from 100 keV to several MeV. On the low energy side, its efficiency is limited by the fact that low energy gamma rays
Measurements of RadioactiviD,
155
(a)
(c) I
In
(b)
....nl
(d)
II
Fig. 5.31. (a) p-type germanium detector; (b) n-type germanium detector; (c) planar germanium detector; (d) low energy germanium detector.
cannot penetrate the wall of the cryostat. High energy gamma rays might not be detected because they just pass the volume of the detector without creating a signal that would reflect all the energy of the ray. If the detector crystal is made of n-type germanium, the outer electrode can be made rather thin. Include a thin aluminium or a beryllium window at the face of the detector, so that low energy electromagnetic radiation will be able to enter the volume of the detector, and you get a gamma ray detector that is good also for soft gamma rays, and X-rays. Instead of a cylindrically shaped detector, one can take only a slice of the monocrystal. This will make a planar Ge detector. With a beryllium entrance window, it can be used for X-rays. The last addition to the world of Ge detectors is the Low Energy Germanium detector. If made by Canberra, it will be called LEGE. It is excellent for low energy gamma-rays; its energy range extends from 10 to 300 keV. This is indicated in Fig. 5.32.
Fig. 5.32. Gamma-ray detection energy interval for different types of detection.
Chapter 5
156
3 keV 9Z --
w~ 1 keV
..-1 9
XtRa
~
~
Lar
100 eV
I
J 10 5.9
106 122
13132
ENERGY Fig. 5.33. Typical resolution vs energy. 0.I
I
I I I IIII
I
I
I
I
I I I IIII
"I
I
I I I I
0.01 --
O.OOl lO
1O0
1000
10000
E N E R G Y / keV Fig. 5.34. Detection efficiency o f the H P G e detector in the Marinelli-beaker configuration.
The detector geometrics also result in different energy resolutions especially for lower y-ray energies. This is shown in Fig. 5.33. Typical absolute efficiency curves for various Ge detectors in the Marinelli-beaker configuration are shown in Fig. 5.34, while Fig. 5.35 shows typical absolute efficiency curves for various Ge detectors with 2.5 cm source to the end-cap spacing.
(ii) Silicon detector Occasionally, high resolution X-ray detectors made of very pure silicon can be found on the market. It seems, however, that the purification process of silicon does not yield the same excellent results as in the case of germanium. It is more of a lucky chance that sometimes the metallurgists succeed in producing a batch of silicon with an extremely low amount of impurities. Then, intrinsic silicon detectors are produced. Most of the time, the silicon crystal with little leakage current is obtained by the lithium drifting process. It should be noted that this process in the case of silicon results in a very stable product. At room temperature (without high voltage connected), a
Measurements of RadioactiviO, I
157 I
'1
I
I
I
I
I
I
10.0
1.0 z 0.!
0.01
I
I
,]
I
I
I
I
I
I
5
10
20
50
100
200
500
1000
2000
ENERGY / keV Fig. 5.35. Typical absolute efficiency curves for various Ge detectors with 2.5 cm source to end-cap spacing: (1) REGe, 15% relative efficiency; (2) LEGe, 100 cm 2 x 15 m m thick; (3) LEGe, 200 m m 2 x 10 mm thick; (4) Coaxial Ge, 10% relative efficiency.
Si(Li) detector can be stored for years, without losing its properties. When it is cooled again, it will display the same characteristics as when it was new. Therefore, the search for ultrapure silicon material does not have the same significance as in the case of germanium. The most frequently used silicon detector is a small pellet of lithium drifted silicon, some 6 mm in diameter and 3-4 mm thick. It is good only for X-rays, in the energy range between 4 and 50 keV. In fact, there are some standard sizes of Si(Li) detectors, areas of 30 and 100 mm 2 are considered normal, anything else is a detector made by specifications of the user. The energy deposited in a silicon detector by an X-ray is small, accordingly the electrical signal produced by collecting the charges from the detector is also small. Obviously, these signals will be correctly detected and analyzed only if they are clearly above the electronic noise. Therefore, the problem of signal to noise ratio is particularly critical with silicon detectors. As a rule, the electronics contributes more than half to the FWHM of an X-ray peak. This was probably one of the reasons that a novel type of electronic gadget was developed: the optically reset preamplifier. The planar germanium and the silicon detector are used for spectroscopy of X-rays. In one aspect, the Si(Li) detector is better: its escape peaks are very small, and do not interfere with the spectrum proper. In the planar germanium detector, as much as 15% of the X-rays coming into the detector will be registered as escape peaks, making the spectrum which is complex anyway, even more difficult to analyze. The planar germanium detector can be recommended if high energy X-rays; or very low energy gamma rays are to be analyzed. The efficiency of detection for X-rays at very low energies, say at 4 eV, strongly depends on the thickness of the entrance windows at the top of the detector. This is
158
Chapter 5 Z
>. r,.) z
100
I
I
I
I
I I li
I
!
/I
I I I
80 _
60 z9 h-
I"
0.0075 mm (0.3 m i l ) - ~ ~
"~
0"0075 mm (0"3 m i l ~ i
- - ~ 3 mm Thick
40
B
0.0075 mm (
i.. >,
20
2:
0
n
1
.a i
0.1
i
I
I I I I I!1
C
N
O
I
I
I I
|
F Nc Na MgAI Si I I 11
I
1..
I
I i
I I II
1
I
I
I
1 I ,I 1 I
10
/
100
ENERGY / keV Fig. 5.36. Efficiency of Si(Li) detector as a function of X-ray energy for different thicknesses of Be window and detector sizes.
usually a beryllium foil, strong enough to hold vacuum, but thin enough to let X-rays enter the detector. Its thickness is specified in thousands of an inch; 1 ml thick Be windows are frequently used, thinner are available. Thicker windows are recommended for countries with high humidity. For X-rays detected by Si(Li) crystal, the efficiency also depends on energy (see Fig. 5.36). At the low energy end, the efficiency decreases toward zero: no wonder, very soft X-rays cannot penetrate into the detector, and if they do not come in they cannot be registered. Some manufacturers are replacing Be windows with a window made of lower Z material which improves low energy response of the detector (see Fig 5.37).
Fig. 5.37. X-ray spectrometer of on detector with low-Z material window.
Measurements of Radioactivi~.
159
At the high energy end, the X-ray might travel through the detector without an interaction; the efficiency will decrease with the energy. Note that X-rays interact with the silicon mainly by the photoeffect; this is good--there is very little of the Compton tail, the spectrum looks cleaner.
5.2.2.4 Thermoluminescent detectors (TLDs) When radiation energy is absorbed by crystals of certain materials (e.g. lithium fluoride, lithium borate and calcium fluoride), the absorbed energy is trapped (stored) as displaced electrons within the crystal structure. If the material is heated later after the exposure, the trapped electrons are released and the stored absorbed energy is released in the form of visible light. This process is called thermoluminescence. Materials having this characteristic are called thermoluminescent. After its exposure to ionizing radiation, a thermoluminescent detector (TLD) is read out in special apparatus (the TLD reader) as follows: The TLD material is heated to a suitable temperature (about 200~ with a heating element or lamp; the trapped electrons in the TLD return to their normal energy state, releasing their extra energy as a light pulse. The light output is converted to an electric pulse by a photomultiplier tube (the working principle of which is illustrated in Fig. 5.26 for the scintillation counter). The size of the electric pulse is measured, and is proportional to the light output from the TLD material, which in turn is proportional to the total radiation energy absorbed, i.e. to the total radiation dose accumulated over the time that the TLD material was exposed. The TLD material can be used again after read-out. Thermoluminescent materials are commonly used in personal dose meters, in so-called TLD badges.
5.2.2.5 Nuclear track detectors When charged particles, e.g. o~ particles, impinge on certain types of plastic materials like polycarbonate or cellulose nitrate, they cause radiation damage tracks in the material. The tracks can be made visually detectable through chemical or electrochemical etching procedures. The visible tracks can be counted using a microscope, microfilm reader or automatic image analyzers. The number of tracks is used to calculate the total amount of radiation to which the detector material was exposed. Nuclear track detectors can be used to determine concentrations of o~-emitting radon and thoron; the number of tracks is related to the radon or thoron concentration. Nuclear track detectors can also be used to indirectly detect fast neutrons. Fast neutrons interact with the base material of a special film and cause recoil protons to be released. These protons then cause damage tracks in the film which can be made visible and counted as described above. The number of tracks can be used to determine the neutron dose.
160
Chapter 5
5.2.2.6 P h o t o g r a p h i c f i l m as a radiation detector
Ionizing radiation affects photographic film in the same way as does light. Films used for radiation measurement are sealed in light-tight packets. When the film is exposed, the radiation causes a latent image which becomes visible as blackening when the film is developed. The optical density ("blackness") of the film is measured by passing a beam of light through it. The radiation dose that the film received can then be inferred from the optical density.
5 . 2 . 2 . 7 N e u t r o n detection
Neutrons have mass but no electrical charge. Because of this they cannot directly produce ionization in a detector, and therefore cannot be directly detected. This means that neutron detectors must rely upon a conversion process where an incident neutron interacts with a nucleus to produce a secondary charged particle. These charged particles are then directly detected and from them the presence of neutrons is deduced. The most common reaction used in neutron detection today is: n + 3He --~ p + 3H + 765 keV where both the proton and the 3H are detected by gas-filled proportional counters using 3He fill gas. In addition, the following reactions are used to detect neutrons: 9 l~ reaction: A neutron collides with a ~~ atom and an c~particle is released in the process. The ionization caused by the o~particle can then be measured. This reaction is used in the BF 3 gas-filled proportional tube. Boron-lined proportional counters and boron-loaded scintillators are other examples of neutron detectors using this reaction. 9 N u c l e a r fission" A gas-filled detector, typically an ionization chamber, is coated inside with a thin film of fissile material like uranium. Absorption of neutrons by this material causes nuclear fission that produces highly ionizing fission fragments, which can be readily detected by the ionization chamber. 9 P r o t o n recoil: Fast neutrons can knock hydrogen nuclei right out of their atoms. The resulting so-called recoil protons can then cause ionization that can be detected and measured. This reaction can be utilized in gas-filled detectors, nuclear track films and scintillators. 9 N e u t r o n activation: This is the term used for neutron reactions that result in the formation of radioactive nuclides. This type of reaction is used in solid-state use-once devices containing one or more materials that are activated by neutron radiation. The induced activity of each material can be measured and the neutron exposure can then be calculated from these activities.
Measurements of Radioactivity
161
5.3 RADIOMETRIC METHODS 5.3.1 Counting The fundamental statistical treatment by Currie (1968) shows that the (low) limit of detection is proportional to the square root of the number of background continuum counts under the peak region of interest, where the proportionality factor varies with the confidence level chosen. Since the detection limit is expressed in counts, a more interesting parameter is the Minimum Detectable Activity (MDA), expressed in becquerels, and defined as the smallest quantity of radioactive nuclide which can be determined reliably, given the prevailing conditions of the specific spectral measurement. The MDA is inversely proportional to the absolute detection efficiency at the full-energy photopeak. Smaller MDAs can be obtained by lowering the background and increasing the detection efficiency (see also Watt and Ramsden, 1964; Friedlander et al., 1964). Therefore, for the measurement of the concentrations of radionuclides in environmental samples, and especially for low concentration levels, it is essential to reduce the background rate as well as improving the counting efficiency. When considering the background rate reduction, it is important to pay attention to the origins of the background. These are: 1. natural and artificial radioactivity in the environmental circumstances; 2. radioactivity in the detector and/or shielding material; 3. cosmic radiation; 4. instrumental noise. In the surrounding environmental materials, for example a concrete wall of a building, the soil, air or water, there are various kinds of natural and artificial radionuclides which give rise to a background. The most important are 4~ 226Ra and its decay products, and 232Th and its decay products. As a result, many lines originating from these nuclides are observed even in a heavy shielding box, during the spectrometry with a Ge detector. The background due to these radiations can be reduced considerably by shielding with heavy materials such as lead and iron. Typically, such background can be reduced to a hundredth by a 10 cm thickness of lead or 30 cm of iron. In making a shielding box, it is important that the detector is surrounded entirely (4rt direction) with shielding. In choosing the shielding material, one must pay attention to any undesirable contamination with radioactive substances. For example, a small amount of 21~ which is a member of the 226Ra series, is inevitably contained in lead. Its daughter nuclide 2~~ emits energetic [3-rays (Emax = 1.17 MeV), resulting in Bremsstrahlung which may contribute to an increase in background rate. The concentration of 21~ is very dependent on the mine where it was produced, and a careful check on the content of 2~~ is important for achieving ideal shielding characteristics. Since the half-life of 2mpb is 22 years, old lead has much better characteristics. Iron blocks are sometimes used instead of lead. However, the modern iron/steel is often contaminated with 6~ and careful checking is necessary prior to the construction of
162
Chapter 5
the shielding assembly. This 6~ contamination originates from the 6~ source which was used to monitor the abrasion of the furnace wall. For shielding purposes, the use of old iron, for example from materials from a sunken ship, is recommended. The background may also originate from the trace amount of radioactivity in the detector itself and its assemblies, and cannot be eliminated by shielding alone and hence, the contamination of the detector materials with radioactive substances should be carefully examined. A typical example is 4~ in the glass used for phototubes. This can be eliminated by the use of a quartz phototube. The 226Ra present in solder sometimes gives rise to a considerable rate of background. The molecular sieve used for keeping a vacuum in the cryostat of a Ge detector sometimes causes a background spectrum. In this sense, electrolytic copper, stainless steel, and perspex are recommended as materials to be used in the shielding box. Even when very thick materials are used for shielding, it is very difficult to stop the penetration of highly energetic cosmic-rays, which have energies in excess of 103 MeV. In order to overcome this undesirable background, an electrical guard technique is employed. The cosmic-rays which give a pulse to the main detector should produce another signal to one of the guard counters. The two signals from the main counter and from the guard counter are produced simultaneously. Therefore an anticoincidence technique is applied quite effectively to reduce the background rate, and an ultimate low background rate of the order of 0.1 cpm can be achieved by the combination of heavy shielding and the anticoincidence guard technique. The coincidence technique is often used for the low-level detection of electrons. As an example, we mention the low background [3-ray spectrometer as developed by Tanaka (1961). A small Geiger-Miiller counter is mounted inside a large plastic scintillator, and scintillation pulses from only those events which are coincident in the two detectors are analyzed by a multichannel analyzer. In this case, signals caused by cosmic rays are also subject to the analysis, but the minimum width of the scintillator is designed so as to produce a very large pulse-height, and this signal is rejected in the course of the pulse-height analyses. In these instruments, the 13-ray spectra can be measured for even very weak samples. By the addition of a logarithmic amplifier, the analysis of 13-ray spectra becomes even easier. Pulse-shape discrimination is also used sometimes for background reduction. Electrical noise is very often caused by the discharge of high-potential circuits, the transition signal of thyristors and the induction signal from spark-gaps, etc., which result in a background signal to the counter. To prevent this noise, electrical filtering in the main power supply, and electrical and/or magnetic shielding are sometimes required. Another essential requirement for reliable low level counting is to increase the counting efficiency and/or to use a larger amount of source sample. In order to increase the counting efficiency, one should use a large volume (large size) detector, and try to improve the source-to-detector geometry. However, it should be noted that with increasing detector size or volume, the background may also increase. As for the improvement of the geometry, the use of a well-type detector is advantageous, but the source size is sometimes limited. On the other hand, to use a larger amount of source, it
Measurements of Radioactivi~
163
is convenient to use a Marinelli-beaker, especially for the measurement of samples such as milk or soil. In principle, one can perform the measurement of activity of any sample by using radiation detection systems and standard sources. Standard sources and solutions are available from many national and international operations. However, for sources one can perform radioactivity measurement without the help of standards. Absolute counting implies of method of radioactivity measurement performed without the help of a standard/reference source. The methods for the absolute measurements can be classified into two categories. The first comprises the direct methods, which include the defined solid angle method, the 4rt counting method, the internal gas counting method, and the liquid scintillation counting method. These methods use relatively simple procedures, and each provides unique applicability. However, some uncertain corrections, such as for self-absorption, are involved in these methods, and the accuracies obtainable are somewhat limited. The second category is based on the coincidence technique, by which we can determine the radioactivity more accurately without any uncertain estimation of correction terms. The most often used methods include: the 2rt o~-counting method, 4n 13-counting with a 4rt gas flow counter, 4n 13-counting with a liquid scintillation spectrometer, and the 4rt 13-y coincidence counting method. 5.3.1.1
2n c~-counting m e t h o d
In cases of radioactivity measurements of an c~-emitter such a s 241Am or uranium source electrodeposited onto a metal disk, a simple 2rt c~counting with a gas flow proportional counter is practical, and a reasonable accuracy can be expected. If the applied potential to the proportional counter is set at the a-plateau region, the counter responds only to a-particles. Since the backscatter of c~-rays is small (a few percent) compared with that of [3-rays, and the value can be estimated as a function of the atomic number of the backing material, the radioactivity, n o, can be readily obtained from the observed counting rate, n, by the following simple relation: 1
n o = 2n ~ (l-a)
1
~ (1 + b )
(5.42)
where the factors a and b represent the corrections for the self absorption and backscattering of the alpha particles (Hino and Kawada, 1990). Self absorption, a, of a uniformly distributed source with superficial density, d, can be estimated by the simple equation: - a = d/(2R)
(5.43)
where R is the range (mg/cm 2) of the c~-particles in source material; this value can be referred from some data books; e.g. from Ziegler (1977). The effective thickness (superficial density), d, of the source can be determined from the line width of the
Chapter 5
164
pulse-height spectrum for a-rays obtained with a Si surface barrier detector mounted in a vacuum chamber as shown in Fig. 5.24. The effective thickness, d (in mg/cm 2unit), of the source material can be obtained from d =~/Vo2bs- V2i
(5.44)
/S
where Fob s and F i a r e the observed and intrinsic line widths (FWHM), respectively, and s is the stopping power of the source material for the specified alpha-particle. In the cases of the measurements of electrodeposited sources of RaDEF(2~~176176 for example, the stopping power of 294 keV cm2/mg for 5.3 MeV s-particle in PbO 2 should be used. If the thickness of the sources ranges from 0.59 to 1.10 mg/cm 2, the self-absorption values from 1.79% to 3.3% would result. Backscatter-values of s-rays are less than a few percent, and can be estimated from the published data as a function of the atomic number of source backing. The correction, b, for the backscattering can therefore be easily obtained with an acceptable reliability. This method is very simple in procedures, and can be applied to the absolute measurements of very weak sources without any standard source. For these advantageous features, this method is useful for the measurements of samples for environmental radioactivity analysis and also for the preparation of laboratory-made standard alpha sources. 5.3.1.2
4rt ~ - c o u n t i n g
w i t h a 4rt g a s f l o w c o u n t e r
The classical method of 47t [3 counting with a 4rt gas flow proportional counter is still useful for the absolute measurements of [3-emitting nuclides provided that good sources with small self-absorption can be prepared. From the observed counting rate, after fundamental corrections for background and counting loss due to dead time, the radioactivity, n o, can be calculated as 1 n o = n ~1 - a y
1 1- as
(5.45)
where aI and a s are the corrections for absorptions in the source-supporting films and in the source material itself. While the foil-absorption can be measured experimentally, for example, by the sandwich method, it is difficult to make direct determination of the self-absorption values only by the use of 47t counting. In practice, it will be convenient to estimate the self-absorption values from the published data taken as a function of [3-end point energy for various source conditions. The self-absorption values are very dependent on both the [3-ray energy and source conditions, and are in the range from 1% to several tens %. In order to get reliable data with this technique, minimizing the foil- and self-absorptions is essential. The source preparation techniques are therefore of great importance. As materials for source support, a VYNS film (copolymer of vinyl chloride and vinyl acetate), whose thickness
Measurements of Radioactivity
165
ranges from about 10 gg/cm 2 to 30 ~tg/cm2 can be easily prepared in the following manner. 1. Dissolve the VYNS powder (for instance from the Union Carbide Co.) in cyclohexanone (CsH,00) 5 times in volume, and keep it for at least one week in sealed glassware to ensure homogeneous mixing. 2. Attach a small amount of VYNS solution onto the edge of a glass plate (e.g. a slide glass for a microscope observation), and contact the edge onto the surface of clean water. A thin VYNS layer will be spread over the water surface. 3. Scoop up the film with thin metal rings for film support, and dry naturally. In this procedure, many source supports can be prepared simultaneously if a number of tings are set on a holder. 4. In order to avoid the accumulation of electrical charges on the source mount in a 4rt counter, the thin VYNS film thus prepared should be rendered electrically conducting by the vacuum deposition of gold or aluminum to both sides (--15 gg/cm 2 each side). The sources are prepared by a simple deposition of a known amount of radioactive solution of sample on the support. The solid contents of the sample solution must be kept to a minimum to minimize the self absorption. Even when carrier concentration is small (0.01-0.05 mg/cm2), a considerable amount of self-absorption is unavoidable. This is mainly due to the local agglomeration or crystallization formed in the dry. Several procedures as illustrated in Table 5.4 are useful to extend the source material uniformly onto the source support and consequently to reduce the self-absorption. For volatile materials such as iodine, mercury and selenium, special care and treatment are needed to form a more stable chemical form. In order to measure the radioactive concentration, i.e. radioactivity in a unit mass or unit volume of sample solution, quantitative deposition of sample solution is necessary. Table 5.4 Source preparation techniques for 4rt counting Name of technique
Procedures
Seeding
After adding a drop of freshly prepared 1:5000 dilution of Ludox-SM(Dupond) with a distilled water, dry in air with or without an infrared lamp irradiation.
Wetting
After adding a drop of a 1:10000 dilution of Catanac, dry in open air with or without an infrared irradiation.
Precipitation
For example: (1) Iodine isotopes: After adding a drop of AgNO 3 solution, dry naturally in open air. (2) Mercury isotopes: Dry in an atmosphere of H2S.
Ammonium complex forming
After adding a drop of pure water, dry in air with an infrared lamp irradiation. Add a drop of water again, and dry in an atmosphere of ammonium.
Chapter 5
166
This is achieved by gravimetrical means with a microbalance or with volumetrical ones with a micropipette. For precise measurements, the former method is preferable, in which a differential weighing with a pycnometer is essential to avoid the appreciable amount of evaporation of solution in the course of the weighing (0.5%). Although the attainable accuracy is limited to around a few percent in this simple 4re [3-counting method, this method can be applied to the measurements of weak source, and hence can play a special role in the environmental radioactivity measurements after chemical separation. 5.3.1.3
4~ ~3-counting with a liquid scintillation spectrometer
A liquid scintillation counter is considered as a version of a 4n counter, since the radioactive material is mixed in a scintillator, and the effective solid angle sustained from the source to the detector is nearly 4n. A large quantity of sample solution is usable in a liquid scintillation counting compared with the samples used for the original 4n. This is one of the advantageous possibilities of liquid scintillation counting. Although the pulse-height may be decreased and distorted due to the quenching phenomena, and accordingly the counting efficiency changes because of the source conditions, the possible counting loss due to the quenching effect can be corrected by using the pulse-height information. The classical means to correct the counting loss are the external standard method and the channel ratio method. However, the efficiency tracing technique has been proved to be more advantageous and convenient to determine the absolute activity. In this improved technique, most of the pure [3- and 13-7 emitting nuclides, including 3Hand '4C, can be measured accurately and readily by the use of only one or two common standard samples without the need to prepare quenching correction curves. In the relationship between the efficiencies for the standard sample and the count rates of the sample to be measured, the counting efficiency of 100% with respect to the standard sample serves to derive the radioactivity of the sample to be measured. Signal pulses resulting from liquid scintillation are analyzed in separate discriminating regions with a multichannel analyzer; the relation between the efficiency, E, and the count rate, n, is empirically expressed by the quadratic regression equation: n
=
aE 2 + bE + c.
(5.46)
Extrapolated value to E = 100% gives the true disintegration rate of the sample. This technique has universal applicabilities, and weak radioactive samples can be measured in terms of absolute activity. These features are considered to be advantageous in the environmental radioactivity measurements. 5.3.1.4
4~ or-counting with a liquid scintillation spectrometer
Since c~-rays emitted from radionuclides are energetic (-4 MeV), and almost monochromatic, the detection efficiency of a liquid scintillation counter for a-rays is
Measurements of RadioactiviO'
167
expected to be near unity. However, if [3-rays are emitted from the source in addition to cx-rays, the spectra originated from cx-rays and from 13-rays are sometimes overlapped partially. In order to overcome this effect, pulse-shape discrimination (PSD) technique is useful to select cz-signals from c~-13 mixing signals. The background due to 13-rays is entirely diminished after pulse-height discrimination. It should be noted that a special scintillator cocktail such as NE-213 should be used in this a-J3 PSD method. A similar cocktail can be prepared by mixing 5 g of PPO, 0.3 g of DMPOPOP, 100 g of naphthalene and 1 litre of xylene. This special cx-counting with a liquid scintillation spectrometer is useful for the (x-radioactivity measurements of environmental samples on an absolute basis.
5.3.1.5
4rt ~3-y coincidence counting method
Among various techniques used for radioactivity standardization, the 4rt [3-y coincidence counting method is undoubtedly the most important technique, and most radionuclides can be standardized by this technique or its extended techniques. Actually, most national standardizing laboratories in the world employ this method for the establishment of the national standards of various kinds of radionuclides. The 4rt ~-y coincidence counting is an improved version of the [3-y coincidence counting method. Here, a 4n ~ gas flow type proportional counter is usually used as the ~-detector, and a NaI(T1) scintillation counter located near the 4n [3 counter wall is employed to detect the y-rays. Source preparation techniques are the same as those explained in the 4n ~-counting. Even in this 4n 13-Y coincidence method, the source preparation to minimize the self-absorption is important to reduce the correction terms and also to shorten the extrapolation range in the 4rt [3-y coincidence extrapolation technique. Hence the special treatments as explained in Table 5.4 are important in the course of source preparation. For the measurements of ot-y emitting nuclide such as 24~Am, the same technique can be used, provided that the applied potential to the 4rt proportional counter is set at the cx-plateau region. This 4re I]-y coincidence counting technique can be readily applied to the absolute measurements of electron capture nuclides such as 51Cr, 54Mn, 88y, 755e etc., since a 4rt gas flow counter filled with P-10 gas (gas mixture of Ar + 10% methane) has considerable detection efficiency for the electron capture events through the detection of the characteristic X-rays and Auger electrons. This should be called 4rt X-y coincidence counting, but the measuring procedures are essentially the same as those for 4rt 13-y coincidence technique. Here the extrapolation technique is again useful for making various corrections. In cases of 4rt X-Y coincidence counting, an alternation of the flowing gas, from P-10 gas to pure methane, is effective to change the counting efficiency of the [3-counter to X-rays. This 4rt ~-y coincidence counting method and its extended technique are very useful for the precise absolute measurements of radioactivity, and therefore these
168
Chapter 5
methods are employed for the establishment of national standards of radioactivities of various kinds of nuclides in most of the national standardizing laboratories. Laboratory-made standard source can be prepared by these techniques. However, these techniques are not adequate for the measurements of low level radioactivities, and hence the direct application to the measurements of environmental samples is difficult. In this arrangement, apart from some corrections required, the disintegration rate (radioactivity), n 0, and counting efficiencies, e~ and ev, of the [3- and y-channels are given by the following simple equations for a ~-y emitter with a simple decay-scheme: n o = nf~nv/n c ~ = nc/n ~
and
ev = n~/n~,
(5.47)
where n 0, ny and n c represent counting rates, after correcting the background, accidental coincidences etc., in the [~-, 7- and coincidence channels, respectively. The above fundamental equations can be easily derived from the following three equations: n~ - noEf~,
nv = n0Ev,
and (5.48)
n c = noe~ev.
The explanation along this line is usually made in most textbooks. However, the ideal conditions are seldom achieved in any practical counting system, and some modifications of the fundamental equations are required in order to correct the possible effects which may disturb the ideal conditions. For example, the 4rt [3- proportional counter has an appreciable sensitivity to y-rays. Furthermore the y-transition is detected by the [~-detector through the internal conversion process, if any. Besides, because a coincidence mixer has a finite resolving time, false accidental coincidences are inevitably produced by chance. In addition to this problem, further consideration must be given when a nuclide with a complex decay scheme is measured. Taking account of all of these effects the coincidence equation becomes __ no
n ~ n v l - ~ R ( n~,__~.~ + n v , ( n C - 2 X R n ~ nv )l, 1 -- n c r,)
1+
(0r ~e + ~;1~,)- ~ + C 1 + ~
(5.49)
e~
and nc
e~
B
1
nr ( 1 - n ~ x )
(5.50)
Measurementsof Radioactivity
169
where e~ = y-sensitivity of the ~-counter; F__,ce counting efficiency of the l-detector to internal conversion electrons; tx = total internal conversion coefficient; e c = probability of obtaining y-y coincidence signal when a [3-particle is not detected; C = correction for the complex decay scheme; z R= resolving time of the coincidence mixer; x = dead time of the [3-channel. The notations with an asterisk denote the counting rate including background rate. These coincidence equations should be valid for all types of coincidence measurements. The effects due to the y-sensitivity of the ~-counter, the response of the ~-counter to internal conversion electrons and the effect originating from the complex decay scheme are all reduced by a factor of (1 - E~)/e~. In the 4n ~-y coincidence counting, the 13-efficiency, e, is expected to be nearly unity, and the corrections for these effects remain small. The attainable accuracy is therefore improved very much in this technique. This is one of the reasons why the 4n 13-y coincidence counting method is superior to the ordinary [3-y coincidence counting. However, when a nuclide with a complex decay scheme is measured, considerable correction is sometimes required. As a practical approach to overcome these difficulties, a series of 4n 13-Ycoincidence countings should be made for various source conditions which provide different ]3-efficiencies, and an extrapolation to (1 - e~)/E~ --~ 0 gives the true disintegration rate. It should be noted that most correction terms such as for the y-sensitivity of the ]3-counter and the response to the internal conversion electrons as well as for the complex decay scheme are all corrected automatically in this procedure. The shape of the efficiency function, graph of the n~n~/n~against (1 - e~)/e~, is dependent upon the decay-scheme and the T-gate setting etc. In the radioactivity measurements by means of coincidence technique, the counting statistics are given by the following formula: :
i
2else v -e:ls -e: v + 1
ncT
1
I noT
(1-~)(1-evi
....noE~E~T
(5.51)
The second term of eq. (5.51) is a function of (1 -e~), which will be reduced considerably in the use of a high efficiency detector, and excellent counting statistics are obtained in a short measuring time. This is another advantageous future use of the 4n 13-y coincidence method.
5.3.2 Gamma spectrometry High-purity detection systems having a very low background are suitable tools for the direct measurement of low-level radioactivity in environmental samples. The background features of the detection system are of considerable importance because they have to be known for one to obtain an estimate of the detection limit and of the minimum detectable activity (Curie, 1968). The natural radioactivity background originates from the uranium and the thorium series from 4~ and from cosmic rays. Natural radioactivity is found in most materials, and it is necessary to shield the
170
Chapter 5
detector using carefully selected materials of high density. The materials used in the detector assembly, and the shielding materials need to have the lowest possible inherent background radiation (see Gilmore and Hemingway, 1995). Containers such as Marinelli beakers can be filled with aliquots of environmental samples and placed on top of the end cap of the detector in an accurately reproducible beaker-detector geometry (Park and Jeon, 1995; Dryak et al., 1989). The radioactivity of the samples can be measured if the detector has been calibrated with Marinelli-beaker standard sources (MBSS) of the same dimensions, density and chemical composition. The calibration procedure should follow as closely as possible that defined in the IEC Standard 697 (1981). This method enables the simultaneous detection of several gamma emitters present in the sample matrix without the need for separation of the radionuclides from the matrix. The method can be applied to a large variety of environmental and biological materials such as air, water, soil, sediments, vegetation (grass, hay, etc.) and, particularly, individual foods of vegetable and animal origin as well as total diet mixtures. This method is suitable for the surveillance and monitoring of radioactivity originating from the operation of nuclear plants, nuclear weapons tests, and releases from nuclear accidents (Kanish et al., 1984). It is recommended that the gamma spectrometry system shall be a fully integrated data acquisition and computation system comprising the following items (IAEA 295, 1989): 9 A vertical, high purity germanium (HPGe) detector is recommended. The detector should have an efficiency of 18-20%. Generally, the efficiency of germanium detectors is specified as the photo peak efficiency relative to that of a standard 7.62x7.62 cm cylindrical NaI(T1) scintillation crystal and is normally based on the measurement of the 1.33 MeV y-ray photopeak of a6~ source with a source-detector spacing of 25 cm used in both measurement systems. The resolution of the detector which is normally specified for germanium detectors as the full width (in keV) at half maximum (FWHM) of the full energy peak of the 1.33 MeV peak of 6~ should be between 1.8 keV and 2.2 keV. It is recommended that the peak-to-Compton ratio of the detector be greater than 46:1. The peak-to-Compton ratio is defined as the ratio of the count in the highest photo-peak channel to the count in a typical channel just below the associated Compton edge and is conventionally quoted for the 1.33 MeV y-ray photo-peak of 6~ 9 A preamplifier is necessary. This is normally an integral part of the detector unit and it is located very near the detector in order to take advantage of the cooling which is necessary for the operation of the detector and which helps the preamplifier to operate with low noise. 9 A biased high voltage power supply is required to supply high voltage to the detector through the preamplifier. A power supply of 1500-5000 V is adequate for the operation of germanium detectors. 9 A linear amplifier is required to process the output signals from the preamplifier.
Measurements of Radioactivio'
171
9 A detector shield will be needed with a cavity which is able to accommodate large (up to 4 1) samples, constructed of either lead or steel with some type of graded line to degrade X-rays. Lead shields have a much lower back-scatter effect than steel shields. Typically, lead shields have walls 5-10 cm thick, lined inside with graded absorbers made of cadmium (--1-6 mm) and copper (--0-4 mm). Other materials, such as plexiglass and aluminium, are also used as absorbers. 9 A multichannel analyzer (MCA) with a minimum of 4096 channels should be connected to a keyboard and display screen for input and output of data and interaction with a computer. Several kits are available for the conversion of personal computers (PCs) into MCAs. Basically there are three types of conversion kits. One makes use of board with an analogue-to-digital converter (ADC) that simply clips into the PC; a second type uses a clip-in board with an external ADC; and the third type uses a multichannel buffer (MCB) connected to the PC. All of these PC-based MCA systems are relatively inexpensive and are very suitable for use in germanium and sodium iodide 7-ray spectrometry. 9 A rapid data-storage and recovery system is needed. It can consist of magnetic tape, hard disk, floppy disk, or a combination of these media. This system can be used for programming, short-term storage of data, and archiving data. 9 A high-speed printer is required for data output. Useful, but not absolutely necessary, is a plotter for archiving spectral drawings. 9 Software for system operation and data reduction is usually supplied with the MCA system. Software packages with varying features and capabilities are available for MCAs based on PCs. Several aspects of gamma spectrometry with such a system deserve some discussion. Interferences associated with gamma determinations may be caused by improper spectral identities, changes in background, errors in calibration and/or geometry, and lack of homogeneity in samples. Many of the problems in ],-ray spectrometry are due to malfunctions of electronic components. Very important also is the calibration of the measuring systems; both energy and efficiency calibration should be performed with care. Energy calibration of a germanium detector system (i.e. establishing the channelnumber of the MCA in relation to a ~,-ray energy) is achieved by measuring mixed standard sources of known radionuclides having well-defined energies within the energy range of interest, usually 60 keV to 2000 keV (IAEA-619, 1991). The use of the lower energy photons emitted by 24~Ammay indicate changes in the intercept. Mixed "f-ray standards are available in various forms and containers from reliable suppliers. A partial list of radionuclides usually available with gamma energies in the range of interest includes: 241Am, l~ and 57Co, 139Ce,51Cr, 22Na,54Mn,and 6~ The energy calibration source should contain a selection of radionuclides with at least four different ),-ray energies. It is recommended that one of the nuclides should be ~37Cs. The gain of the system should be adjusted so as to position the 662 keV photo-peak of ~37Csat about one-third full scale. It is also recommended that the gain of the system be adjusted to 0.5
Chapter 5
172
10 ~ 10: 9
~
,~
r
'~
-
10 ~ I
~.~
10 ~ 10 2
10' 10~ 0
I
I
1000
2000
....
/
3000
4000
PULSE HEIGHT / CHANNEL Fig. 5 . 3 8 . G a m m a - r a y
spectrum of an energy calibration source.
keV/channel. Once these adjustments are made, the gain of the system should remain fixed. As an example we show energy calibration with a point source incorporating 6~ 133Ba, and 137Cs. The peaks corresponding to the following energies are present in the measured spectra: 81.00; 301.85; 356.00; 661.65; 1173.24; and 1332.50 keV (see Fig. 5.38). The energy calibration, represented by a quadratic equation: E = a + bX ~+ cX 2
(5.52)
is determined (from the six pair data of peak centre channel (X) and the energy (E)) by using least square method for determination of the coefficient (a, b and c). An accurate calibration of the efficiency of the system is necessary to quantify the radionuclides present in a sample. It is essential that this calibration be performed with great care because the accuracy of all quantitative results will depend on it. It is also essential that all system settings and adjustments be made prior to determining the efficiencies and should be maintained until new calibration is undertaken. Small changes in the settings of the system components may have slight but direct effects on the counting efficiency. In general, it is not so easy to obtain the absolute peak efficiency e ( E , x ) which is a function of both y-ray energy (E) and geometry (x) between source and detector. The following method also includes an approximation that the energy dependency of the peak efficiency is scarcely affected by geometry, and the peak efficiency consists of two components being independent of each other. e(E,x)
=f(E)- e(x)
(5.53)
where, f i E ) represents the energy dependency, which was normalized at the specific energy and is often called "relative peak efficiency", e(x) gives the absolute peak
Measurements of Radioactivity
em
v'~
--
--
173
~
,
~
~,D
--
9
t~.
4
< ~.
3
9 ~
2
'
'
0.0
'
i
,
i
!
!
!
0.5
,,
,
9
!
,,i,,
|
1.0
,,!
!
.....
!
!
1.5
i
,
2.0
G A M M A RAY ENERGY / MeV
Fig. 5.39. Gamma-ray spectrum of the standard for efficiency calibration.
efficiency that was obtained for various geometries using the radioactivity standard sources emitting y-rays of the specific energy. The experimental procedure for determination of peak efficiency vs. energy curve involves the following steps: 9 Place a standard source (5~Cr, 54Mn,57Co,6~ 85Sr, 88y, 109Cd ' 137Cs ' 139Ce) at the position of 2 cm apart from the top face of the Ge detector using a source holder. Figure 5.39 shows the gamma-ray spectrum of such a standard source. 9 Accumulate the spectrum for a time period (t) long enough to reduce the counting error; approximately 1000 s is recommended. 9 After the measurement has finished, the spectrum shown in Fig. 5.39 is accumulated (integral net count of peak region) for each peak corresponding to the energy: 88, 122, 136, 166, 320, 514, 662, 835, 898,1173, 1332,1836 keV. 9 Calculate the peak efficiencies for all peaks using the following equation Peak efficiency: e =
peak area / t a.A
100%
(5.54)
where a is a fraction of y-ray emission per disintegration, and A is the present radioactivity (Bq). 9 Plot data of e(%) vs. E(keV) on log-log graph paper, then draw a smooth curve through all points. The result of such work is shown in Fig. 5.40, as an example.
Chapter 5
174
--
,.e--o--
-- o , _
i / e'
" qk
1
f(E)
xo``
',% "t,
0.5 ``o
0.2 0.1
I i III 0.05
I 0.2
0.1
!
I
I I I III 0.5 1
I 2
ENERGY / MeV Fig. 5.40. Relative peak efficiency as a function of energy.
The equation for the efficiency can be written in the form:
(5.55)
fiE) = exp [a + b- In(E) + c . ln(E) 2]
Quantitative y-ray spectrum analysis of environmental samples such as soil, water or ash of food requires the peak efficiency for volume sample. A Marinelli beaker is often used as a container for a large quantity of sample such as water or soil. As the volume of sample to be measured is usually fixed, the absolute peak efficiency is dependent on energy only, but affected by self-absorption which depends on density of matrix and the elemental composition. The method to be used for determination of the peak efficiency is based on experimental procedures involving the following steps: First, the Q-peak efficiency for an aqueous sample is determined by: 9 placing an aqueous standard sample (5~Cr, 54Mn, 57Co, 6~ 85Sr, 88y, 109Cd ' 137Cs ' ]39Ce) in the Marinelli beaker with the Ge detector; 9 accumulating the spectrum for a time period (t) long enough to reduce the counting error (approximately 1000 s of preset time is recommended); 9 after the measurement has finished, one should analyze the spectrum. Read the peak area for each following peak from the printed data 88, 122, 136, 166, 320, 514, 662, 835,898, 1173, 1332, 1836 keV; 9 calculate the observed peak efficiency Observed efficiency" e =
peak area / t
a.A
100%
(5.56)
where a is a fraction of y-ray emission per disintegration, and A is the present radioactivity (Bq) of each nuclide.
M e a s u r e m e n t s o f Radioactivio'
175
Let the peak efficiency for 662 keV be e~. Second, the Q-peak efficiency f o r a heavy liquid sample is determined by: 9 Replacing the above sample with a heavy liquid sample (ZnBr 2 in water; 80% in weight; p =2.5 g/cm3; weight composition: Zn(0.232), Br(0.568), H(0.02), 0(0.178)) in a Marinelli beaker of the same size. Then, measure the spectrum for approximately 1000 s. 9 After the measurement has finished, analyze the spectrum, read the peak area for 662 keV from the printed data, then calculate the peak efficiency. 9 Peak efficiency should be corrected by self-absorption. 9 Determine the self-absorption correction factor (K) by Eq. (5.57). K-
1
In e_L
~t 29 2 -- ~t l P l
(5.57)
E2
where P is density (g/cm 3) of matrix" 91 = 1.0, P2 = 2 . 5 ; ~t is the mass attenuation coefficient; ~t1= rt of water for 662 keV = 0.0857 cmZ/g; ~t2 = ~t of heavy liquid for 662 keV = 0.0742 cmZ/g. 9 Calculate the peak efficiency e* without self-absorption for each energy: E* = E/e -K~tp
where ~t is mass attenuation coefficient (cruZ/g) of water, and p is density of water (=1). 9 Plot the data of e* vs. E(keV) on log-log graph paper, then draw a smooth curve through all points. 9 From the curve, read the efficiency for each energy shown below as exactly as possible. 9 The coefficients (a, b and c) of absolute peak efficiency without self-absorption can be calculated separately for two regions of energy
E*(E) = exp[a + b. ln(E) + c. ln(E) 2]
(5.58)
9 Thus, the peak efficiency for a sample with linear attenuation coefficient (It" 9) is represented by the following equation.
e(E) = e*(E) 9e -K~
(5.59)
Most laboratories involved in radiation measurements now use personal computers and commercially available software for the analysis of "/-ray spectra. Some of these programs allow the user to control the multichannel analyzer (MCA), calibrate the detector for various geometries, and provide analysis results. The programs are easy to use and do not require the user to be an expert in "/-ray spectrometry. As an example we refer to work by Heimlich et al. (1989). In their work, the gamma-ray spectrum analysis program, GAMANAL, has been modified to operate on a small computer. The program uses an algorithm involving a Gaussian and a tailing term for fitting and resolving peaks obtained from spectrometers using germanium detectors. Gamma-ray energies, intensities and absolute photon emission rates can be determined. A graphical output showing the original and fitted data can also be
176
Chapter 5
obtained. The results generated by the program are stored on disk for further analysis. This allows the use of other computer programs and languages in tasks such as decay curve analysis, radionuclide activity measurements and neutron activation analysis. Sanderson (1989) evaluated commercially available IBM PC-compatible software in 1987. At that time, it was reported that most of the programs satisfactorily detected peaks and resolved doublets of equal intensity. Problems arose when the doublets were of unequal intensity or the analysis of a complex spectrum was needed. The suppliers of the programs involved in that study have corrected some of these deficiencies. Since many of these programs have undergone numerous revisions, and a few programs have become available, a re-evaluation was performed (Decker and Sanderson, 1992). Six programs were evaluated in this study: GAMMA-W (Gesellschaft ftir Kernspektrometrie, Germany), INTERGAMMA (Intertechnique Instrumentation Nucleaire, France), QSMPlus (Aptec Nuclear, Inc, Canada), OMNIGAM (EG&G Ortec, USA), GDR (Quantum Technology, Inc., USA), SAMPO 90 (Canberra Nuclear Products Group). Hardware requirements were similar for all the programs tested (IBM PC compatibility, 384-640 K of memory, a hard disk drive). Two programs, QSA/Plus and SAMPO 90, also required a maths co-processor. The QSA/Plus program had to be installed in Windows 3.0. All of the other programs operated directly from DOS. Except for GAMMA-W, all the programs control data acquisition. GAMMA-W, which is written by an independent company that does not manufacture nuclear instrumentation, does not allow the user to read ten different formats of y-ray spectra. The conclusions reached by the authors were as follows. All of the programs satisfactorily found small peaks using sensitivity values recommended in the program manuals. However, these values may not be optimal for every situation. When the sensitivity values were lowered, additional valid peaks were found. When the recommended sensitivity values were used, only two programs did not report any false peaks. All of the programs were able to resolve equal-intensity doublet peaks with only a 2 keV (4-channel) separation. The resolution of doublets of unequal intensity, especially where the smaller is on the high-energy side of the predominant one, has improved since the last evaluation. However, some programs still require improvement in this area. All the programs for analysis of gamma spectra measured with germanium detectors generally allow the following steps (among others) to be carried out: a. transfer of spectra from the multichannel analyzer to the computer; b. search for and identification of photopeaks; c. energy and efficiency calibrations; d. resolution of multiplets, e. calculation of activities and the dosimetric magnitudes deriving from them. Most of these programs, however, are not specifically designed for the development of plans for environmental radiological monitoring, where one must measure periodically and systematically activity levels of a set of man-made gamma emitters in different types of environmental samples. Such samples often present very low counting statistics for these emitters.
177
Measurements of Radioactivity
Moreover, since most of these programs employ elaborate algorithms for the search and fit to the shape of the gamma-peaks, they are coded in FORTRAN (Aarnio et al., 1984; Koskelo et al., 1981; Szekely, 1985) and are therefore implemented on minicomputers or, in more reduced versions, on IBM-type personal computers (Helmer and McCullagh, 1983; Hensley et al., 1988). Also commercially available programs (Canberra, 1987) are difficult for the user to modify, requiting usually a great deal of programming effort to adopt them to the peculiarities of each laboratory. Baeza et al. (1992) have developed ESPEC: a set of programs especially designed to undertake gamma spectrometric analyses of environmental samples. The processes enumerated above (a-e) are carried out in a simple form. The package is designed so that it can be implemented not only on an IBM-type personal computer, but even on one with less memory and lower performance, such as a 64 kbyte memory microcomputer. The code is written in BASIC and it is easily modified to suit the specifications of any multichannel analyzers.
5.3.3 Beta particle spectrometry Table 5.5 lists beta-emitting radionuclides; since beta emitters show a continuous emission spectrum, the average energy E 8 and the maximum energy E~" are given. Measurement uncertainties are shown in brackets after the respective value (in units of the last significant digit). The classical way to measure low-level beta particle activity is with a GeigerMiiller (GM) gas-flow counter. The block diagram of the GM counter is shown in Fig. 5.41. In anticoincidence with a guard detector the GM gives a fairly low background (0.003 s-1) and a counting efficiency of 40%. The disadvantage, however, is its inability to give energy resolution. Table 5.5 Some common beta emitters Nuclide
Half-life
E'~
EB( % )
3H
12.35(1) a 5730(40) a 14.29(2) d 87.44(7) d 2.75(2) a 96(4) a 50.5(1) d 28.7(3) a 64.1 (1) h 59.3(2) d 22.3(2) a
18.6 156.48 1710.4 167.47
5.68 49.47 695.0 48.80
65.87 1492 546.0 2284
17.13 583.1 195.8 934.8
16.5 63.0
4.15 (80) 16.13 (20)
14C
32p 35S 55Fe
63Ni 89Sr 9~ 90y 125I 21~
Er
Ks 5.9, K~ 6.5
Ka 27.4, KI331.0
178
Chapter 5
IR High-voltage power supply
C
II
Amplifier
Scaler
Pulse shaper and discriminator Timer
,k Source Fig. 5.41. Block diagram of the counting electronics associated with a G-M tube.
As the characteristics of the GM tube may gradually change over a long period of use, it is necessary to examine the characteristics of the tube and background count occasionally. The counting efficiency of a GM counter for detecting [3-rays is defined as the ratio of the count rate to [3-activity of the source. Thus, the counting efficiency is one of the most important factors to determine the characteristics of the detector. There is the following relation between activity A (Bq) and net count rate n (s -~) in the case of activity measurement using the GM counter n = A - f g - f / - f s o - f ~ "fa
(5.60)
where f~ = geometry factor of counting setup; f = intrinsic efficiency of GM tube for [~-rays; f~ = source self-absorption factor; f~ = source-mount backscattering factor; f~ = correction due to absorption between source and GM tube. Equation (5.60) is rewritten as eq. (5.61), iff~, f , fsa, fs is replaced by rl, n
Afo where rl is called counting efficiency; it represents the rate of count rate n corrected for the adsorption f~ to activity A. The counting efficiency 11 is obtained by using the standard source of the known activity A, and the absorbers with the known thickness d,,. If the count rates for 13-rays are defined as n and n o respectively, which correspond to those passing through with and without absorber, f~ is equal to n/no; where f~ represents the transmission rate of [3-rays for the absorber thickness d m. When the thickness of absorber d m is not so large compared with the maximum range of [3-rays, n can be represented by the experimental formula n = no e-"'dm
(5.62)
Therefore, if one plots the n values as a function of the thickness d m on semi-log graph paper, the n values decrease linearly with the d m values increase. The slope of this line is
Measurements of Radioactivi~
179
p,,, which is called the mass absorption coefficient. If one puts din=0, the count rate n o (=n/fa) can be obtained. Thus the equation becomes no
rl = ~ A
(5.63)
If 1"1is already known, the unknown activity A' is obtained by V/t0
A' = ~ 11
(5.64)
where n 0 is the count rate extrapolated to dm=O. Furthermore, in the case that the actual count rate n is larger than 100 s-~, it is necessary for the count rate to correct the resolving time "c[s] of the GM counter. It has been shown (Holm et al., 1990) that ion implanted detectors can be used not only for alpha particle spectrometry but also for beta particle spectrometry. Their drawback, however, is the high background around 100 keV and noise below 1 keV. Olssen et al. (1992) have described a detector system for low-level beta particle spectrometry where the good characteristics of gas flow and silicon detectors are used. The gas-flow GM counter used in the detector system is a windowless single-channel version of the GM-25-5 multicounter developed in the Riso National Laboratory, Denmark. The gas-flow counter utilizes a GM counter with a diameter of 22 mm and a guard of 80x90• mm, using argon (99%)/isobutane (1%) as counting gas. Background counts, produced in the sample counter by cosmic radiation, are reduced by means of the guard and attached anticoincidence circuits. As the energy-discriminating detector, a passivated implanted silicon detector, with an active area of 450 mm 2 and a depth of 300 ~tm, was used. The PIPS detector has a 0.5-~tm-thick aluminium coating on the front surface. The detector is placed on top of the gas-flow counter, and integrated into the gas detector unit, allowing the aluminized front surface to act as one of the ground-electrodes of the sample counter (Holm et al., 1990). The source is placed in a cavity inside the detector, between the sample and guard counter. The instrumentation used for PIPS detectors is identical to that used for alpha particle spectrometry. The authors have demonstrated that it is possible, by the coincidence technique, to reduce the contribution of noise and background from a PIPS detector by a factor of ten and to improve its energy resolution. Such a detector system could be a useful tool for quality control of low-level, low-energy, pure beta emitters such as 63Ni from environmental samples. t
5.3.4 Alpha particle spectrometry Common radioisotope sources of alpha particles are listed in Table 5.6. All but the first one listed are members of radioactive decay chains. Decay chains are classified into four groups according to their mass numbers. They are Th-series whose mass number is 4N (N is integer), U-series of 4N+2, Ac-series of 4N+3 and Np-series of 4N+l. An Np-series does not exist naturally because the half-life of its longest-lived member is
180
Chapter 5
Table 5.6 Common alpha emitters Nuclide
Half-life
Alpha energy/MeV (% Branching)
147Sm
1.07• 60.55 m 138.38 d 0.296 ~ts 55.6 s 3.824 d 1600 y 1.405x101~ y 2.45• y 7.038x 108 y 4.468x109 y 2.413x 104 y 6570 y 432 y 18.1 y
2.232(100) 6.051 (72), 6.090(28) 5.304(100) 8.784(100) 6.288(99.93) 5.490(99.9) 4.602(5.5), 4.784(49.5) 3.954(23), 4.013(77) 4.723(27.5), 4.775(72.5) 4.368(12.3), 4.400(57) 4.150(23), 4.197(77) 5.105(11.5), 5.144(15.1) 5.157(73.3) 5.124(26.4), 5.168(73.5) 5.443(13.1), 5.486(85.2) 5.763(23.6), 5.805(76.4)
212Bi
210po 212po 22~ 222Rn 226Ra
232Th 234U 235U 238U
239pu 24~ 241Am
244Cm
three orders of magnitude shorter than the age of the Universe. In three natural decay chains activities of each parent and succeeding daughter nuclides are equal. This condition is called radioactive equilibrium, because half-lives of parent nuclides, 232Th, 238U and 235U are much longer than those of their daughter nuclides. Alpha particles emitted from nuclides which decay to a single level are observed as mono-energy particles. On transitions given the branching ratio in Table 5.6, multiple alpha energies are observed. Such a fine structure in the alpha spectrum comes about because an alpha emitter may decay to any one of several discrete energy levels of its daughter. 241Am is commonly used as a standard source. The detection of alpha particles is based on the physics of the processes which take place when the particles pass through matter. During this passage the alpha particle loses its energy by excitation and ionization of the atoms. The energy loss per unit path length is called stopping power(-dE/dx). Stopping powers of various elements are given (Ziegler, 1977; Northcliff and Schilling, 1970). Stopping power at energy above 1 MeV is inversely proportional to alpha energy. At the energy range of less than 1 MeV stopping power is nearly proportional to the energy. Because the velocity of alpha particles is slow compared with orbital electrons, so alpha particles capture electrons and their effective changes for ionization decrease. When alpha particles pass through matter, they make small angle scattering. Therefore an asymmetric line spectrum which has a tail extending to the lower energy side is observed after penetrating relatively thick foils, because of asymmetry in their trajectory path.
Measurements of Radioactivi~
181
Source thickness has an important effect on the observed alpha particle spectrum. As alpha particles have a relatively large stopping power, the observed spectra for thick sources show some degradations caused by self-absorption. For high resolution spectrometry, thin alpha sources or samples are required. They are usually prepared by electrodeposition or vacuum evaporation. With a source-thickness less than 10 ~tg/cm2, no effects due to self-absorption are observed. For such thin sources, a Gaussian-shape line spectrum whose width is limited by the detector energy resolution, is observed for a mono-energetic alpha source. When there is an absorber between an alpha source and a detector, the observed energy is reduced by the energy loss in the absorber. Absorbers are the entrance window of a detector, the covering material which prevents contamination of a source, air, and so on. In the case of a thin source, prepared by chemical separation, the emitted alpha particles are observed as line spectra by spectrometers having high energy resolution. The intensity of each alpha emitter is easily estimated from the area of the corresponding peak. Even if the peak is superimposed on the tail of another peak, the peak area can be calculated from the shape of the line spectrum. An example of this situation are plutonium isotopes, 239puand 240r~ vu. They are used for estimation of a burn up of nuclear fuel. As the energy difference of these alpha emitters is only 10 keV, the alpha particle spectrum is observed as an overlapped single peak. However, when a Si detector is used, which has an energy resolution of less than 10 keV (FWHM), the overlapped peaks can be analyzed by the least squares fitting technique. Alpha particle spectrometry is usually performed using Si detectors, which are especially useful for thin and small-area alpha sources. Two types of Si detectors are commonly used; the surface barrier type and the ion-implanted PN-junction type. As an alternative to Si detectors, a Frich grid ionization chamber is sometimes used as a spectrometer of alpha particles. This ionization chamber has a grid between a cathode and anodes and a sample is put on the cathode electrode. Electrons and ions, which are generated between the grid and the cathode by ionization, drift towards the anode or the cathode, and a signal pulse is obtained. The pulse height obtained from the cathode depends on the emission angle of alpha particles. Only the drift of electrons is observed due to their drift velocity being 1000 times greater than that of ions. However, the height of the anode pulse is proportional to the alpha energy. The advantage of this counter is that the areas of samples can be made larger than with Si detectors. A commercially available counter of this type has an area larger than 1000 c m 2. The energy resolution of this counter is 40-50 keV (FWHM) for alpha particles. Several other types of spectrometers can be used in some specific applications. For example, an organic liquid scintillation counter is useful for detection of extremely low-level alpha activities. Alpha emitters, chemically separated from samples, are mixed into the scintillator. The geometrical efficiency is then 100% and a relatively large amount of source can be introduced into the scintillator. The main drawbacks of the counter are its poor energy resolution and its relatively low light output in excitation by alpha particles compared to that by electron.
182
Chapter 5
Track detectors are also often used for alpha detection. The length of tracks produced by alpha particles is measured with a microscope after appropriate chemical etching. The energy spectrum of alpha particles is obtained from the distribution of their trace lengths. Advantages of this detector are its high sensitivity, good discrimination against [3- and y-ray backgrounds and the acquisition of alpha emitter distribution in samples. Its main drawbacks are a poor energy resolution and the time-consuming processing of the microscope reading. It should be mentioned that the energy of alpha radiation is one of the most important characteristics of radionuclide sources. Knowledge of the alpha particle energy is necessary for the determination of other major characteristics of external radiation of alpha sources, such as the flux and energy flux density, and the absorbed dose rate. Information on the alpha radiation energy is also used for calibrating semiconductor alpha spectrometers. The results of accurate measurements of alpha particle energies play an important role in the evolution of the atomic-mass scale and comprise nearly 60% of the input data for atoms with A > 200, and are deciding factors in the design of precision spectrometers for the high-precision energy measurements of alpha particles from radionuclide sources with the minimum attainable uncertainties. A very precise measurement of alpha particle energies can be achieved by the measurement of the alpha particle time-of-flight (Frolov, 1992). A special problem is the assay of alpha particle emitters in water samples. There are several procedures in the literature for assaying alpha particle-emitting radionuclides in natural waters. However, few of them are well suited for low-level analysis since they require the handling of large water samples and the application of concentration techniques. Typical water samples do not contain sufficient amounts of some important nuclides for a precise or reasonably rapid measurement. The most common concentration technique is the coprecipitation of the heavy elements by the addition of a carrier. Some of the best-known methods are coprecipitation with iron as the hydroxide and with calcium and strontium as the oxalates (Livingston et al., 1975). However, these methods present some disadvantages. These include the transportation of large volumes of water (several, to hundreds of litres) to the laboratory, the isotopic exchange of tracers, the use of large containers and laboratory ware, and the difficulty of recovering the precipitate. As a consequence of these difficulties, interest in studying in situ concentration techniques with appropriate adsorbents has increased. These methods eliminate the need for preservation, storage, evaporation or coprecipitation of large water samples. One such approach has been described in detail by Crespo et al. (1992). According to the authors MnO 2 and A1203 can be used as absorbents in the assay of the low levels of alpha emitters in the waters. This technology requires more development work before it can become routine. Another problem of importance to safeguards and reactor fuel technology is the measurement of the relative abundance of plutonium isotopes. In addition to 7- and X-ray spectrometry (Gunnick, 1981), alpha particle spectrometry has also been used (Bland et al., 1992; Amoundry and Burger, 1984; Bortels and Collaers, 1987). This
Measurements of Radioactivity
183
usually results in a complex alpha spectrum, with difficulties in obtaining correct amplitudes for overlapping peaks which tail towards lower energies. The 239pu and 24~ peaks overlap when alpha particle energies are measured using Si detectors. Some sort of deconvolution procedure is required in this case (see, for example, Bland et al., 1992). The radiation dose to man due to radon and its decay products in the environment amounts to more than 50 percent of the total radiation dose to man from natural radiation sources. Recently the risk of developing lung cancer from exposure to inhaled radon daughters has been attracting considerable attention. Thus it is very significant to measure alpha-activity levels in the air. Let us briefly describe a filter-sampling method used to assay the gross alpha-activity concentration of radon daughters in the air by means of the gross alpha-particle counting with ZnS(Ag) scintillation counters. An air sampler is used to collect natural radionuclides onto a filter paper by passing air through, because some nuclides of radon progeny (RaA=2~SPo, RaB=Zl4pb, RaC=214Bi and RaC'=21apo etc.) tend to attach to aerosol particulates floating in the air. Supposing only one radionuclide is attached to the surface of a filter paper, the rate of change with time in the total number of the radioactive atoms on the filter paper can be written dN dt
= ~qn - L N
(5.65)
where t = sampling time, N = total number of radioactive atoms on filter paper at time t, = flow rate of sample air (constant), n = concentration of radioactive atoms in the air (assumed to be constant), ~, = decay constant of the radionuclide, ~qn = rate of collection, and LN = rate of decay. The solution of Eq. (5.65) for the initial condition N = 0 at t = 0 is N = ~qn (1 - e ~' )
(5.66)
The radioactivity A is given by LN, or A = ~qn(1 - e -z')
(5.67)
The counting rate R of alpha particles emitted from the filter paper at time t is related to the radioactivity A" R = e~A
(5.68)
where e = counting efficiency of alpha ray counter and ~ = emerging efficiency of filter paper. From eqs. (5.67) and (5.68) the concentration n of the alpha-emitting nuclide in the air is obtained"
184
Chapter 5
R
n=
(5.69)
e ~ q ( 1 - e z' ) The alpha-activity concentration Q in the air is given by ~,n, or a
e ~ q ( 1 - e ~' )
R 0.693 1 e~q T 1 - ( 1 / 2 ) '/r
(5.70)
where T (= 0.693/X) represents the (alpha-decay) half-life of the radionuclide. For long-lived radionuclides such as 238U, 239ptl and 226n t~a, eqs. (5.66), (5.67) and (5.70) become N = ~qnt
(5.71 )
A = ~,~qnt
(5.72)
Q=~
R
(5.73)
e~;~qt
respectively.
5.3.5 Liquid scintillation measurement method Liquid scintillation counting (LSC) has been accepted as the generally preferred method of counting weak beta emitters and is useful to a lesser extent for t~- and ),-emitters. The counting sample consists of three components, the radioactive material, an organic solvent or solvent mixture, and one or more organic phosphors. A particle or radiation emitted by the sample material is absorbed in, and its energy transferred to the solvent and then to the phosphor which emits a scintillation of light photons. These photons are absorbed by the photocathode of a photomultiplier tube which converts them into an electronic pulse. The pulse, after suitable amplification, is registered as a count corresponding to the emission of the particle or radiation (Neame and Homewood, 1974). Configuration of a typical LSC is shown in Fig. 5.42. The technique has the following characteristics. 1. Self-absorption is usually negligible. 2. There is no absorption of radiation by air or a detector's window between the radioactive sample and the sensitive region of the detector. 3. There is no radiation scattering prior to incidence upon the detector. 4. 4rt counting is performed, because the radioactive material is completely surrounded by the liquid scintillator.
Measurements of Radioactivio'
185
Fig. 5.42. Configuration of a typical liquid scintillation counter.
Based on the above merits, the liquid scintillation measurement is extremely sensitive to the low-level radioactivity existing in the environment and food. Since the radioactive sample material from most methods of sample preparation is in intimate contact or in actual solution with the phosphor, the detection of emitted particles or radiation is highly efficient and may even approach 100%. Problems of self-absorption of the emissions are thus absent, or considerably smaller than those associated with planchette counting of solid samples. This is of particular importance for the measurement of low energy beta emitters such as tritium and carbon- 14. On the other hand, the measurement method has intrinsic drawbacks such as quenching and chemiluminescence. Currently, the liquid scintillation counter has been employed not only for the measurement of low energy [3 emitters, but also for pure [3, 13-y, and o~-emitters and further Cherenkov radiation. The liquid scintillator consists mainly of organic solvent and fluorescent material (i.e. solute), and sometimes a surfactant or other material is added to the solution. The characteristics of the liquid scintillator depend mostly on the sort and amount of these chemicals. The liquid scintillator plays the role of an energy transducer, converting radiation energy into photons. The organic solvent which
Chapter 5
186
comprises most of the liquid scintillator should satisfy the following conditions (Neame and Homewood, 1974): 1. The energy should efficiently transfer in the process of luminescence. 2. An absorption spectrum of solvent should never overlap an emission spectrum of solute. 3. The radioactive sample and solute must be able to be incorporated with the solvent. 4. The solvent must be of high purity. In the past many kinds of chemicals were employed as solvents, but nowadays only a few typical solvents are being used including toluene, xylene, pseudocumene and droxane. The characteristics of typical solvents used for liquid scintillators are shown in Table 5.7. The solute is classified into a primary solute and a secondary one. The former is a main fluorescent material, and the latter serves as a wavelength shifter which gives rise to an emission spectrum having long wavelength. These characteristics and molecular structures are respectively represented in Table 5.8. The sample to be measured is easily prepared by incorporating the radioactive sample into the liquid scintillator such as xylene-base-, toluene-base-, or emulsive scintillator. The xylene (or toluene)-base-scintillator is appropriate for hydrophobic Table 5.7 Typical solvents used for liquid scintillator Solvent
Molecular weight
Solidifying point (~
Absorption spectrum (A) ~. max*
Emission spectrum (A) ~. max*
)~ mean**
Toluene Xylene Pseudocumene Dioxane
92.13 106.16 120 88.1
-95 -20 -60.5 12
2.620 2.660 2.690 1.880
2.870 2.890 2.930 2.470
2.840 2.880 2.900 -
Relative pulse height 1.00 1.09 0.65
*Wavelength giving maximum value; **wavelength giving mean value.
Table 5.8 Typical solutes used for liquid scintillator Solute
Primary solute PPO Secondary solute DMPOPOP bis-MSB
Absorption spectrum ~. max*
Emission spectrum k max*
~. mean**
Optimum Decay time of concentration scintillation (g/l)
3.030 3.630
3.640 4.290
3.703 4.273
4~7 0.2--0.5
1.6 1.5
3.470
4.120
4.219
-0.5
1.3
*Wavelength giving maximum value; **wavelength giving mean value.
Measurements of Radioactivity
187
samples, to form a homogeneous solution which provides efficient energy transfer and light-counting efficiency. The simplest mixture consists of a primary scintillator and a secondary scintillator dissolved in a primary solvent. The primary solvent converts the kinetic energy of radiations into excitation energy. The primary solvents are usually aromatic hydrocarbons, and therefore only non-polar radioactive materials can be dissolved in them. The most widely used primary solvent is toluene. Others are p-, and m-, and mixed xylenes. Pseudocumene (1,2,4-trimethylbenzene) is becoming a popular solvent for new, commercially produced scintillation cocktails. It offers the highest energy conversion efficiency of the solvents known, and has fewer restrictions on shipping and storage as a combustible liquid because of its high flashpoint, low volatility and lower toxicity. Owing to the presence of organic solvent, the mechanism of luminescence for the liquid scintillator is much more complicated than that for a crystal scintillator. The energy successively transfers to generate finally the luminescence, as follows: 1. excitation of solvent molecule due to the absorption of radiation energy; 2. solvent-solvent energy migration; 3. solvent-solute energy transfer; 4. luminescence from solute molecule. Since the solvent molecules have a majority in the liquid scintillator, they are initially excited by the radiation energy. The rt electron of the solvent molecule plays an important role in the process of energy transfer due to its active mobility. When the solute absorbs the excitation energy liberated from the solvent, the solute molecules in the ground state are excited up to the excited electron level or its vibrational one. Then, the following processes compete with one another as intramolecular behaviours of the solute: 1. internal conversion of molecule, 2. fluorescence emission, 3. intersystem crossing, 4. phosphorescence emission. The quenching is a phenomenon caused by the energy loss in the process of the energy transfer inside the liquid scintillator. It should be taken into account, whenever the liquid scintillation measurement is performed. Owing to the existence of a quencher, which means the material giving rise to the quenching, the counting efficiency finally decreases as follows. 1. Energy is partly lost during the energy transfer. 2. The number of photons emitted from the solute decreases, and/or some of the photons are wastefully absorbed in solution. 3. The height of electric pulse created by a photomultiplier tube becomes lower. 4. A pulse height distribution shifts toward a lower pulse height. 5. The number of pulses which are present within a countable region decreases, leading to the lowering of the counting efficiency. Most materials are regarded as quenchers whose quenching strength depends on the material itself and its amount. The quenching and optical (or colour) quenching are
188
Chapter 5
practically significant. The impurity quenching is the quenching which arises in the process where the excitation energy of the solvent is provided to the solute. Its mechanism can be explained by the theory of an exciplex which means the excited complex formed by two different kinds of molecules. Final energy loss is represented by the following scheme. 1M* + Q ~ (M -~+ Q~)* --~ M-* + Q~ (dislocation)
(5.74)
exciplex 3M + Q ~ ~M + Q (transition to ground state)
(5.75)
(through intersystem crossing) where 1M = solvent molecule in singlet state, 3M = solvent molecule in triplet state, Q = quencher molecule, and * = excited state. When the material whose absorption spectrum overlaps, more or less, the emission spectrum of the solute is contained in a prepared sample, the fluorescence generating from the solute is partly absorbed to cause the optical quenching. The strength of the optical quenching depends on the sort and density of colours. In general, red and yellow cause strong quenching, but blue causes weak. Even though a quencher is colourless, the optical quenching occurs if the quencher has an absorption spectrum which partly overlaps an emission spectrum of solute. The liquid scintillation measurement method is a quantitative technique to determine radioactivity, but not a qualitative one to find radiation energy or the sort of nuclide. Based on the internal-sample counting, the measurement method has great merits for low-level radioactivity. The following three measurement methods are in general often used.
(1) External standard method The index of quenching with respect to sample is found from Compton spectrum created by "f-ray irradiation to the sample. A set of quenched standards are employed for the construction of a quenching correction curve which represents the relationship between the index of quenching and the counting efficiency. The radioactivity of the sample to be measured can be determined by using the quenching correction curve.
(2) Sample channel ratio method The index of quenching in this method is obtained from an areal ratio or a centre of gravity with respect to a 13-ray spectrum. The radioactivity is determined through a similar quenching correction curve to that of the external standard method.
(3) Automatic efficiency tracing method As a matter of fact, the foregoing two methods are confined to the measurements of only tritium and carbon-14. The automatic efficiency tracing method, however, makes it possible to determine the radioactivity of many sorts of radionuclides such as pure 13,
Measurements of Radioactivit3,
189
13-7 and o~-emitters. The radioactivity can be found from a quadratic regression equation which is constructed by the relationship between the counting efficiencies of a standard sample and the counting rates of a sample to be measured. Sample preparation techniques involve the chemical separation of the specific chemical forms of a radionuclide, and the preparation of the counting sample by mixing the separated nuclides with the proper liquid scintillator. A large number of recipes for scintillation mixtures have been published, but for many purposes a few simple ones will suffice. The properties of a suitable mixture are as follows: 1. It should generally be clear, colourless and homogeneous after addition of the radioactive sample, although in certain cases suspensions or gels may be satisfactory. 2. It should quench as little as possible. This is particularly important when measuring radioisotopes such as tritium whose beta particles are of low energy. 3. It should not be expensive. 4. Its constituents should be stable. Some scintillators are known to be unstable when exposed to light and other materials may deteriorate on storage, with the formation of impurities which cause quenching or chemiluminescence. Scintillation mixtures should be stored in dark bottles. The simplest mixture consists of a primary scintillator and a secondary scintillator dissolved in a primary solvent. The primary solvent converts the kinetic energy of radiations into excitation energy. Primary solvents are usually aromatic hydrocarbons, and therefore only non-polar radioactive materials can be dissolved in them. The most widely used primary solvent is toluene. Other primary solvents are p-, m-, and mixed xylene. Pseudocumene (1,2,4-trimethyl benzene) is becoming a popular solvent for new, commercially produced scintillation cocktails. It offers the highest energy conversion efficiency of the solvents known, and has fewer restrictions on shipping and storage of a combustible liquid because of its high flash point, low volatility and lower toxicity. The primary scintillator converts excitation energy into light. Two commonly used primary scintillators are PPO (2,5-diphenyl-oxazole) and butyl-PBD (2-(4'-tert-butylphenyl)-5-(4'-biphenyl)-l,3,4-oxadiazole). The pulse voltage produced by butyl-PBD is about 20% greater than that produced by PPO, but its usefulness is reduced because of its limited solubility. Suitable concentrations of PPO and butyl-PBD are about 5 g/1 and 7 g/1 respectively. With alkaline samples butyl-PBD produces a brownish colour in the scintillation cocktail. In highly quenched samples, a higher concentration may be required for optimum efficiency. The secondary scintillator shifts light wave length and may be needed in the mixture if the emission wavelength of the primary scintillator does not match the wavelength to which the photomultiplier is most sensitive. Commonly used secondary scintillators are POPOP (1,4-bis-(2-(5-phenylloxazolyl))-benzene) and dimethyl-POPOP (1,4-bis-(2-(4-methyl-5-pheniloxazolyl))-benzene), used at concentrations varying from 0.05 to 0.5 g/1 (usually 0.1 g/l). POPOP is the less soluble of the two but is slightly cheaper.
Chapter 5
190
Both should be given adequate time to dissolve completely, preferably overnight. The best general purpose secondary scintillator is bis-MSB (p-bis-(o-methylstyryl)benzene). Bis-MSB is readily soluble in solvents and has a fast rate of dissolution. Bis-MSB is used at 0.5 to 1.5 g/1 and does not react chemically with most liquid scintillation samples, while Dimethyl POPOP can react with acids to produce a yellow to greenish colour. A suitable mixture for toluene-soluble materials is: toluene containing 5 g PPO/1 or 7 g butyl-PBD/l 0.1 g POPOP or dimethyl-POPOP/1. This mixture cannot be used for measuring the radioactivity in aqueous samples since water is immiscible with toluene. A further solvent (secondary solvent) miscible with both water and primary solvent must be added to this mixture to enable water to be incorporated. This amount of aqueous solution which can be incorporated in a particular mixture without phase separation depends on the identity of the secondary solvent, the relative proportions of primary and secondary solvents and the temperature. The secondary solvents used for this purpose are a surface-active agent or emulsifiers. Triton X-100 and Triton N-101 are nonionic surfactants commonly used in laboratory-prepared scintillation cocktails. Anionic surfactants give better sampleholding capacities than the nonionic surfactants for certain types of salt solutions. As a general rule, the greater the ratio of secondary to primary solvent, the greater the amount of water which can be incorporated, but the extra secondary solvents and the extra water increase the quenching. The properties of a good emulsifier scintillator are as follows: 1. Mixture of scintillator and water should show high counting efficiency and a low background counting rate. 2. It should show high water capacity. 3. The region of phase separation between the homogeneous liquid region and the stable emulsion or gel region should be narrow. 4. It should produce minimum chemiluminescence. 5. Mixture of aqueous solution should be stable for a long period. Mixtures having water content between 10 and 20 vol% should be avoided because they are unstable and show two phase separation. Table 5.9 shows some useful mixtures used in liquid scintillation measurements.
Table 5.9 Some useful mixtures used in liquid scintillation measurements Mixture 2
Mixture 3
PPO
6 g/l
PPO
6 g/1
DMPOP
0.1 g/1
bis-MSB
0.5 g/l
Toluene
667 ml
Xylene
700 ml
Triton X-100
333 ml
Triton N101
300 ml
Conc. HNO 3
10 I~1
Conc. HNO 3
10 ~tl
Measurements o f Radioactivi~. ,
191
The scintillation vial is the container for the analyte and the scintillator and permits light transfer from the scintillation cocktail to the photomultiplier tubes. Many materials have been used for scintillation vials. Low-potassium, borosilicate glass is the best material to meet the requirements for most liquid scintillation counting. It is non-permeable and nonreactive with chemicals used in liquid scintillation counting. It has good optical clarity for light transmission. The glass should be selected for low radioactivity content, providing a low background counting rate. Polyethylene vials are popular because of their low cost. They show a low background counting rate and good counting efficiency. The opacity of polyethylene vials makes it difficult to detect phase separation and precipitation in the sample mixture. The major disadvantage of polyethylene vials is that they are permeable to aromatic hydrocarbons. As the primary solvent permeates into polyethylene, it carries the scintillators with it, changing the background counting rate and affecting the efficiency. The radioactivity may also be carried into the vial wall leading to geometry and self-absorption losses. Other materials for vials are quartz and Teflon. They are the best materials, but they are very expensive to make and used only in applications requiring very low backgrounds and long counting times. When one wants to measure low-level radionuclides with good precision and accuracy, one has to adjust the counter settings properly and to select the most suitable scintillators and vials to ensure the highest sensitivity and reproducibility. For these purposes, the "Figure of Merit" is adopted as a quantitative criterion: Figure of Merit (FM) = (EM)Z/B,
(5.76)
where E is the counting efficiency (%), M is the mass (g) of a sample introduced in the vial and B is a background counting rate (cpm). The performances of several emulsifier scintillators prepared in a laboratory and available in the market show no significant difference in sensitivity. Some of the radionuclides commonly distributed in environmental samples include 3H, ~4C, 6Co, 898r and 9~ Tritium and ~4C emit low-energy betas which are efficiently counted by LSC. Here we present some details of sample preparation for LSC. Aqueous samples containing tritium are distilled to eliminate impurities and mixed with an emulsifier scintillator with the proportion of 40-50 vol% water. Tritium in biological samples exists as tissue free water tritium (TFWT) and organically bound tritium (OBT). Water containing TFWT is obtained by vacuum distillation (lyophilization) or azeotropic distillation using organic solvents such as toluene and benzene. The OBT is converted into HTO for counting by combusting dried samples. The tritium concentration in the natural environment is less than 10 Bq/1 and sometimes less than 1 Bq/l. Only low-background type LSCs detect these low tritium levels. To measure tritium below 10 Bq/l with good accuracy, using the conventional type LSCs, it is necessary to enrich the tritium in a large water sample by electrolysis. This procedure can increase the tritium concentration 20-30-fold. Carbon-14 exists mostly as carbon dioxide (CO 2) and as organic compounds in the environment. Carbon dioxide in the atmosphere or in the effluent gas from a
192
Chapter 5
combustion system for biological samples is first absorbed in alkaline solutions such as aqueous NaOH or NH4OH. If necessary, calcium chloride or barium chloride is added to the alkaline solution to precipitate C a C O 3 or B a C O 3, which are then purified and stored in a sealed bottle for future analysis. The alkaline solution containing carbon dioxide or carbonate is acidified by titrating with a strong acid solution to generate carbon dioxide. The carbon dioxide gas is then absorbed in a solution of organic amine absorber and liquid scintillator mixture (1:1). The procedure for 6~ measurement is as follows: after adding cobalt salt as carrier to sea water, 6~ is first precipitated as cobalt hydroxide in alkaline solution. The precipitate is separated from sea water and dissolved in HC1 solution. 6~ is separated and recovered in the effluent from an anion exchange column. The effluent containing 6~ is evaporated to dryness. The residue is redissolved in dil. HC1 solution and transferred into a counting vial. The emulsifier scintillator is added into the vial and mixed well. The counting sample is counted by means of a low background LSC. This technique can be applied to 6~ in biological samples by using an ashing procedure beforehand. Strontium isotopes can be measured by utilizing the fact that high energy beta rays in aqueous solution emit Cerenkov photons which can be counted by photomultipliers in a LSC without any phosphors. Strontium radioisotopes in a sea water sample are separated and purified by precipitation of strontium carbonate and subsequently strontium oxalate precipitate followed by the cation exchange. Yttrium-90, the daughter nuclide of 9~ is scavenged from the solution by Fe(OH)3 co-precipitation. The acidified aqueous solution containing strontium isotopes is transferred into a counting vial and counted after the total volume of the solution is adjusted by adding distilled water. By setting the lower level of the discriminator properly, the LSC can count only Cerenkov light signals produced by 1.49 MeV beta rays of 895r, without counting the 0.546 MeV beta rays of 9~ 9~ is determined by counting Cerenkov photons generated by 2.28 MeV beta rays of 90y which grows into the solution after separation.
5.3.6 Radiochemical analysis 5.3.6.1 Introduction
There is a considerable amount of attention focused on long-lived artificial radionuclides such as 9~ and 137Csreleased into natural environments. Their inventories began to grow as a result of nuclear weapon testing programs since 1950 and the nuclear power reactor accident of Chernobyl in 1986. The new sources contaminating the environment with radionuclides are recently thought to be liquid effluents from nuclear power stations and nuclear fuel reprocessing facilities. The long-lived fission products and their characteristics are given in Table 5.10 (WHO 1966; Sugihara, 1961; Norton, 1967; Chase and Rabinowitz, 1967; Harley, 1972; Greenberg et al., 1981).
193
Measurements of Radioactivi~ Table 5.10 Long-lived fission products Nuclides
Decay mode
Half-life (year)
79Se 95Kr 97Rb 9~ 93Zr 99Tc 106Ru
~13- (y) ~I]~~~- (y) I~~- (y) ~- (y) 1~- (y) ~- (y) I~- (y) [3- (y) ~~- (T) I]- (y) ~- (y)
6.5x104 10.7 4.8• l~ 28.8 1.5• 2.14• 1.00 6.5x106 14.6 55 2.7 lxl05 1.57X107 2.06 3x106 30.17 2.62 4.9
l~ 113mCd
121msn 125Sb 1265n 1291 134Cs 135Cs 137Cs 147pm 155Eu
It is not practicable to present here all the methods used; we outline only the reliable methods used for 895r and 9~ for tritium and for actinides (for details see IAEA-295, 1989).
5.3.6.2 Analysis of strontium Strontium-90 is one of the most important fission products because of its relatively high yield (about 6%), long physical half-life (29 years) and its uptake and retention in biological systems. For assessing the integrated exposure to large populations, not only the direct measurements of biological materials must be done, but also the monitoring of the transportation of the nuclide in the environment, e.g. in oceans and streams. The assay of low-level strontium-90 in biological and radiotoxicological samples requires time-consuming and laborious techniques because both the nuclide 9~ and its daughter 90y are pure beta-ray emitters. Therefore, the radiostrontium must be completely separated from other radionuclides prior to the beta-ray counting. The most commonly used method for separating strontium is by nitrate precipitation. With some modifications this method can be applied to all kinds of environmental samples and foods.
194
Chapter 5
The chemical yield varies according to the type of material. The use of 85Sr tracer to determine chemical yield is a general practice. When determining yield in this manner, it is important that the tracer is pure SSSr, i.e. free from 89Sr and 9~ Although the method is time-consuming, it is reliable and safe. Rapid methods for 9~ analysis exist, and it has been shown that they can be used after short-lived nuclides have decayed. In fresh fallout situations, the nitrate precipitation method has been shown to be more reliable, also, during periods of fresh fallout, the amount of 89Sr is of interest and the rapid methods can only analyze for 9~ In the case of higher contamination with 9~ the daughter 90y can be separated without waiting for equilibrium. Within 10 hours the activity concentration of 90y will be approximately 10% of the equilibrium value and may be sufficient for a reliable 9~ analysis (IAEA-295, 1986). A special application of liquid scintillation counters is in the measurement of Cerenkov radiation produced by beta emitters with beta energies greater than 260 keV. This application can be used for screening samples for 9~ (Carmon and Dyer, 1986; Eakins et al., 1986; IAEA- 118, 1970; Johns et al., 1979; Kleinberg and Smith, 1982; Krieger and Whittaker, 1980a, 1980b; Regan and Tyler, 1976; Szabo, 1982). An outline of the method used for determination of radiostrontium in various materials by nitrate precipitation is as follows. The ashed material is dissolved in nitric acid in the presence of strontium and barium carriers. The nitric acid concentration is then increased to precipitate all the strontium and barium (and part of the calcium) as nitrates. After further nitric acid separations, barium chromate and iron hydroxide scavenges are carried out. The subsequent treatment depends somewhat on the circumstances but the following is normal practice. Yttrium carrier is added to the purified strontium solution and, after a delay of about 14 days for the growth of 90y, the yttrium is separated, mounted and counted. The storage period for the growth of 90y can be reduced if sufficient 9~ is known to be present, and the appropriate growth factor applied. For samples of very low activity, as well as for measurement of 89Sr, strontium is precipitated from the solution remaining after the removal of yttrium and mounted for counting. In many cases the determination of the natural inactive strontium content of the material is required so that the strontium chemical yield can be corrected. In the case of milk, direct application of the nitric acid separation to a solution of the ash usually gives low strontium yields. The calcium, strontium and barium are therefore concentrated by an initial phosphate precipitation. The mixed phosphates are then dissolved in an acid and the general procedure continued from that point. In the case of cereals, and vegetation generally, the ash is very variable in composition and contains numerous elements other than calcium: a mixture of hydrofluoric and perchloric acids is necessary to decompose and dissolve the ash. After heating to remove the hydrofluoric and most of the perchloric acid, the residue is dissolved in dilute acid and the alkaline earth precipitated as phosphates. For the details of procedure, see report IAEA-295 (1989) and references therein. Here is procedure designed to separate 90y from a partially purified Sr fraction isolated from various marine samples. It has been used successfully in soil, sediment
Measurements of RadioactiviO,
195
and ashed biological samples up to 300 g in weight; and, in fresh water and sea water samples, up to 600 litres in volume. 9~ levels as low as 1 dpm have been measured in these samples. Several important facts should be kept in mind when planning 9~ analysis on marine samples. Average sea water contains 8 ppm of stable Sr, or about 8 mg/kg, while average carbonate soil is about 1% (10,000 mg/kg) Sr. Thus even modest size samples require some thought in sample handling and about chemical yield measurements, especially if stable Sr is to be used for a yield monitor. This procedure has been developed to accommodate a Sr carrier from 20 mg to 4 g. For most routine analysis, however, 85Sr, a gamma emitter readily measured by gamma spectrometry and free from 9~ contamination, is the preferred yield monitor. If stable strontium is used for chemical recovery, the strontium content of the original sample has to be determined. In average sea water the 226Racontent is about 0.2 dpm/1 and 238U(or 234U)is about 2 dpm/1. The uranium usually presents no problem, but a good separation of the final 90y from Ra and its daughters is important, especially for deep ocean water. During the separation steps, this procedure effectively removes all radionuclides that are chemically similar to yttrium or rare earths from the Sr fraction; and, after the establishment of 9~176 radioactive equilibrium, the final purification steps further remove any interfering radionuclides in the ingrowth 90y fraction before measurement by beta counting.
5.3.6.3 Analysis of tritium Tritium is measured by liquid scintillation counting of a portion of a distilled sample. Several reagents (such as sodium sulphite and silver iodide) can be added in the distillation to prevent interference by radioiodine. The allowed concentration of tritium in water for human consumption is relatively high; thus the method presented here is normally adequate for routine determinations. However, if required, lower concentrations of tritium in water can be determined by electrolytic enrichment. The principles of the tritium determination procedure are as follows. The water sample is distilled to remove non-volatile quenching materials and nonvolatile radioactive materials. Prior to distillation, sodium carbonate (Na2CO 3) and sodium thiosulphate (Na2S203) are added to the sample. The majority of the constituents that might interfere remain in the residue together with any radioactive iodide and bicarbonate that might be present. If the tritium content of non-aqueous biological samples is required, the sample can be converted into water by oxidation. The distillation is carried out to dryness to ensure complete transfer of the tritium to the distillate. An aliquot of the distillate is mixed with a scintillation solution in a counting vial. The mixture is cooled and counted in a liquid scintillation spectrometer (coincidence type). In this sample (usually an emulsion) the kinetic energy of the tritium beta particles is partly converted into light photons. When certain boundary conditions are satisfied (e.g. simultaneous detection by two or more photomultiplier tubes connected in coincidence) these photons are counted as pulses.
196
Chapter 5
Standard tritium and background samples are prepared and counted identically to minimize errors produced by aging of the scintillation medium or instrumental drift. The counting rate is a measure of the tritium activity concentration, the sensitivity (counting time 100 min) is generally of the order of 20-200 Bq/1. Details of the method are described in IAEA-295 (1986). For additional reading see Krieger and Whittaker (1980b); Budnitz (1974); Bush (1968); Fox (1976); Horrocks, (1974); IAEA-246, 1981; Konno and Suguro (1986); Lieberman (1984); Lieberman and Moghissi (1970); Momoshima et al., 1986); Volchok and de Planque (1983).
5.3.6.4 Caesium analysis For many samples one collects, the concentration of 137Cs is determined directly by gamma spectrometry. Only water samples and a small percentage of the soil, sediment, and biota samples require preconcentration of Z37Csand measurement by beta counting and/or in some cases in low background well-type Ge detectors. The radiochemical procedure for the determination of 137Cs in aqueous samples is based on the batch extraction of caesium onto a microcrystalline cation exchanger, ammonium molybdophosphate (AMP), and subsequent purification from potassium and rubidium activities by ion-exchange separation using a strongly acidic cation exchange resin (BIO-REX-40). Natural K and Rb have radioactive isotopes that interfere with the beta counting of 137Cs. The purification of caesium is also necessary to determine the chemical recovery. Gamma spectrometry is used when samples contain both ~34Csand 137Csactivities. In theory, using the proper absorbers, 134Cs can also be resolved from 137Cs activity because of the difference in beta energy. However, the absorbers greatly reduce the counting efficiency, thereby eliminating any gain in the beta measurements of samples containing both 134Csand 137Cs. The following are steps to be followed for sample preparation and preconcentration in the case of aqueous samples. If other radionuclides (Pu, Am, and Sr), are also analyzed in the same sample, the MnO 2preconcentration steps are generally performed first to remove the transuranics, followed by Cs extraction with AMP; and then, the Sr fraction is removed last as the oxalate. After the completion of MnO 2 separation, the water sample is transferred to the proper size polyethylene processing container. Adjust to pH 1--4 with nitric acid, add AMP as a slurry in water to extract the caesium (use 0.2 g AMP/1 of sample), stir the sample thoroughly and let the AMP settle, filter or decant the supernate (discard the supernate or save for Sr analysis, if required), separate the AMP by centrifugation and purify the Cs for beta or gamma counting. The steps to be followed in the sample preparation procedures used for soil, sediment, or ashed biota are: dissolve or leach the sample with c o n c . H N O 3 and separate the residual materials. The acid sample is diluted with water and adjusted to pH 2--4 with NaOH. Add 1-2 of AMP to extract the Cs from solution. Use a minimum of 1 g AMP/litre of sample. Larger amounts of AMP are required for soil, sediment, or
Measurements of Radioactivity
197
biota samples than sea water samples, because of the higher ionic strength in the acid leached sample solution, which reduces the extraction efficiency of AMP for Cs. Separate the AMP as described above for the water sample and purify as described below. The steps required for ~37Cs purification are described next. The amount of AMP collected from the preconcentration step, especially from large water samples, is invariably greater than 1 g. The following steps reduce the amount of AMP to about 1 g in order to perform the ion exchange procedure for the separation of Cs from K and Rb. 1. Dissolve the AMP with a minimum amount of 10 M NaOH. Add an 0.5 ml excel of 10 M NaOH for each gram of AMP dissolved. Transfer the sample solution to a 400 ml beaker with a few ml of water, and heat the solution on a hot plate at a medium setting, without a cover glass, to decompose the AMP and to evaporate the ammonia. 2. Periodically check the vapour phase (just above the hot beaker) with a wet pH paper or ammonium specific test paper to check for the presence of ammonia fumes. The sample solution must be kept strongly basic with NaOH for the AMP decomposition to be effective. When the vapour no longer shows the presence of ammonia, and the solution is strongly basic (pH greater than 13+), stop the heating, cool the sample to room temperature, and add water to dissolve any salts that may be formed during cooling. 3. Dilute the sample to about 200 ml with water, add 1-2 drops of methyl red indicator, adjust the pH to 1-4 with 8 N HNO 3, add 1.0 g of AMP, let the AMP settle, decant the clear liquid, transfer the AMP slurry to a clean 50 ml C-tube with water, centrifuge, and discard the supernate. 4. Dissolve the AMP with a minimum amount of 10 M NaOH and add 10 ml of 2% EDTA-07.5 NaOH solution. The sample solution should be clear. If any precipitate forms, centrifuge and decant the clear solution into another clean 50 ml C-tube. Discard the precipitate. 5. Load a medium size ion exchange column with a goose-neck adapter (see Fig. 5.43) with 20 ml of BIO-REX-40 20-50 mesh cation exchange resin. Precondition the column with 100 ml of 3 N HC1 followed by 150 ml of 5% NaC1 solution and 50 ml of water. 6. Using a Teflon coated stirring rod, carefully pour the sample solution from Step 4 directly onto the top of the resin column. Try not to splatter any solution on the upper part of the reservoir. 7. After the sample has drained to the top of the glass wool plug, rinse the column walls and the reservoir three times with about 5-10 ml aliquots of water. 8. Rinse the column with another 40 ml of water (necessary to remove any dissolved AMP from the resin and preventing the possibility of AMP reforming in the column when the acid rinse is added in the next step). Total water rinses in Steps 7 and 8 are not critical but should not exceed about 60-70 ml. Discard the rinses. 9. Wash the column with 160 ml of 0.75 N HC1 (to remove K and Rb). Discard the wash.
198
Chapter 5
Fig. 5.43. Glass columns for anion exchange separation. 10. Elute the Cs with 125 ml of 3 N HC1 and collect the sample in a 150 ml beaker. 11. Evaporate the caesium eluate to dryness and prepare the Cs for beta counting as described in the next section. Afterwards Cs samples should be prepared for beta counting following the steps: 1. Dissolve the caesium salts from Step 11 above with 1-2 drops of 8 N HNO 3 and 2-3 ml of water. Transfer the solution to a 50 ml C-tube. Rinse the beaker twice with 2-3 ml of water. 2. Add 1 ml of 10 N NaOH, dilute the sample to about 10 ml with water, and add 2 ml of 0.12 chloroplatinic acid (H2PtC16) to precipitate Cs2PtC16. 3. Cool sample in a refrigerator or ice bath for 30-40 minutes. 4. Prepare a tarred glass-fibre filter paper. (a) Assemble a filtering apparatus with a 2.54 cm base. (b) Cut a 2.54 cm diameter glass-fibre filter paper disc. (c) With the vacuum off, centre a filter disc on the base of the filter holder, wet the filter with water, apply vacuum, wash the filter with 2-3 ml of water, and 2-3 ml of acetone. (d) Dry the filter under a heat lamp. (e) Weigh the filter to _0.01 mg. 5. Filter the sample through the tarred filter from Step 4. 6. With the vacuum on, remove the filter chimney, wash the filter and the Cs2PtC16 thoroughly with a few ml of cold acetone. Turn off vacuum, transfer the filter and precipitate to a petri dish. 7. Dry the filter under a heat lamp and cool the filter to room temperature.
Measurements of RadioactiviO,
199
Fig. 5.44. Electroplating cell.
8. Weigh the sample to constant weight (_+0.01 mg). 9. Mount the filter containing the Cs2PtC16 on the ring and disc holder as shown in Fig. 5.44. Count the sample using a low background beta detector or by gamma spectrometry.
5.3.6.5 Determination of actinides Actinides in the environment can be classified into two groups: (i) the uranium and thorium series of radionuclides in the natural environment and (ii) neptunium, plutonium, americium and curium which are formed in a nuclear reactor during the neutron bombardment of uranium through a series of neutron capture and radioactive decay reactions. Transuranics thus produced have been spread widely in the atmosphere, geosphere and aquatic environment on the earth, as a result of nuclear bomb tests in the atmosphere, and accidental release from nuclear facilities (Sakanoue, 1987). Most of these radionuclide inventories have deposited in the northern hemisphere following the tests conducted by the United States and the Soviet Union. In actinide series, the elements of greatest interest as environmental contaminants are neptunium, plutonium, americium, and curium, because their presence at relatively high concentrations in ecosystems would represent potential health problems (Katz et al., 1986). Nuclear data for actinide isotopes are presented in Table 5.11.
200
Chapter 5
Table 5.11 Nuclear data of actinide isotopes Nuclide
a-Energy (MeV)
Yield (%)
Half-life (years)
244Cm 242Cm 243Am 241Am
5.81, 5.76 6.11, 6.07 5.28, 5.23 5.49, 5.44 4.90, 4.86 5.16, 5.12 5.16, 5.14, 5.10 5.50, 5.46 5.77, 5.72 4.79, 4.77 4.20, 4.15 4.40, 4.37 4.77, 4.72 5.32, 5.26 4.02, 3.96 4.69, 4.62 4.90, 4.85 5.42, 5.34
77, 23 74, 26 88, 11 86, 13 74, 26 76, 24 73, 15, 12 71, 29 69, 31 51, 36 77, 23 57, 18 72, 28 69, 31 77, 23 76, 23 11, 56 73, 27
1.81xl0 6.09x 106 7.37x103 4.32x102 3.76x105 6.57x103 2.41x104 8.77x 10 2.85 2.14x106 4.47x 109 7.04x 105 2.45x105 7.18x 10 1.41x 101~ 8.03x104 7.34x103 1.91
242pu 24~
239pu 238pu 238pu 237Np 238U 235U
234U 232U 232Th 23~ 229Th 228Th
As a result of nuclear bomb tests in the atmosphere, accidental release from nuclear facilities, and the accidental fall of artificial satellite SNAP 9A, plutonium isotopes have been spread widely in the earth' s atmosphere, geosphere and aquatic environment from the 1960s. They include 241pu, 24~ 239pu, and 239pu (Harley et al., 1973). These plutonium isotopes in the geosphere and the aquatic environment are incorporated metabolically into plants and ultimately find their way into man through the food chain; these radionuclides in the atmosphere are also incorporated into man directly by inhalation. From the standpoint of a dose assessment, these radionuclides are important because they are mostly alpha-emitting radionuclides and have half-lives of 13-2.4x104 years. From the standpoint of the safety assessment of the disposal of high-level radioactive waste, it is important to clarify the migration behaviour of plutonium in ground strata. Consequently, it is essential to know the plutonium concentrations in food and environmental samples for these studies. The commonly applied methods for determination of actinides in environmental samples may be classified as follows. 1. Preparation: (a) drying, (b) ashing, (c) scavenging.
Measurements of Radioactivi~'
201
2. Solubilization and equilibration: (a) fusion, (b) leaching. 3. Concentration and separation: (a) coprecipitation, (b) ion exchange, (c) solvent extraction. 4. Electrodeposition and alpha spectrometry. For a description of a procedure for plutonium separation in large volumes of flesh and saline water by manganese dioxide coprecipitation, see Wong et al. (1975). Among ion exchange separation methods for transuranics, strong base anion exchange in hydrochloric and nitric acids is important (Keder, 1962; Keder et al., 1960; Horowitz et al., 1990; Chu, 1971; Wong, 1971; Diamond et al., 1954; Korkisch, 1989). Among solvent extraction reagents for transuranics, thenoyltrifluoroacetone (TTA) and trioctylamine (TOA) are important (Keder, 1962; Chieco et al., 1990). Each transuranic element has many valencies and their behaviour in aqueous solution is very complicated because of disproportionation reactions. As stated above, the ion-exchange and solvent-extraction behaviours of transuranics are dependent on their valency state. Therefore valency control is very important in their analysis (Katz et al., 1986). Additional references discussing this problem include Karkisch (1989), Diamond et al. (1954), Stevenson and Nervik (1961), Abuzwida et al. (1987), Bernabee et al. (1980), Budnitz (1973), Chu (1971), Fukai et al. (1976), Hampson and Tennaut (1973), Hindman (1986), Holm et al. (1979), Irlweck and Veselsky (1975), Jiang et al. (1986), Johns (1975), Scott and Teynolds (1975), Sekine et al. (1987). Procedures are described for the determination of plutonium and americium in environmental samples by anion exchange (HNO3). Procedures are also described for the determination of plutonium, americium and their sequential analysis by anion exchange (HNO3) and TOA extraction (Chieco et al., 1990). Livens et al. (1989) developed the method for the determination of Am in environmental samples (Fig. 5.45). Yamamoto et al. (1989) developed the method for the determination of Np in environmental samples (Fig. 5.46). There follows a description of plutonium separation by anion-exchange. Plutonium as well as several other heavy elements (e.g. U, Th, Np, Am, Cm), may be separated and purified by anion-exchange chromatography. If only 239+24~ activity is to be determined, a single anion-exchange column separation, done carefully, is usually sufficient. However, if the 23Spu activity is also required, then a second small column separation will be necessary to eliminate any trace of interfering activities from the naturally occurring elements, Ra and Th, present in all soil and sediment samples. The steps to be followed in plutonium purification are: 1. Prepare anion-exchange columns as shown in Fig. 5.43. Use the medium size column for sample volumes less than 500 ml or the large size column for samples with final volumes greater than 500 ml.
Chapter 5
202 Sample leachate (HCI)
Fe solvent extraction (diisopropyl ether)
Anion exchange (HC1)
~ Pu fraction
Cation exchange (HC1)
BiPO~ or Fe(OH)3 precipitation
Anion exchange (HNO3 - CH3OH)
Anion exchange (HNO3)
~ Am fraction
Fig. 5.45. Flowsheet for Pu/Am method (after Livens and Singleton, 1989).
2. Precondition the columns by passing about 20 ml of 8 M H N O 3 through each column. 3. Check the sample solution for particulate material. If insoluble silicate (white, gelatinous materials) or other particles are noted, centrifuge or filter the solution through a glass fibre filter before loading the sample onto the preconditioned column. This step is important in order to maintain a continuous flow of the sample solution through the column. 4. Load the sample solution onto the column as fast as the column will allow. The plutonium is retained on the ion-exchange column. If insoluble material has been removed from the sample and there is no outgassing in the column, the average flow rate in the medium size column is about 3-6 ml per minute, depending on the viscosity of the sample solution. 5. After the sample has drained to the top glass-wool plug in the column, rinse the wall of the column reservoir thoroughly with 8 M H N O 3. This can be done conveniently with a polyethylene wash bottle. Do this at least 3 times with about 5-10 ml of 8 M H N O 3 and allow each rinse to drain to the top glass-wool plug. 6. After the last rinse, wash the column-volumes of 8 M H N O 3 and discard the rinse (about 150 ml for the medium column or 350 ml for the large column). 7. After the 8 M H N O 3 rinse, wash the wall of the column reservoir with about 5-10 ml of conc. HC1. Repeat the conc. HC1 wash three times, each time draining the liquid to the top of the glass-wool plug. 8. Now wash the column-volumes of conc. HC1 (about 150 ml for med. column or 350 ml for the large column). Discard HC1 rinse.
Measurements of Radioactivio,
203
Soil or sediment sample
23'~Nptracer Aqua-regia
Heat
1
I
Res.
Sup.
I
Evaporate to dryness I 10M HCI Isopropyl ether extraction
1 Aq.
} I
Sup. (Pa, U)
I
ppt.
(LaF3) ] 8M HNO3saturated with AI(NO3)3
'--[Anionexchange ] i I Feed solution ~ I I ~ 8MHNO3 ~ ~ ~ 10MHC1 ~ ~ ~
(Np)
4M HC1 ~
~
I
I Evaporate to dryness
I
(Fe)
10% TOA-xylene Extraction (x3) I
I
i
Org.
(Th, Pu, Aq. Am)
I 10M HCI wash
I M HC1-0. 1M HF strip Continued to Pu and Am separationprocedure
NH2OH x HC1 Conc. H2SO4 La carder
(U) (Th) (Pu)
10MHC1-0.1MHI ~
Org. 0.05M HI (Heat)
Dewox l xS, 100-200 mesh
1Anion exchange I
e .so tion
4MCH3COOH
~
L[
~
.... ] (Np)
I I
Electroplate from 2M NH4C1solution Plated NI2
I
r-ray spectrometry
I o~-ray spectrometry
Fig. 5.46. Separation scheme of neptunium for environmental soil or sediment sample (after Yamamoto et al., 1989).
9. Elute the Pu with 5 column-volumes of NH4I-HC1 solution. Use 100 ml of NHnI-HC1 for the medium size column. Collect the Pu eluant in a 150 ml beaker. 10. Add approximately 5 ml of c o n c . H N O 3 to the Pu eluant, mix, and evaporate to dryness on a hot plate. A small residue may be visible on the bottom of the beaker after the evaporation, which appears to have no significant effect on the analysis. This material is normally seen and probably results from the decomposition products of strong acids with the resin. The following is a the description of plutonium purification by anion-exchange separation. 1. Cool the sample to room temperature. Rinse the wall of the beaker with 8-10 ml of conc. HC1. Add 2-3 drops of 30% H:O 2, 3-5 drops of 1 M NaNO 2, and heat on a hot plate for about 5 minutes. Again, cool the sample to room temperature. 2. Prepare a small sample with Dowex 1x8.50-100 mesh and precondition it with 1 ml of conc. HC1. Collect the eluant in a 50 ml C-tube. 3. Load the sample from the beaker onto the column. 4. Rinse the beaker 2 times with 2-3 ml of conc. HC1 and load the rinse onto the column.
204
Chapter 5
5. When the liquid has drained to the top glass-wool plug of the column, rinse the wall of the reservoir with 2-3 ml of conc. HC1. 6. Rinse the column 2 more times with 2-3 ml of conc. HC1. 7. Wash the column with 10 ml of conc. HC1. 8. Discard the rinse from Step 2 through Step 7. 9. Elute the Pu with 20 ml of NHaI-HCI solution. Collect the Pu in a 50 ml glass beaker. 10. Add 2-3 ml of conc. HNO 3and evaporate the solution to dryness on a hot plate. The steps to be followed in the implementation of the electrodeposition of plutonium procedure are: 1. Assemble the plating cell. 2. Fill the cell with water to test for leakage. 3. Add 1 ml of conc. sulphuric acid to the sample. 4. Heat the sample on a hot plate until copious white fumes evolve. 5. Cool the sample to room temperature, carefully rinse the wall of the beaker with 2 ml of 1 N H2SO 4. Add 2 drops of 0.1% methyl red indicator. 6. Transfer the sample to the electroplating cell. 7. Rinse the beaker with 2 ml of 1 N H2SO 4. Add the rinse to the plating cell. 8. Repeat Step 7 three times. 9. Add conc. NH4OH dropwise until the colour of the sample changes from red (pH 4.4) to yellow (pH 6.2). Mix the solution by swirling the plating cell. 10. Add 1 N H2SO 4 dropwise to the red end point (pH 4.4) and then add 2 drops excess. 11. Complete the assembly of the electroplating cell by attaching the platinum anode with plastic insulation tape. Position the platinum wire anode about 0.5 cm from the stainless steel disc (cathode). 12. Connect the anode and cathode of the electroplating cell to a constant current power supply. 13. Electroplate at 1.0 amp for 60-70 minutes. 14. Before turning off the power supply when electrodeposition is completed, add 1 ml of conc. NHnOH to the cell and continue plating for about one minute then turn off the power supply and as quickly as possible: (a) Disconnect the cell from the power supply. (b) Discard the solution from the cell. (c) Rinse the cell with diluted NHaOH from a wash bottle. (Make the diluted NH4OH solution by adding 0.5 ml conc. NHaOH to 500 ml of water) (d) Disassemble the plating cell. (e) Rinse the plated disc with diluted NHaOH. (f) Rinse the plated disc with acetone and let the disc air dry on a clean paper tissue. 15. Count the plated disc and determine the activity of plutonium isotopes by alpha spectrometry.
Measurements of Radioactivity
205
5.3.7 Rapid methods The need for rapid methods is apparent in accident and similar situations. An impulse to the development of rapid methods is also provided through the so-called coordinated research program of IAEA, Vienna. Here we shall present some of the methods reported in a Research coordination meeting on Rapid Methods held in Vienna in 1991. As an example of rapid methods, we mention here the work by Brodzinski and Perkins (Brodzinski and Perkins, 1992). They have described a completely portable instrument, operable by one man, which is capable of quantifying the radioactive content of drums. Eleven radioisotopes are measured simultaneously in just a few minutes. The assayer uses two measuring techniques: segmented 7-ray spectrometry and neutron counting. A drum (or other container) to be assayed is placed on a rotating turntable by a self-contained electric hoist. A collimated high-resolution germanium y-ray spectrometer vertically scans the rotating drum to measure the intensity of y-rays as a function of the energy emanating from the drum. Most fission and activation products and some transuranic radionuclides emit measurable quantities of monochromatic photons that serve as "fingerprints" of those radiosiotopes. Comparison with emission rate from known standards provides a quantitative measure of radioactivity from each y-ray emitter in the drum. This germanium spectrometer is used to measure the bremsstrahlung radiation from 9~ By manipulating the software with the on-board computer, the intensity of the 9~ bremsstrahlung in the assayed drum is also compared to that of standards, and the 9~ concentration is quantified. The reported sensitivity for transuranic radionuclides is approximately 1 nCi/g, while that for gamma emitters is of the order of 0.1-1 pCi/g. Also, based on the bremsstrahlung radiation measurement, 9~ can be determined at concentrations of 100 pCi/g.
5.3.7.1 Rapid determination of transuranic elements and plutonium The rapid methods are based on fast removal of the transuranic elements from interfering materials so that they can be electro-deposited as a group, and measured by alpha energy analysis. The procedure involves the following basic steps (Thomas, 1991). 1. The sample is first brought into solution. 2. Radioisotope tracers, including 242,-, r'u, 243Am ' and 234Th (if appropriate), are added. 3. A small amount of Fe carrier (10 mg) plus sodium sulphite is added to this solution, which is subsequently made basic by addition of ammonium hydroxide to allow the formation of an iron hydroxide precipitate. This precipitate serves to carry the thorium and the transuranic elements. 4. The mixture is centrifuged and the solution discarded: the precipitate is dissolved in dilute hydrochloric acid then diluted with water. It is then made basic with ammonium hydroxide, which results in a second iron hydroxide precipitate forming.
206
Chapter 5
5. Following centrifuging and discarding of the solution, the precipitate is dissolved in dilute hydrochloric acid, diluted with water, and a small amount of sodium sulphite added to maintain the transuranic elements in their lower valence states. 6. Small amounts of calcium and oxalic acid are then added and the pH is adjusted to approximately 3 to allow formation of an oxalate precipitate. The iron forms a very soluble oxalate, thus remaining in solution. This and two subsequent oxalate precipitations serve to remove any remaining iron. 7. The final oxalate precipitate is then dissolved in a small amount of sulphuric acid (0.5 ml of concentrated H 2 5 0 4 ) , and the pH adjusted using dilute ammonium hydroxide. 8. The solution is then placed in an electro-deposition cell, where the transuranic elements are electro-deposited on a 1 cm 2 area of a 2.5 cm diameter stainless steel disc. 9. Electro-deposition is conducted for a 1-h period at a current of 1 A. 10. Immediately before turning off the current, 1 ml of concentrated NH4OH is added to the cell and the electro-deposition continued for an additional minute. The current is then turned off, the solution discarded, and the electrode washed with water, then ethanol and air-dried. Following alpha energy analysis, the radiochemical yield, as determined from the radioisotope tracer content and the concentrations of the radionuclides of interest, are calculated. Samples with a large amount of iron such as soil extracts or vegetation ash may require partial removal of iron prior to initiation of this procedure (Thomas, 1991).
5.3.7.2 Rapid determination of 9~ Up to now, two different approaches have been used successfully in fast radiochemical separation procedures for the determination of strontium-90 in environmental sample: 9 investigations dealing with the extraction of yttrium-90; 9 investigations which tailor precipitation methods for the strontium-90 separation to the needs of special sample types. We shall mention only some of the work in these fields. Vajda et al. (1991) have described a simple and rapid method for the separation and successive determination of total radiostrontium in soil by using a crown ether. The method consists of three basic steps: oxalate precipitation to remove bulk potassium, chromatographic separation of strontium from most inactive and radioactive interferences utilizing a crown ether, oxalate precipitation of strontium to evaluate the chemical yield. Radiostrontium is then determined by liquid scintillation counting of the dissolved precipitate. When 10-g samples of soil are used, the sensitivity of the method is about 10 Bq/kg. The chemical yield is about 80%. The separation and determination of radiostrontium can be carried out in about 8 h. Another method for 9~ determination in food and environmental samples has been described by Shuzhong et al. (1991). It is based on the use of a tributylphosphate for
Measurements of Radioactivio,
207
extraction of 90y, the daughter of 9~ The method is shown to be sensitive to 0.2 Bq per kg of dry grass and milk powder and 2 Bq per kg of soil. In the case of a nuclear accident, most of the radioisotopes in the environment and food can be reliably and quickly assayed by gamma spectroscopy. There is a problem with some important isotopes which are pure beta or alpha emitters and which cannot be identified directly by gamma spectroscopy. The activity of the isotopes of the strontium group 89Sr, 9~ and 91y after a three-year reactor fuel cycle can reach about 8% of the total in-core activity and one of them, 9~ is important for the long-term health consequences. It has been shown (Vapirev and Hristova, 1991) that the Ba/Sr reactor core ratio can be used for estimation of the upper limit of strontium activity in the fallout immediately after an accident. The Cs/Sr ratio can be used for estimation of the strontium fallout in the late post-accident period.
5.4 N O N - R A D I O M E T R I C M E T H O D S
There is a variety of situations in which it is better to determine the concentration of a radionuclide by a mass measurement rather than by measuring the activity present. This approach is possible using a wide range of instrumental methods of non-radiometric elemental analysis; analytical measurements can be performed also by elementspecific chemical methods, some of which are extremely sensitive. The most important criterion for selecting an analytical method is whether the technique is sufficiently sensitive to measure the amount of radionuclide present in the sample. This is a very different problem when considered from the viewpoint of analytical chemists who use radiometric methods and those who use non-radiometric methods. Limits of detection in radiometric methods can be as low as 10-4 Bq, although -1 mBq is a more generally attainable detection limit. For non-radiometric methods, the detection limit is expressed in terms of mass and the relationship between radiometric and non-radiometric limits of detection will depend upon the half-life of the radionuclide of interest. Table 5.12 (from McMahon, 1992) shows the mass corresponding to 1 mBq for a number of important radionuclides. Although 1 mBq of 232Th is readily measured by a number of nonradiometric methods, 1 mBq of 137Cscould only be detected by the most sensitive of methods and is probably best determined radiometrically. Non-radiometric methods offer a variety of features and their use may be favoured for reasons other than improved sensitivity or isotopic selectivity. They can, in some instances, be used to perform analyses with less sample preparation and greater speed or sample throughput, and allow remote analysis or provide elemental or isotopic maps or depth profiles (McMahon, 1992). The instrumental methods of elemental analysis can be conveniently grouped as follows:
Chapter 5
208
Table 5.12 The mass of 1 mBq for a selection of radionuclides with a variety of half-lives (after McMahon, 1992) Radionuclide
Half-life (years)
Mass of 1 mBq (g)
232Th 238U
1.4x 101~ 4.5 x 109
2.5x 10-7 8.1 x 10-8
1291 99Tc 239pu
1.7x107 2.1x105 2.4x 104
1.6x10 -l~ 1.6x10 -12 4.3x I 0 -13
14C 137Cs
5.8x103 30
6.1x10 -15 3.1x10 -16
(i) Methods based on X-ray fluorescence analysis
X-ray fluorescence analysis Total reflection X-ray fluorescence analysis Electron microprobe analysis Particle induced X-ray emission Synchrotron radiation induced X-ray emission
(XRF) (TXRF) (EMA) (PIXE) (SRIXE)
(ii) Methods based on ultraviolet or visible spectroscopy
Atomic absorption spectroscopy Graphite furnace AAS Atomic fluorescence spectroscopy Inductively-coupled-plasma optical-emission spectroscopy Glow-discharge optical-emission spectroscopy Laser-excited resonance ionization spectroscopy Laser-excited atomic-fluorescence spectroscopy Laser-induced-breakdown spectroscopy Laser-induced photocoustic spectroscopy Resonance-ionization spectroscopy
(AAS) (GFAAS) (AFS) (ICPO-ES) (GC-OES) (LERIS) (LEAFS) (LIBS) (LIPAS) (RIS)
(iii) Methods based on mass spectrometry
Spark-source mass spectrometry Glow-discharge mass spectrometry Inductively coupled-plasma mass spectrometry Electro-thermal vaporization-ICP-MS Thermal-ionization mass spectrometry Accelerator mass spectrometry Secondary-ion mass spectrometry Secondary neutral mass spectrometry Laser mass spectrometry Resonance-ionization mass spectrometry Sputter-initiated resonance-ionization spectroscopy Laser-ablation resonance-ionization spectroscopy
(SSMS) (GDMS) (ICP-MS) (ETV-ICP-MS) (TIMS) (AMS) (SIMS) (SNMS) (LMS) (RIMS) (SIRIS) (LARIS)
209
Measurements of Radioactivity
Table 5.13 Analytical techniques classified by amount of isotopic information and amount of sample required (after McMahon, 1992)
No isotopic information Minor isotope determination Trace isotope determination
Bulk samples
Small samples profiling
Imagingand depth
ICP-OES
XRF, GFAAS, LEAFS, TXRF ETV-ICP-MS
PIXE, SRIXE
ICP-MS, GDMS, SSMS
SIMS, SNMS, LMS, SIRIS
TIMS, RIMS, AMS
McMahon (1992) has reported on intercomparison of non-radiometric methods for the measurement of low levels of radionuclides. He has classified the above analytical techniques according to the amount of isotopic information and the amount of sample required. The conclusions are presented in Table 5.13.
5.4.1 Methods based on X-ray spectrometry The electronic transitions which give rise to X-ray emission spectra involve core electrons and are therefore relatively insensitive to the chemical and physical form of the determinant (Bertin, 1978). As a result, analyses can be performed with a minimum of sample preparation directly on materials in the condensed phase. This insensitivity of sample matrix applies to the wavelength of the emitted X-rays, not to their intensities and as quantitation is based on intensity measurement, closely matched standards are required. X-ray emission spectra can be excited by primary X-rays in a fluorescence experiment or by changed particles via collisional excitation. The cross sections for excitation of X-ray emission are rather low and this is combined with the low efficiency of collection, collimation, diffraction and detection of the emitted X-rays. This low overall efficiency leads to a relatively low sensitivity in some cases and is compounded by high backgrounds either from scattered primary radiation in a fluorescence experiment or due to bremsstrahlung in the charged-particle-excitation methods. Methods based on X-ray spectrometry do not provide isotopic information about the sample. Nonetheless, there are a number of radio analytical problems which can be solved by methods based on X-ray spectrometry. The following instrumental methods of elemental analysis are based on X-ray spectrometry. XRF X-ray Fluorescence Analysis TXRF Total Reflectance X-ray Fluorescence Analysis EMA Electron Microprobe Analysis PIXE Particle Induced X-ray Emission SRIXE Synchrotron Radiation Induced X-ray Emission
210
Chapter 5
XRF is the simplest of these methods. It allows bulk analysis of solid or liquid samples with detection limits of approximately 0.1 ~tg. The method can thus only compete with radiometric methods for the longest lived of radionuclides. It has approximately the same sensitivity for 232Th as alpha spectrometry but has the advantage that little sample preparation is required and that analysis is rapid and easily automated. XRF would be the method of choice for measurement of airborne thorium collected onto filter papers, for example. The more sophisticated methods address the problem of the low overall efficiency of generation and acquisition of the X-ray spectrum. The low fluorescence cross section is addressed by using a highly intense X-ray source, a synchrotron in the SRIXE method. The high intensity of synchrotron X-rays allows the beam to be focused and collimated whilst retaining significant intensity. The method can therefore be used in a microprobe mode and by moving the sample in a raster pattern across the incident X-ray beam, elemental images can be generated with micron spatial resolution. The scattered primary radiation background can be reduced by using the total-reflectance technique in TXRF (Knoth and Schwenke, 1978). The instrumental geometry limits scattering of primary X-rays in the direction of the detector, however this is at the expense of increased sample preparation. The gains in sensitivity achieved by each of these methods may be compounded in a method which uses a total reflectance sample geometry in combination with a synchrotron X-ray source. The charged-particle-beam methods EMA and PIXE also allow elemental imaging within the sample. These methods generally require that the sample be enclosed in a vacuum. The approximately 15 keV electrons used in an EMA instrument, penetrate only 1-2 ~tm into the sample. This rapid declaration of the charged particles generates bremsstrahlung X-rays which generate a strong background signal in the spectral region of interest. EMA thus has relatively poor detection limits. The method can be used for analysis of electrodeposits such as sources prepared for alpha-particle spectrometry where the element of interest is present at high concentration in a very thin surface layer. The approximately 2.5-MeV-proton beam used in PIXE analysis penetrated much deeper into the sample than the EMA electron beam. The resulting proton bremsstrahlung is less intense and backgrounds are therefore reduced. PIXE can thus achieve much lower detection limits. PIXE (Johansson and Campbell, 1988) and SRIXE (Jones and Gordon, 1989) have similar imaging capabilities and detection limits but both suffer from the drawback that they rely on major pieces of hardware, an accelerator in the PIXE experiment and a synchrotron X-ray source for SRIXE. 5.4.2 Methods based on ultraviolet-visible spectroscopy
Atomic spectroscopy in the ultraviolet-visible region involves transitions of valence shell electrons and the spectra are thus sensitive to the chemical and physical form of the element of interest. For sensitive quantitative work the sample is normally converted to free atoms in the gas phase. This can be achieved by vaporization from a furnace, by
Measurements of Radioactivit3,
211
aspiration of a solution into a flame or inductively coupled plasma, by sputtering in a glow discharge or by laser ablation. Producing free, gas phase atoms is a particular problem for thorium, uranium and plutonium as these elements react with traces of oxygen, even in a high vacuum system, to give oxides and dioxides. The following methods are based on types of atomic spectroscopy in the ultraviolet visible region: Atomic Absorption Spectroscopy AAS Graphic Furnace AAS GFAAS Atomic Fluorescence Spectroscopy AFS Inductively-coupled-plasma optical-emission spectroscopy ICPOES Glow-discharge optical-emission spectroscopy GDEOS Laser-excited atomic-fluorescence spectroscopy LEAFS Laser-induced-breakdown spectroscopy LIBS Resonance-ionization spectroscopy RIS The methods range from simple, inexpensive absorption spectroscopy to sophisticated tunable-laser-excited fluorescence and ionization spectroscopies. AAS has been used routinely for uranium and thorium determinations (see for example Pollard et al., 1986). The technique is based on the measurement of absorption of light by the sample. The incident light is normally the emission spectrum of the element of interest, generated in a hollow-cathode lamp. For isotopes with a shorter half life than 23Su and 232Th, this requires construction of a hollow-cathode lamp with significant quantities of radioactive material. Measurement of technetium has been demonstrated in this way by Pollard et al. (1986). Lawrenz and Niemax (1989) have demonstrated that tunable lasers can be used to replace hollow-cathode lamps. This avoids the safety problems involved in the construction and use of active hollow-cathode lamps. Tunable semiconductor lasers were used as these are low-cost devices. They do not, however, provide complete coverage of the spectral range useful for AAS and the method has, so far, only been demonstrated for a few elements, none of which were radionuclides. Absorption spectroscopy measures the difference in intensity between an incident and transmitted signal. Lower detection limits can be potentially obtained by monitoring a single low-intensity signal, as in emission or fluorescence spectroscopy. LEAFS uses tunable lasers to efficiently excite fluorescence and, by passing the sample atoms repeatedly through excitation-fluorescence cycles, very high sensitivities can be obtained. Again, LEAFS has been demonstrated for only a limited number of elements, none of which were radionuclides. A particularly sensitive approach is to excite fluorescence by a two photon process. In this way the wavelength of the fluorescent light is much shorter than that used to excite fluorescence and scattered primary radiation can easily be discriminated from the fluorescent signal. As an alternative to observing the fluorescent signal in a LEAFS experiment, the state which has been resonantly excited by a tunable laser can be further excited by further laser photons to produce an ion. Ions can be collected and detected with electron multipliers with high efficiency leading to the extremely high sensitivity of ionization spectroscopies. LEAFS and RIS combine the high selectivity of laser spectroscopy with high sensitivities. Both these components are required to give low detection limits.
212
Chapter 5
Isotope effects are observable in high resolution ultraviolet-visible spectroscopy. At very low atomic numbers and at very high atomic numbers (especially for the actinides) isotope shifts can be observed and coupling of the electron spin with the nuclear spin in odd-mass-number elements. Ultraviolet-visible spectroscopy can thus potentially provide isotopic information in these regions of the periodic table although routine methods are not yet available. For those elements where isotopes cannot be distinguished from their simple atomic spectra an isotope-specific resonance ionization method has been suggested by Letokhov (1987). An instrument is being developed, based on this suggestion, for the determination of 9~ in which strontium ions are accelerated to an energy of about 50 keV and neutralized collisionally. At this kinetic energy, the different strontium isotopes are travelling at sufficiently different velocities that collinear resonance ionization spectroscopy can differentiate between isotopes on the basis of their different doppler shifts (Monz et al., 1993). Thus some methods of atomic spectroscopy can provide isotopic information (Hurst and Payne, 1988).
5.4.2.1 Inductively coupled plasma-optical emission spectrometry (ICP-OES) Atomic spectroscopy is widely used in inorganic chemistry to determine total element concentrations in many sample types, and generally allows rapid sample throughput. The optical techniques allow determination of atomic concentrations down to sub ng/ml levels (10 -8 M and below) in samples of a few millilitres or less. The recent introduction of a new mass spectrometric technique allows isotope-specific measurements to be made with the ease of use and sample throughput of the atomic spectroscopic techniques. The ICP is a stable argon plasma heated by inductive coupling of argon cations and free electrons, but is perhaps best thought of as simply a hot flame. Temperature measurement indicates that the plasma has a temperature approaching 6-7000~ Samples in solution are nebulized (at about 0.4 ml min -~ solution consumption) to produce an aerosol of fine droplets. A spray chamber is used to select only the smallest droplets for analysis, in practice those below approximately 5 micron. The selected droplets are swept into the centre of the plasma by an argon stream. In the plasma, droplets undergo rapid heating causing firstly desolvation of droplets, and then breakage of molecular bonds. The resulting free atoms are electronically excited; many are ionized. As atoms leave the plasma and cool, they relax leading to emission of light. Detection of this light is the basis of ICP-OES. The wavelengths emitted are characteristic of the elements present, and the intensity proportional to their concentrations. ICP-OES limits of detection for many elements lie in the range 1-100 ng/ml (ppb) in solution. A few elements, notably Li, Be, Mg, Ca, Sc, Ti, Mn, Cu, Sr, Y and Ba, have limits less than 1 ng/ml. ICP-MS on the other hand generally is more sensitive (by 1-3 orders of magnitude) and gives isotope specific information; ICP-OES to a first approximation only gives total element concentrations. The principal advantages of the technique is that it is multi-element and that data acquisition takes approximately 1 min with a changeover time between samples of a similar order. The technique has drawbacks:
Measurements of Radioactivit3,
213
spectral interference is possible, depending on other elements present; therefore in ICP-OES a high resolution optical spectrometer may be required. The technique is best suited to solution, although direct solid sampling techniques are being developed. 5.4.2.2 Laser excited resonance ionization spectroscopy (LERIS)
To achieve both high isotopic selectivity and high sensitivity at the same time, collinear laser spectroscopy is combined with resonance ionization. The principle of resonance ionization spectroscopy is the following: the atoms are excited by one or several resonant optical excitation steps into an energetically high-lying state. Subsequently the atoms are ionized either by absorption of another photon, by collisions, or by field ionization. The photo-ions produced in this process can then be detected with high efficiency. This technique has proved to be extremely useful and sensitive in numerous applications. For the combination of collinear fast beam laser spectroscopy with resonance ionization detection, the excitation into high-lying Rydberg states with subsequent field ionization is best suited because of the effective suppression of background. This technique has already been successfully applied for trace analysis of 3He in environmental samples as well as for the sensitive study of radioactive Yb-isotopes at the on-line mass separator facility ISOLDE at CERN. Monz et al. (1993) have described the use of LERIS for low level detection of 9~ and 895r in environmental samples. The experimental set-up is shown schematically in Fig. 5.47. After chemical separation from the environmental sample, the Sr is inserted into the ion source. The ions are accelerated to an energy of 30 keV and pass through the mass separator, where the stable isotopes are strongly suppressed. The 895r or 9~ ions enter the apparatus for resonance ionization in collinear geometry and are deflected by 10 ~ to enable collinear superimposition of the laser beam. Neutralization takes place inside a charge exchange cell filled with caesium vapour. Remaining ions are removed afterwards from the resulting fast atomic beam by different electrostatic deflectors. Subsequently, the selective excitation into high-lying Rydberg states is induced by the laser light. The Rydberg atoms are field-ionized in a longitudinal electric field and the ion source
L. e ector separator magnet
"~n.tll~~ neutralization -" L
_ /
optical excitation
ion
ion filter
~
energy
detector
t.d field ionization
Fig. 5.47. Simplified scheme of the experimental set-up for detector of 89'9~ by collinear fast beam laser excitation and resonance ionization detection.
214
Chapter5
resulting ions are deflected out of the atomic beam for counting with a particle detector. A total isotope selectivity of 885r/9~ > 10 ~~and an overall efficiency of 5x 10-6 have been achieved. With these values, a detection limit of l xl08 atoms of 9~ in the presence of more than 10 j7 atoms of stable isotopes has been demonstrated. The trace determination of such a contamination can be carried out with an accuracy of 30% within one working day including all chemical extraction steps. The chemical procedure of separation of strontium from air filters is carried out without the usual addition of strontium carrier to keep the content of stable strontium low. Such a chemical procedure has been worked out. Water and soil samples may have higher contents of stable strontium and thus require still higher values for the selectivity of the method. The performance of the technique might be increased by an optimization of the ion source efficiency and higher optical excitation probability affecting both the overall efficiency and the selectivity. These improvements should enable us to lower the detection limit for 9~ and extend the measurements to 895r (see also Bolshov et al., 1981; Letkhov, 1987). 5.4.3 M e t h o d s b a s e d on m a s s s p e c t r o m e t r y
The electric and magnetic fields, used for the analysis of ions, provide only information about the two quantities E/q and M/q where E, M and q are the energy, mass and charge of the ion respectively. There are four ways in which the quantities E/q and M/q may be determined: 1. magnetic selection (Bp)2 = 2(M/q)(E/q) 2. electrostatic selection Ep = 2(E/q) 3. Cyclotron selection 1/f= (rc/B)(M/q) 4. velocity selection v2= 2(E/q)(M/q) where B is the magnetic field, p the radius of the ion path, E the electric field, f the cyclotron frequency, and v the ion velocity. Low resolution measurements that separate neighbouring isotopes as their final output have the quantities M/q and E/q. Since M can be regarded as an integer, ambiguities can arise if M and q have common factors. It is for this reason that some flexibility in the choice of q is desirable. If it is possible to measure the energy of the ion also, then it is possible to determine q and so determine the mass from the ratio M/q. The use of energy, mass and charge signatures, at energies such that charge state 3+ or higher is dominant, is the basis for the accelerator mass spectrometry of almost all stable isotopes. The methods listed below are based on mass spectrometry, differing mainly in the design of the ion source used: Spark-Source Mass Spectrometry SSMS Glow-Discharge Mass Spectrometry GDMS Inductively-Coupled-Plasma Mass Spectrometry ICPMS Electro-Thermal Vaporization-ICPMS ETV-ICPMS Thermal-Ionization Mass Spectrometry TIMS
Measurements of RadioactiviO'
215
AMS Accelerator Mass Spectrometry SIMS Secondary-Ion Mass Spectrometry SNMS Secondary Neutral Mass Spectrometry LMS Laser Mass Spectrometry RIMS Resonance-Ionization Mass Spectrometry SIRIS Sputter-Initiated Resonance-Ionization Spectroscopy LARIS Laser-Ablation Resonance-Ionization Spectroscopy Let us briefly discuss again the limitations of radioactive decay measurement. The observation of the radioactive decay of a single atom is possible, consequently, with efficient apparatus for the detection of the decay particles and a radioactive species with a half-life of seconds and minutes, it is possible to detect all or nearly all of a small number of radioactive atoms in the presence of a large number of nonradioactive atoms with radiation detection techniques. However, as the half-life increases, the time taken to carry out an experiment with a small number of radioactive atoms naturally increase, for half-lives, of say, 106 years efficient detection of the radioactive decay products becomes impossible unless the measurement can be continued for 106 years. Therefore, studies of long-lived radioactive isotopes invariably use very large numbers of atoms and the apparatus detects the decay of only a small fraction of the total during the experiment. In this situation the mass spectrometric detection sensitivity surpasses by far the sensitivity of radioactive counting methods. An important example is the study of the ~4C(half-life = 5730 years), generated in the atmosphere by cosmic rays, in connection with radiocarbon dating. The observed beta-ray counting rate from one gram of contemporary carbon of biological origin is about 15 per minute per gram. However, this low counting rate is supported by the presence of 6.5x 10~~ of 14C in the one-gram sample. If 14C atoms could be counted efficiently by accelerator mass spectrometry, it would be possible to determine the ~4C content of very small quantities of carbon. This has now been accomplished for milligram carbon samples even though the ratio ~4C/~2Cis near or below 10-~2. A comparison of measurements of long-lived radioisotopes at natural levels with beta-ray counting and AMS (after Elmore and Phillips, 1987) is shown in Table 5.14. Atomic mass spectrometry is inherently sensitive and by its nature provides isotopic information. The goal of methods of elemental analysis based on mass spectrometry is to produce a spectrum of singly charged atomic species. Again this can be a problem for elements, such as uranium, which readily form oxides. If molecular or multiply charged species "contaminate" the atomic mass spectrum, they can give rise to background signals at the mass of interest, or if these molecular ions contain the element of interest, then the signal due to that element is distributed between the atomic singly-charged ion, the multiply-charged ions and any molecular species formed. Atomic mass spectra are simple to interpret, however great care must be taken to avoid molecular interferences especially at very low concentrations. SIMS analysis of electrodeposited 232Th alpha-particle sources gives rise to higher signals for the ThO § and ThO2 + than for Th § This leads to difficulties in quantitation as the oxide to atomic ion ratios will be sensitive to local oxygen concentrations. Isobaric
216
Chapter 5
Table 5.14 A comparison of measurements of long-lived radioisotopes at natural levels with beta-ray counting and AMS (after Elmore and Phillips, 1987) Radioisotope l0B 14C
26A1
36C1
1291
Half-life (years) Stable isotopes Stable isobar
1.6x106 9Be I~
5730 13C 14N
7.05x105 12C 26Mg
4.05x105 27A1 36Ar
1.57x107 35C1
Chemical form Sample size (mg) Atom per sample AMS run time (minutes) Decay counting interval (years)
BeO 0.2 2x105 10 110
C 0.25 2x105 7 3
A120 3 3 4x105 40 250
AgCI 2 5x105 30 86
AgI 2 2x106 20 1130
36S
atomic interferences also present a problem: 99Ruand 99Tc,for example, have the same nominal mass and cannot be discriminated between on the basis of mass except by high-resolution mass spectrometers. Even with high mass resolution, if the interfering isobar is present in excess then discrimination at high mass resolution will be difficult and in any spectrometric method there is a trade-off between resolution and sensitivity. If there is a vast excess of an isotope of adjacent mass, even this may interfere with the signal of interest. The ability for a mass spectrometer to discriminate against such an interference is termed the "abundance sensitivity". Methods such as ICPMS and TIMS must discriminate against isobaric interferences by chemical separation methods prior to instrumental analysis. AMS, which is most commonly used for radiocarbon dating, discriminates against interferences in a number of ways. The InNinterference in ~4C measurement is removed by generating a beam of anions and relying on the instability of the nitrogen anion. Molecular interferences are removed by high-energy (several MeV) collisions in a gas cell or thin foil. Further discrimination can be achieved by charge stripping in the same collisional processes to produce highly charged ions or even ions in their maximum charge state. The combination of discrimination processes used depends upon the isotope of interest and the potential interferences. RIMS approaches the same problem by selectively ionizing only the element of interest prior to mass spectrometric separation. A selectivity of approximately 1 in 105 can be achieved per resonant excitation step in the ionization process and two or three such steps are frequently used. In combination with the selectivity of the mass spectrometer the method should potentially offer elemental selectivities in excess of 1015. 5.4.3.1 ICP-MS ICP-MS uses an inductively coupled plasma as an ion source for a mass spectrometer. The basic units of an ICP-MS system, in the order used, are the sample introduction
Measurements of Radioactivi~,
217
computer [ ~
ion lenses MCS -
~.quadrupole ir-x
[
r ...............
9
to vacuum
data acquisition and handling
I/
k~U
ion filtration
]
sample introduction
Fig. 5.48. Component of a typical ICP-MS system.
device, the plasma, the plasma/mass spectrometer interface, the ion focusing/ion filtering system, the detector, and the data acquisition/data handling system (Fig. 5.48). Beauchemin (1992) gives a helpful comparison between ICP-MS and ICP-optical emission spectrometry. The sample introduction device introduces liquid samples as either a dry vapour or a fine mist into the plasma, with several options available. These include: pneumatic nebulization (the most common), ultrasonic nebulization, electrothermal vaporization (ETV, which uses a graphite furnace), flow injection (Denoyer and Stroh, 1992) and direct injection (Wiederin et al., 1991). The transport efficiencies of sample into the plasma for pneumatic nebulization, ultrasonic nebulization, and ETV are about 1%, and 100% respectively. Laser ablation is a common method for introducing solid samples into the plasma. This and other sampling methods for solids are reviewed by Baumann (1992). An inductively coupled argon plasma is used most frequently in ICP-MS, with argon as the cooling, carrier, and auxiliary gas. The high-temperature plasma (5000-8000~ is sustained by radio frequency fields at the tip of a quartz torch. The plasma desolvates (if necessary), atomizes, and ionizes the sample. Horlick (1992) notes the use of helium or nitrogen-based microwave-induced plasmas to eliminate interference from argon-based background species. Smith et al. (1991), and Lam and McLaren (1990) have used mixed carrier gases to reduce argon-based background ions. The plasma/mass spectrometer interface allows the import of a stream of ions from the plasma, at atmospheric pressure, into the mass spectrometer, which is under vacuum. The interface has a sample cone and a skimmer cone, usually of nickel and typically with 1.0 mm and 0.8 mm apertures, respectively. The turbomolecular pump is presently the most widely used type in the vacuum system. The extraction lens, held at a negative voltage, attracts the positive ions as they emerge from the skimmer cone. The negative ions are repelled and the neutral species diffuse away. The accelerated positive ions are focused by the ion-lens stack and then enter the quadrupole region. Positioned between the extraction lens and the first ion
218
Chapter 5
lens, the photon stop prevents stray photons from reaching the detector. Voltages applied to the quadrupole, four metal rods mounted in a square array, produce an electric field that affects the trajectories. For any specific applied voltage, only ions of a very narrow range of mass/charge (M/Z) ratios have stable trajectories and reach the detector. A single M/Z ratio may be continuously focused onto the detector, or a selection of M/Z ratios can be sequentially focused onto the detector. A channel electron multiplier, the commonly used detector, responds to each incoming positive ion and produces a measurable pulse. The number of pulses measured is proportional to the number of ions of the selected M/Z ratio reaching the detector. In the pulse counting mode, maximum gain is obtained by applying a high voltage to the multiplier so that individual ion arrivals at the detector are recorded and ultimate detection limits are obtained. The analog mode uses lower voltages, which produce lower gains. The use of lower gains extends the useful analytical range but results in higher detection limits. A Faraday cup replaces the analog mode in some instruments. The data acquisition/data handling system consists of a multichannel scalar (MCS) and a computer system. Signal pulses from the detector accumulate into memory channels of the MCS according to their M/Z ratios. Signal pulses from replicate scans are sorted into the appropriate channels and accumulated until the analysis of that sample is complete. A computer program retrieves the totals from the MCS memory and stores the data for later manipulation or display. As mass spectrometry has continued to gain sensitivity and reliability, inductively coupled plasma]mass spectrometry (ICP-MS) has become increasingly useful in the measurement of radionuclides. The optimization of ICP-MS is improving our ability to use the atomic detection of radionuclides in that it allows the near-complete isotopic analysis of any form of sample. Aqueous samples are generally introduced into the plasma source, and solids or individual particles, and organic solutions, may be atomized and continuously introduced into the plasma source. ICP-MS sensitivity, which is currently ~8x 109 atoms, can be improved by: 9 the use of more efficient sample introduction techniques, 9 understanding of the basic principles of ion and gas dynamics in the ICP-MS interface, and 9 the use of high-resolution mass spectrometers with high ion transmission. The ultimate sensitivity could approach -107 atoms, which would result in a superior detection capability for all radionuclides with half lives greater than 1 year. For radionuclides with half-lives of thousands of years and longer, ICP-MS has two principal advantages over radiation counting, which are speed of measurement and sensitivity. Most radiation counting times range from 50 to 2500 minutes for most samples and most backgrounds. In contrast, an ICP-MS analysis requires only a few minutes per sample or blank, whether it is introduced via nebulizer, ETV unit, or other device. The analysis time is independent of the half-life or decay scheme of the radionuclide. The analysis time is also not greatly lengthened by a lower required MDA. Indeed, this advantage of ICP-MS over radiation counting becomes greater with
Measurements of Radioactivit3,
219
increasing half-life and decreasing MDA. Analysis by ICP-MS may be the preferred method even when its sensitivity is not as great as the one obtained by radiation counting, because of its speed. Using ICP-MS quantitatively becomes feasible for radionuclides with half-lives greater than about 1• years. However, the sensitivity that is routinely achievable with ICP-MS is not as high as that achieved with radiation counting for radio-nuclides with half-lives less than about 1• 104 years, unless the decay scheme is unfavourable for radiation counting. For present-day instruments, at least 107-108 atoms are necessary to qualify a nuclide by an ICP-MS with an ETV unit (Smith et al., 1992). To use an ultrasonic nebulizer requires at least 108-109 atoms; while a pneumatic nebulizer requires at least 109-10 j~ atoms. The corresponding masses vary according to the atomic weight of the nuclide. (For 239ptl, 2• 107 atoms equals 8 fg). Under favourable operating conditions, an instrument with an ETV unit should just meet an 8 fg detection limit for 239pu (Ill 2 = 2.41• y). This equals an MDA of 0.001 dpm (1.7• -5 Bq), which is five times lower than the contractual MDA given earlier. This corresponding MDA for 2• 107 atoms of 24~ (tl/2 -" 6.57• 103 y) is 0.004 dpm (6.7• -5 Bq). The sensitivity of ICP-MS for heavier elements is better than for lighter ones, due to lower background in the higher mass region and more stable trajectories for more massive ions (Jarvis et al., 1992). This is an advantage when one is interested in analysing for long-lived radionuclides of the rare earth and heavier elements. Essentially all of the inert sample matrix needs to be removed when performing radiation counting of all alpha particle emitters and low-energy beta-particle emitters because of sample self-absorption. Complete matrix removal may not be necessary for analyses by ICP-MS, depending on the elemental composition of the sample, the analytes, the sampling device, and the required sensitivity. Partial or less-complex matrix decompositions and separations of analytes may suffice. For example, Hursthouse et al. (1992) compare the extent of chemical purification necessary to obtain good results for 237Np via ICP-MS, alpha-particle spectrometry, and neutron activation analysis. Depending on the dissolved solids content, natural waters may need filtration only and/or treatment with acid. A chemical separation of a group of elements may be satisfactory. Preparation of purified samples for alpha-particle spectrometry is usually by either electrodeposition or micro-coprecipitation. Either technique takes at least an hour. Many beta-particle emitters are precipitated with several milligrams of carrier and weighted for determining the chemical yield prior to counting, which also takes time. In contrast, a few millilitres of solution are satisfactory for ICP-MS. If the analyte concentration should exceed the linear part of the calibration curve, a simple dilution overcomes the problem. Mixtures of beta-particle-emitting nuclides of more than one element usually have overlapping spectra. This is also often true for alpha-particle-emitting nuclides. Some of these mixtures can be analyzed by ICP-MS without internal interferences. For example, the alpha-particle spectra of 237Np and 242pu partially overlap even under the best of conditions, whereas ICP-MS is appropriate for analysing the long-lived
220
Chapter 5
radionuclides in a Np-Pu mixture. Only X- and gamma-ray counting are comparable to ICP-MS in the number of radionuclides that can be measured simultaneously. As with all analytical techniques, ICP-MS has its problem areas (Olesik, 1991). Although the mass spectrum of a sample is usually much simpler than an atomic emission spectrum for the same sample, spectral interferences from isobaric interferences and peak overlap can still be a problem. Isobaric interferences result from two situations. The first occurs when two elements in a sample have nearly identical M/Z values. An example of this is the presence of ]~3In+ interfering with the analysis of JJ3Cd+. The second situation occurs due to the formation of "background species". These are ions, usually polyatomic, formed from the plasma gas alone or in combination with elements from the solvent used in the sample preparation. Examples, together with the ions for which they cause the greatest interference, are Ar2+ (8~ ArO + (56Fe+), Ar § (4~ + and 4~ and O, § (3'-S+). Tables of common background species for an argon plasma are readily found in the literature. Most of these interfering species have m/Z < 81. Fortunately, most long-lived radionuclides have masses greater than 81 amu. Peak overlap occurs when major constituents in the sample have such massive peaks at particular M/Z value channels. Examples of this type of interference would be a massive peak for uranium at mass 238 that tails into the 237 and 239 mass channels, thus complicating the analysis of 237Np+ or 239pu+. Other problems may be caused by the matrix of the sample itself. If chloride is present, a series of polyatomic chloride-containing species may cause major interferences. As an example, 4~ is an intense peak that interferes with 75As+. Arsenic is monoisotopic; therefore appreciable levels of chloride in the sample will seriously compromise the precise determination of arsenic. Components of the sample matrix may also contribute to oxide formation. Oxides of the form MO § give rise to peaks at the (M/Z)+ 16 position. One or more of these may interfere with nuclides of interest. An 64 + example of this is 48Ti]60* interfering with the analysis of Z n . The four other naturally occurring titanium isotopes would, of course, also give interference at their respective (M/Z)+16 values to any analytes with these masses. Formation of the oxide of the analyte also reduces the signal measured at M/Z. The sample matrix may also induce changes in the analyte signal intensity. High concentrations of contaminant elements generally cause suppression of an analyte signal, although under certain conditions signal enhancement has been observed. In general, the lower mass elements are more subject to suppression than are higher mass elements, and higher mass elements are more likely to cause signal suppression of lower mass elements than the reverse (Jarvis et al., 1992). An ICP-MS instrument will not tolerate dissolved solids at concentrations that can be run with an ICP-atomic emission spectrometer. In addition to increasing the probability of interelement (isobaric) interferences and signal suppression, high levels of dissolved solids condense on the sample-cone orifice. This deposition degrades the sensitivity and stability of the analytical signal. Typically, a maximum of 0.1% dissolved solids is recommended for continuous nebulization with a pneumatic
Measurements of RadioactiviO'
221
nebulizer. Dissolved solids should be kept below about 0.01% with an ultrasonic nebulizer, due to its desolvation effect. Liquid samples containing up to about 1% dissolved solids can be run with ETV and flow injection. The sensitivity achievable for an element is inversely related to its ionization potential (I.P.). Thus, for example, the ICP-MS sensitivity for iodine (I.P. = 10.34 eV) is not as good as it is for nearby cesium (I.P. = 3.89 eV). Finally, the sample aliquant is consumed during an ICP-MS measurement, whereas with radiation counting the sample aliquant usually can be retained and can be remeasured. Some of the reported ISP-MS applications to radionuclide determination are presented next. Kim et al. (1989a) measured the 24~ ratio in two soils and an estuary silt after performing radiochemistry. They also measured this ratio in these soils by the fission track-etch technique and found that ICP-MS gave better precision. They measured the 239pu concentration on separate aliquants by alpha spectrometry, with 3.76x10Lyear 242pu tracer as the chemical yield monitor. However, at the 239pu concentrations in their samples, ICP-MS could have measured these directly. Determination of low levels of 99Tc in environmental samples by ICP-MS was reported by Nicholson et al. (1991). In earlier work Nicholson et al. (1989) used ICP-MS with the beta-particle-emitting nuclide 99Tc, in salt marsh soil, seaweed, and sea water. They employed 95mTcas a yield tracer, and radiochemically isolated Tc from the matrix. They confirmed the chemical removal of any interfering isobaric 99Ru by monitoring for other stable Ru isotopes. The chemical yield was measured by gammaray spectrometry. Beals (1992) used 2.6-million year 97Tc as the yield monitor in ICP-MS measurements of 99Tc in river water, thereby eliminating the need for a separate yielding measurement. In purified water, they have reliably detected 0.05 ng/ml of very low specific-activity l l3Cd by ICP-MS with pneumatic nebulization. For environmental waters, better sensitivities could be achieved by sample concentration and a clean-up that includes the removal of interfering ~3In. It is entirely impractical to detect 0.05 ng of ~3Cd by beta-particle counting. James et al. (1989) demonstrated, with diluted aqueous standards, an MDA of approximately 8 fg of 239pu (1.7x10 -5 Bq) with ETV-ICP-MS in the peak dwelling mode. Comparison measurements by ICP-MS and alpha spectrometry on radiochemically processed urine with moderately higher activity gave good agreement, considering the very low amount present. Plutonium-244 (t~/2 = 8.3x 107 years) is also amenable to analysis by ICP-MS, whereas 238puis not. ICP-MS was found to be compatible with LC for the trace metal speciation. The role of ICP-MS in trace element speciation studies at the FSL was described (Crews et al., 1987). The characteristics of LC-ICP-MS for the study of metalloprotein species were assessed and the chromatographic efficiency of ICP-MS was found to be similar to that obtained with a UV detector (Dean et al., 1987). Information about the chemical nature of trace elements from food can be obtained by first treating the foods in vitro with enzymes to broadly simulate the action of enzymes in the gastrointestinal tract (Crews et al., 1988). The soluble components can be separated by size exclusion chromatography (SEC) and an estimate of their molecular size obtained. By coupling SEC
222
Chapter 5
directly to ICP-MS, the trace element content of the chromatographic fractions can also be measured. This approach has been used at the FSL to investigate the speciation of cadmium in raw and cooked pig kidney (Crews et al., 1989). The sensitivity of ICP-MS enabled the researchers to study retail samples in which the levels of multielement data obtained indicated that, while the feeds were contaminated with a number of elements, only lead presented a serious problem in parts of the rest of the food chain. For example, while samples from affected cattle were not allowed to enter the food chain, experiments with meat on contaminated bones showed that lead did not migrate significantly from the bone under a variety of cooking conditions (Baxter et al., 1992). Kim et al. (1991) have reported the measurement of some long-lived radionuclides, such as 99Tc, 226Ra, 232Th, 237Np, 238U,239pu and 24~ using high-resolution inductively coupled plasma mass spectrometry (HR-ICP-MS). By using HR-ICP-MS with an ultrasonic nebulizer, the detection limits of these nuclides were 0.002-0.02 pg ml -~ and the sensitivities were 10 times better than those obtained using HR-ICP-MS without the ultrasonic nebulizer. More accurate isotopic data were also obtained using HR-ICP-MS than with quadrupole ICP-MS at lower concentrations of the analyte because of improvement in counting statistics that can be obtained with HR-ICP-MS due to the greater efficiency of ion transmission. Morita et al. (1993) have applied ICP-MS to the determination of technetium-99 in environmental samples. The determination of eliminating the interfering element (Ru) before the ICP-MS measurements are made. Technetium-95m is used as the chemical recovery tracer. Compared with conventional methods, the method sensitivity is 10 to 100 times higher and the counting time is 300 to 10,000 times shorter. Tye and Mennie (1994) have reported development of the performance of a new interface for the Plasma Quad ICP-MS, which enhances the signal-to-noise performance of the standard instrument by a factor of ten. In order to maintain the flexibility of the instrument, the new interface is designed such that the enhanced performance can be easily switched on and off, offering the benefit of routine performance plus the high sensitivity mode when required. These improvements in signal-to-noise make it possible for the routine monitoring of many actinide elements directly by ICP-MS, potentially shortening the analytical cycle from days to hours. A further improvement on these impressive limits of detection is possible if a high efficiency nebuliser is used to introduce samples into the new instrument, giving the capability of single figure ppq detection limits in an analysis which takes minutes, not hours. 5.4.3.2 AMS
For isotopes with long lifetimes (> 1 year), it may often be more advantageous to use atom-counting techniques rather than traditional decay-counting methods. This is especially true for measurements where efficiency is a criterion, as for small samples, or if high precision is required. While atom counting has a counting rate that is essentially independent of decay lifetime and sample size, the decay-counting rates are comparable only if the isotopic half-life is less than one year for a sample size of the
Measurements of RadioactiviO,
223
Table 5.15 Long-lived cosmogenic isotopes detected with accelerator mass spectrometry Isotope
Half-life
Interfering stable
(years)
Isotopes 6 3
AMS detection
Range of terrestrial
Isobars
limit (*)
concentration (*)
9B e 12,13C
I0B 14N
7• -15 0.3• -15
10-8_10 -14 10-12_10 -16
l~ 14C
1.5• 5.7•
26A1 36C1 41Ca
7.2• 5 3.1• 1.3x105
27A1 35,37C1 4~
26Mg 36S 41K
10• -15 36Ar 500• -15
_10 -14 0.2• -15 10-15_10 -16
1291
15.9•
1271
129Xe
100•
...10-16
-15
*Compared to the stable isotope of the same element.
order of 1 mg. Of course, if sufficient material is available, the decay-counting rate can always be improved by using more material (Litherland, 1987; Kilins et al., 1992). Accelerator mass spectrometry (AMS) extends the capabilities of atom-counting using conventional mass spectrometry, by removing whole-mass molecular interferences without the need for a mass resolution very much better than the mass difference between the atom and its molecular isobar. This technique has been used with great success for the routine measurement of 14C, 1~ 26A1,36C1and, recently, 129I (see Table 5.15). Analysis of 14C by AMS can, for example, generate dates with a precision that is at least equal to the best conventional beta-particle-counting facility. In many cases, where small sample analysis is required, the AMS method has proved superior (Benkens, 1990). A complete description of AMS can be found in review articles (Litherland et al., 1987; Elmore and Philips, 1978) or recent conference publications. The application of AMS to 129I measurement has been discussed in detail in Kilins et al. (1992). Accelerator mass spectrometry (AMS) is an analytical technique that uses an ion accelerator and its beam transport system as an ultrasensitive mass spectrometer. Accelerator mass spectrometry (AMS) was introduced in 1977 by Muller, who suggested that a cyclotron could be used for detecting 14C, 1~ and other long-lived radioisotopes, and independently by the Rochester group, who demonstrated that 14C could be separated from the isobar 14N by relying on the instability of negative ion 14N-. Presently AMS measurements are being made at about 30 accelerator laboratories around the world, and half of these are dedicated to AMS measurements of long-lived radioisotopes. Six long-lived radionuclides beyond uranium exist which have half-lives greater than 100 ka (236Np, 237Np, 242pu, 244pu and 248Cm). The first two are natural by-products of the nuclear industry. Nuclear-weapons tests will generate the plutonium and curium isotopes although attempts have been made to detect pre-solar system 244ptl in ores (Hoffman et al., 1971) or 244pu from more recent supernova debris. The detection of these isotopes is still in the development stage. Unlike the natural elements, isobaric
224
Chapter 5 i
.... ~
analyzer 1
velocity selector
stripper AT +HV terminal
negative ion source
/
"x..
iiiii
.......
~. magnetic
5 / / . ....
analyzer
preaccelaration
INJECTOR
,
TANDEM ACCELERATOR '
magnetic analy
[
gas-filled k4r ionization \ ~ chamber " ~ time-o;flight
I: magnet DETECTOR SYSTEM
ES analyzer POSITIVE ION ANALYSIS
Fig. 5.49. Components of a typical AMS system.
interferences are not a major problem as all isotopes will be equally rare or non-existent because of their very short decay half-lives compared to the lifetime of the solar system. The components of accelerator mass spectrometry, AMS, system are shown in Fig. 5.49 and they include: ion source, injector, tandem accelerator, positive ion analysis and detection system. A caesium sputter ion source is used for most AMS work. This is essentially a secondary ion mass spectrometry (SIMS) instrument that has been refined to produce high current of negative ions. Generally, solid samples are used; gas samples can give intense beams, but the problem of contamination from the previous sample ("memory") is difficult to overcome. For radioisotope studies, sample sizes are 1 to 10 mg of processed material and beam currents of 1 to 50 ~tm are typical, depending on the element and ion source model. Some sort of multiple sample changing system is used at most AMS installations. For example, the main features of the 846B model highintensity sputter source (High Voltage Engineering Europe) include: a hemispherical ionizer giving a focused Cs beam spot of less than 0.5 mm, an x-y scanning stage to limit cratering effects and a 60-sample carousel with automated remote loading for throughput work. Currents of up to ~tA ~2C- have been quoted for this source from graphite targets. Mass analysis of the negative-ion beam with a resolution sufficient to separate isotopes of heavy elements is needed prior to acceleration. For example, an electrostatic analyzer is used at the University of Toronto to sharpen the energy distribution of ions produced from a caesium sputter ion source. A pre-acceleration of the negative ion beam
Measurements of Radioactivity
225
to 100 to 400 keV is used with large tandem accelerators to ensure that the injected ion beam is focused at the central terminal where the stripper canal is located. The name "tandem" refers to a dual acceleration design. The negative ions are accelerated to the terminal of the accelerator, which is held at a constant positive voltage, typically in the range 2 to 10 MV. The electron stripper at the terminal removes several electrons while energetic negative ions pass through; positive ions are then accelerated from the terminal to the end of the accelerator (ground potential). Tandetrons operate reliably below 3 MV using a solid-state power supply, and tandem Van de Graaff accelerator use a rotating belt or chain to charge the terminal up to 25 MV in some models. Tandem accelerators have the following characteristics: (i) the ion source and detector are located at ground potential for tandetrons; (ii) they do not require pulsed beam; (iii) the electron-stripping step need to eliminate molecules is an integral step in the operation of the accelerator; and (iv) transmission through the accelerator and subsequent analyzers can be made independent of small changes in the terminal voltage. Analyzers positioned after the accelerator remove scattered particles accepted by the injector analyzer, molecular fragments, and unwanted charge states. Magnetic analyzers alone are not sufficient. An electrostatic analyzer or velocity selector is necessary to remove particles that have different mass but would otherwise have the correct mass-energy product to pass through the magnetic analyzers. At 1 MeV/amu energies, the dE/dx and total energy measurements are made with either gas ionization detectors or silicon surface-barrier detectors or a combination of these. The time-of-flight detector serves as an additional positive-ion mass analysis stage. It is most useful for the heaviest (slowest) ion such as 129I and consists of two time-pickoff detectors with time resolution of a few hundred picoseconds. Isotope ratios are obtained by alternately selecting each stable isotope and measuring its beam current in a removable or offset Faraday cup and then by measuring the radioisotope (rare nuclide) counting rate in the detector. Standards (samples with a known isotope ratio) are periodically measured for normalization, and blanks (samples containing no detectable nuclides to be measured) are used to measure background. Ratios are corrected for time-varying linear mass fractionation when more than one stable isotope is measured and for nonlinear fractionation, which arises from the stripping process and from stray magnetic fields in the accelerator, by comparison to the standard. The precision of ratios ranges from 1% to 10%, in the AMS measurement, depending on the value of ratios and counting time (if background is low enough). The long-lived radioisotopes ~~ ~4C,26A1, 36C1, 41Ca and 129Ican now be measured in small (mg) natural samples having isotopic abundances in the range 10-~2 to 10-~5 and as few as 10 5 atoms. At elevated energies (> 1 MeV/amu), ions can pass through thin or equivalent gas with virtually no attenuation of the particle beam and little energy loss. As a result of electron capture and loss interaction, an ion passing through matter is characterized by the fraction of the total ions (Fq)in a given charge state (q) where ZFq = 1. The resulting charge state distribution is determined by the electron capture and loss cross section of
226
Chapter 5
an ion in a gas or solid. An equilibrium distribution will be established, the character of which depends only on the ion velocity and the target material. This equilibrium distribution is independent of the initial ionic charge or the target thickness, and approximation is valid as long as the energy loss remains insignificant. The passage of an ion through matter with the subsequent removal of electrons from molecules will decrease the bond strength among the constituents. Generally, after a reduction of two electrons in ionically bound molecules, no bond is possible and the molecule is broken up by Coulomb force. A sufficient number may remain to leave the charged molecule in a stable or metastable configuration. To avoid the possibility of long-lived (> 1 ms) metastable molecules, at least three electrons must be removed. At present no 3+ molecules are known to exist. For light ions (Z < 20) an energy of at least 3 MeV is needed to maximize the production of charge state +3 ions in a gas cell. As the mean ionic charge rises approximately linearly with energy, the higher charge states will dominate at the 8 MV accelerating potentials. Usually one of these charge states was chosen to provide the highest conversion efficiency and no molecular interference. Separation of isobars can be accomplished by using different approaches:
(i) By chemical separation: No two isobars will have the same atomic number and rarely will they belong to the same chemical group of elements. Consequently, an initial reduction of the isobaric interference is always possible by some form of chemical processing. The degree to which this is effective depends on the specific chemical differences among the isobars and the level of isobaric contamination. Because the mass ambiguity is always formed from stable elements which have had the benefit of the last 4-5• 109 years to achieve some level of contamination in all materials, chemistry can not usually eliminate this type of ambiguity significantly below one ppb. One part in 106 (1 ppm) is the typical level of purity for most reagents. At this level of refinement, the interference is still at least many orders of magnitude greater in concentration than the rare long-lived radioisotopes. For some very rare isobars, specifically 36S (0.017%) in the case of 36C1, careful chemical processing is the primary means of isobaric reduction.
(ii) Using negative ions: Some isobars can be eliminated by exploiting the instability of negative ions. For example, noble gas negative ions are known to be metastable or unstable, thereby removing 36At and 129Xeas a source of interference in the measurement of 36C1and J29I. Others such as JaN- and 26Mg- do not form readily or have metastable states which decay in a time scale << 1 ms. Since the life-times of these metastable states are small in comparison to the transit time (> 1 ms) through the ion analysis system, attenuation factors in excess of 106 were possible with negative ions for the detection of ~4C and 26A1. In contrast to the high probability for scattering and multiple charge changes for positive ions used at low energies (keV), the scattered negative ions from the more
Measurements of Radioactivit3.,
227
intense isobaric beams are greatly reduced in intensity after an interaction with the residual gas in the ion source. This can be attributed to the low binding energy of the negative ions.
(iii) By fully stripping the electrons: At sufficiently high energies and for cases where the radioisotope has a higher Z than the stable isobar, separation by fully stripping with subsequent magnetic analysis can be accomplished. This method has been investigated for the systems 3He2+-3H~+, r . 25+ - 59r,~ 26Al13+-26Mg12+, 36Cl17+-36516+,41Ca2~ 41K19+,53Mn25+~ 53Cr24+and 59-~1,~1 ~o 24+ . Isobar separation by fully stripping actually allows the simultaneous acceleration of the radioactive and stable isobar, with the latter of sufficient intensity to provide beam feedback signals.
(iv) By energy loss measurement: When ions with energies about 1 MeV/amu are passing through matter, the energy loss per unit path length (dE/dx) of an ion (Z) traversing a solid or gas with velocity (V) is governed by the Bethe-Block equation: dE/dx = k (Z/V) 2 where k is a constant. The effectiveness of energy loss measurements decreases as the atomic number increases. The percentage difference in energy loss between ~4C and ~4N is 30% at 40 MeV in isobutane gas. This rapidly decreases to 6.8% between 36C1 and 36S. Beyond calcium, very high energies (>100 MeV) are required. (v) By gas-filled magnet: The gas-filled magnet is a powerful isobar separation instrument developed only recently for AMS. An ion passing through a gas changes its charge frequently by electron capture and loss. If this charge-changing occurs frequently enough in a magnetic field region, the trajectory is determined by the average charge state of the ion, which depends on Z. Development of the gas-filled magnet should make possible AMS of 36C1 at lower energies and aid in AMS detection of other heavier isotopes. (vi) By lasers: Lasers are very powerful instruments to separate elements. Since the separation of isobars from different elements is the most difficult task in AMS, the use of lasers in connection with AMS could provide a very effective clean-up of background. The basic idea in a recent proof-of-principle experiment at the Rehovot AMS facility was to clean a negative ion beam from unwanted isobaric background ions by selective electrons detachment. 32S- ions which have an electron affinity of 2.08 eV were effectively neutralized by interaction with 2.33 eV photons from a pulsed Nd:YAg laser. The same photons did not affect 37C1 ions whose electron affinity is 3.62 eV. This clearly demonstrated that a laser depletion of 36S background in 36C1 measurements is feasible, opening up the possibility for sensitive 36C1measurement at small AMS facilities where the ion energy is too low to perform isobar separation. However, for actual applications in AMS measurements, a substantial improvement in overall efficiency of the laser depletion process is necessary.
228
Chapter5
The developments in instrumentation include production of dedicated machines. For example, with the introduction of the "Attomole 2000", HVEE (High Voltage Engineering Europe B.V., P.O. Box 99, 3800 AB Amersfoort, The Netherlands, Phone +31 33 4619741; Fax +31 33 4615291 ) has made available a compact ~4C Isotope Ratio Mass Spectrometer (14C IRMS) for biomedical applications. The system provides 14C//12C ratios down to 10 -13 from sub-milligram samples in typically a few minutes. Both solid samples (carbon) as well as CO 2 can be analyzed. The Attomole 2000 combines a compact instruments package with the extreme sensitivity of large tandem accelerators that are normally found in big research centres. The main application of the instrument will be for tracer kinetic and pharmacological measurements in biomedical studies. Essentially, ~4C IRMS is an alternative technique to 13C Isotope Ratio Mass Spectrometry (~3C IRMS), which is widely accepted within the biomedical community. 14C IRMS is at least 100-1000 times more sensitive than 13C IRMS. This allows, for example, that the kinetics of lnc-labelled enzymes, aminoacids or carcinogens can be studied in the human body at levels that are comparable to actual environmental exposure. However, a widespread acceptance of ~4C IRMS in the biomedical community is presently handicapped by the size of the existing ~4C IRMS systems and their need for expensive support personnel. These shortcomings of the present instrumentation are overcome with the introduction of the Attomole 2000. The system is a compact, turnkey instrument that is user-friendly. Its characteristics are reflected in the following specifications: System Footprint: 2.22x 1.25 meters Sample medium: Solid graphite or CO 2 Solid graphite: minimum 100 ~tg; CO2 sample: Required sample size: 0.2 [amole CO,. (minimum 10 x modern) ~4Cfl2C ratios Output: Accuracy: Better than 3% >250 cnts/sec, for 10x modern samples Counting rate: Less than 10 13 for ~4Cfl2C ratios (-0.1 modern) Background: Detection efficiency: 1-2% approx. 4 attomole (10 TM mole) ~4C for 3% Detection limit: accuracy 400 sample/day (based on samples with an Throughput: average of 4x modem ~4C content) Furthermore the system is fully automated, self-tuning and needs little or no maintenance. The operator will consider the instrument as an analytic tool; the fact that an accelerator is involved is incidental. Up to 50 solid graphite samples can be loaded in a carousel prior to analysis. CO 2 samples can be admitted on line to the ion source. The ion source uses a primary caesium beam to sputter the sample under investigation to form a negative carbon ion beam. The ion beam is accelerated through the system to reach the detector with an energy of 2.5 MeV. A detailed description of the system can be found in: Nucl. Instr. and Meth. B 123 (1997) 159.
Measurements of RadioactiviO'
229
5.4.4 Laser-induced photoacoustic spectroscopy (LPAS) Growing interest has been recently directed to the application of photoacoustic sensing techniques to the spectroscopic analysis of various optical absorbers in very dilute concentrations. For this purpose a laser is commonly used as a light source. Since the discovery of the photoacoustic effect by A.G. Bell in 1880, its application has a long history of development. Renewed interest in photoacoustics has emerged, starting with the work of Kreuzer in 1971 who analyzed trace amounts of gas molecules by laser-induced photoacoustic generation. The theory, instrumentation and application of laser-induced photoacoustic generation developed in recent years have been thoroughly reviewed by Patel and Tam (1981) and more recently by Tam (1983, 1986). Other reviews are also available in the literature from different authors: Pao (1977), Somoano (1978), Rosencwaig (1978), Colles et al. (1979), Kirkbright and Castleden (1980), Lyamshev and Sedov (1981), Kinney and Stanley (1982), West et al. (1983) and Zharov (1985). Because of difficulties involved in handling radioactive preparations, the photoacoustic sensing technique had not been applied until some years ago to the spectroscopy of aqueous actinide ions. A relatively simple detection apparatus of photoacoustic spectroscopy for the spectral work of actinide ions using pulsed laser as a light source has been developed. This detection apparatus can be used for radioactive c~-emitting aqueous samples without restriction to corrosive solutions and facilitates the spectroscopic investigation of actinide solutions, particularly transuranic ions, in very dilute concentrations. The creation of the photoacoustic signal in the solution is based on the conversion of absorbed optical radiation into heat by non-radiative relaxation processes. This principle is illustrated in Fig. 5.50. The geometry of laser beam and photoacoustic signal generation is schematically presented in Fig. 5.51. The spectroscopic system has been introduced to different nuclear chemical laboratories and further developed for a variety of purposes. Most of these developments are confined primarily to the spectroscopic investigation (i.e. speciation) of actinides in very dilute solutions or natural aquatic systems in which the solubility of actinides is, in general, very low ( < 1 0 -6 mol L-~). Optical spectroscopy of high sensitivity is an indispensable tool for the study of the chemical behaviour of actinides in natural aquatic systems, which has a newly developing research field in connection with nuclear waste disposal in the geosphere. For this reason, not only is photoacoustic spectroscopy attracting great attention but also thermal lensing spectroscopy and fluorescence spectroscopy, all using laser light sources, are in growing use for the same purpose. Actinides have particular spectroscopic properties which are characterized primarily by thef--->ftransitions within the partially filled 5fshell and thus by a number of relatively weak, sharp absorption bands. The optical spectra of actinides are characteristic for their oxidation states, and to a lesser degree dependent upon the chemical environment of the ion. Thus spectroscopic investigation provides information on the oxidation state of an actinide element and also serves to characterize the chemical
230
Chapter 5 Modulated light source e.g. Laser pulse
II
Ion specific absorption
II
Generation of heat by nonradiative relaxation
II
I
Modulated
]
volume expansion
II Generation and propagation of acoustic wave
Detection of compression wave by piezoelectric detector
Fig. 5.50.
Generation and detection of photoacoustic signals.
Radial intensity distribution Laser beam
[
Sample cell
k. 1D Halfwidth
Temporal intensity distribution
A R
Halfwidth
Xp
Detector
Fig. 5.51. Geometry of laser
beam and photoacoustic signal generation.
states, such as hydrolysis products, various complexes and colloids. Hence, laserinduced photoacoustic spectroscopy with its high sensitivity can be conveniently used for the speciation of aqueous actinides in very dilute concentrations. For a summary of the present knowledge of laser-induced photoacoustic spectroscopy, as regards theoretical backgrounds, instrumentation and radiochemical applications to particular problems in aquatic actinide chemistry, see Kim et al. (1990). Since there is no other radiochemical application known in the literature, except the measurement of tritium decay by an acoustic sensing technique, the present discussion is limited to application to actinide chemistry, particularly in aquatic systems. The most interesting field of application is and will be the geochemical study of long-lived
Measurements of Radioactivit),
231
radionuclides, namely man-made elements (transuraniums). The main importance for such a study is not only the detection of a migrational quantity of radioactivity but also the characterization of their chemical states and hence their chemical behaviour in given aquifer systems. Knowledge of this kind will facilitate a better prediction of the environmental impact of transuranic elements which are being produced in evergrowing quantities and will be disposed of in the geosphere. Since LPAS application to actinide chemistry is in its infancy, only a limited number of works are available in the published literature. Experiments hitherto performed are confined to either hydrolysis, complexation reactions with carbonate, EDTA and humate ligands and a variety of speciation works for Am(III) and to much lesser extent for U(IV), U(VI); Np(IV), Np(V), Np(VI); Pu(IV), Pu(VI). Of considerable interest is the LPAS application to the direct speciation of actinides in natural aquifer systems, where the solubility of actinides is in general very low and multicomponent constituent elements as well as compounds are in much higher concentrations than actinide solubilities. The study of the chemical behaviour of actinides in such natural systems requires a selective spectroscopic method of high sensitivity. LPAS is an invaluable method for this purpose but its application to the problem is only just beginning.
5.5 QA/QC PROCEDURES Quality assurance to determine radionuclides in food and environmental samples ensures that the quality of data obtained is maintained at an adequate confidence level, and is objectively evaluated. Quality assurance includes quality control, which involves all those actions by which the adequacy of equipment, instruments and procedures are assessed against established requirements. For the purpose of quality assurance, the following items must be ensured: (1) equipment and instruments function correctly, (2) procedures are correctly established and implemented, (3) analysis are correctly performed, (4) errors are limited, (5) records are correctly and promptly maintained, (6) the required accuracy of measurements is maintained and (7) systematic errors do not arise. In general, the design of a quality assurance program should take the following factors into account: a. quality of equipment and instruments, b. training and experience of personnel, c. verification of procedures by the routine analysis of control samples and the use of standard methods for analysis, d. frequency of calibration and maintenance of equipment and instruments (variability in the measuring system is an important aspect of this), e. the need for traceability of the results of determinations to a national standard, f. the degree of documentation needed to demonstrate that the required quality has been achieved and is maintained.
232
Chapter 5
It is important to have each item of the quality assurance program established. Intercomparison is also necessary to generally evaluate the quality assurance of the determinations. By this process, it is possible for data to be compared between laboratories or within a laboratory at different times. The concept of "quality control" should be discussed in general comparison with that of the concept of "quality assurance." The basic concept of quality assurance is that quality should be assured comprehensively and wholly from the beginning to the end of a fixed volume of successive procedures. It assures that whole data acquired by using the fixed volume come to have a signification result to meet intended objectives. On the other hand, quality control is related only to definite and practical control of respective procedures, and is limited to only some portions of those procedures. Its main objective being maintenance of quality of results within a specific limit. The scope of quality assurance is extended from one laboratory, to a group of laboratories in a region, then on to those in a country, and then to international groups of laboratories. The wider the scope of subjects to be assured the more effective the quality assurance. A smaller scope can be utilised as part of a larger one. This means that more effects are found in scopes of quality assurance in ascending order from a single laboratory to a region, a country, a continent, and to the whole world. The causes of errors which are treated as problems in quality assurance for radioactivity analysis and measurements are (1) collection of samples; (2) sampling; (3) transportation of samples; (4) labelling of samples; (5) storage of samples; (6) pretreatment of samples; (7) procedures for measurements; (8) measuring instruments; (9) human errors; (10) erroneous conversion; (11) reporting and notification; (12) environmental changes; and (13) misinterpretation of data. The causes from items (1) collection of samples to (5) storage of samples are related to procedures to handle them. Quality assurance is established by comprehensively quantifying the partial uncertainty of errors which are to be generated from these causes. There are a few items for implementation: a. Organisation must be implemented to aid in establishing quality assurance. Firstly, a laboratory or research institute to play the central role has to be set up or appointed in a region or a country. It should act as a centre, play a leading role, and handle the clerical work in that country. A committee may also be set up in the central laboratory for specialist members to offer guidance and advice. Ideally, a network should be formed to cover all the laboratories concerned, with the central laboratory acting as leader. In some smaller operations, a network will be composed only of laboratories who agree to join. The items to be executed for QA involve: 9 To form a network of quality assurance coveting all laboratories concerned, with a central laboratory acting as leader. 9 For the central laboratory to check, compare, and analyse the work of all the laboratories including itself. 9 For the central laboratory and/or all the laboratories to conduct periodic calibration and stability checks of instruments.
Measurements of Radioactivi~.,
233
9 For the central laboratory and other related laboratories to make comparative measurement and analyses either continually or periodically. 9 To carry out exercises related to the network. b. It is easier to maintain technological levels if instruments are subjected to periodic calibration and stability checks. These can be performed by the respective laboratories. Needless to say, checks should be done whenever operators are changed, instruments are installed, replaced, or moved, or environmental conditions are changed. c. Intercomparison and comparative measurement have to be conducted continually and periodically for assuring quality. Irregular and/or short term checks never represent real assurance and have the least effect as a quality assurance system. d. Quality cannot be assured completely from the very beginning. Quality assurance will take time before it is refined. To improve the level of quality assurance, the following steps must be carried out: 9 Unify the subjects to be sampled and the sampling methods. 9 Standardise measuring procedures. 9 Standardise specifications of measuring instruments. 9 Have specialists to operate the measuring instruments. 9 Standardise forms to make results accessible to all concerned. 9 Understand regional characteristics. e. The procedures to keep up quality assurance levels must be incorporated into the quality assurance system by all laboratories. The list includes: 9 Secure (a) operators with full expertise, (b) appropriate methods, and (c) appropriate, well organised locations and space. 9 Supply standard samples. 9 Examine materials prior to application. 9 Calibrate and adjust instruments. 9 Make use of reference and standard samples which have appropriate records. 9 Check quality assurance procedures. 9 Effect continuous review of related data. 9 Check whether objectives are met or not. 9 Use divided samples. 9 Compare data with those of other laboratories. 9 Correctly handle requests. 9 Review results in an organised manner 9 Correct errors, if any, by means of continuous measurements. Quality control measures are necessary to provide documentation to show that the analytical results are reliable. This is very important since analytical results can form a basis upon which economic, administrative, medical and/or legal decisions are made. It is essential to develop a quality assurance (QA) programme that covers sample collection, sample handling, and methods for on-site and laboratory analysis, data handling and record keeping. The QA programme should address the variety of different scenarios likely to be encountered. Appropriate calibration and analytical
234
Chapter 5
standards and a variety of reference materials will be needed. To keep costs down, one should carefully design a QA programme that recognises that for some signatures high precision data are not required. If, for example, one analyses for a typical short-lived radionuclide which does not exist in nature, background measurements are unnecessary, however low the reported concentration. In other cases where one looks for faint anomalies in certain isotope ratios, the QA programme should demand a knowledge of background values and their variability; this would be much more expensive. The protocols should include "blank" samples as well as "background" samples. In the case that an attempt is made to find an undeclared facility adjacent to a declared one, the analyst should try to take "background" samples from a plant somewhere else, which is similar to that part of the installation which is being examined. When attempting to find an undeclared nuclear facility at a declared site, the optimum background samples would be from similar facilities which are a part of the declared installation. Reliability of results is a function of precision (reproducibility) and accuracy (true value). The precision of results can easily be determined by internal measurement. The determination of accuracy in most cases, however, requires more detailed procedures such as the following: 9 Analysis by as many different methods, analysts and techniques as possible. In cases where agreement is good, the results are assumed to be accurate. 9 Control by as many different methods, analysts and techniques as possible. In cases where agreement is good, the results are assumed to be accurate. 9 Control analysis with reference materials that are as similar as possible to the materials to be analysed. Agreement between certified and observed values is then a direct measure of accuracy for that particular determination. 9 Participation in an interlaboratory comparison. Samples used in such an intercomparison should be, as far as possible, similar in composition and concentration to the samples to be analysed on a routine basis. The agreement between the results received from a particular laboratory and the most probable mean value obtained from statistical evaluations of all the results will be a measure of the accuracy for that particular determination.
5.5.1 lntercomparison For practical reasons, most analytical laboratories are not in a position to check accuracy internally, without an external source of reference materials. To overcome some of the difficulties in checking the accuracy of analytical results, the IAEA provides the Analytical Quality Control Services (AQCS) Programme to assist laboratories in assessing the quality of their work. AQCS co-ordinates intercomparison studies and supplies reference materials. Participation is on a voluntary basis and at minimum cost. Information supplied by laboratories taking part in the intercomparisons is treated as confidential.
Measurements of Radioactivi~.'
235
Table 5.16 Analytical quality control services (AQCS) Year
Intercomparison
Reference available
Materials distributed
1986
24
39
1450
1987
24
38
1680
1988
33
46
2700
1989
27
50
1800
1990
19
58
1850
The IAEA has traditionally played an important role in the development and testing of analytical methodology for determination of radionuclides and through the AQCS programme provides a service by offering laboratories the option of determining their accuracy by distributing reference and intercomparison materials containing radionuclides in different types of materials. The analytes of interest in these samples include naturally occurring radionuclides and radionuclides of fission and activation products. The activities of the IAEA AQCS programme are shown in Table 5.16. Currently the orders for reference and intercomparison materials are running at the level of about 3000 units per year for the whole AQCS programme. The distribution of reference and intercomparison materials is co-ordinated by the Chemistry Unit of the Agency's Laboratories at Seibersdorf, but it also receives input from other Sections of the IAEA, including the Hydrology Section, the Nutrition and Health Related Environmental Studies Section, the Safeguards Analytical Laboratory, Monaco. Intercomparison studies organised over the last twenty years are generally based on recommendations of consultants' group meetings, and in response to the demands of many of the IAEA Member States for assistance in developing methodologies for the measurement of radioactivity. The Chemistry Unit distributes every four years a questionnaire concerning the need for organising intercomparison tests and the preparation of reference materials. Using this data the AQCS programme collects different kinds of environmental and foodstuff bulk samples, some of which were affected by fallout radioactivity following the Chernobyl nuclear reactor accident. The general policy is to organise intercomparisons with those materials which are in most demand and have various levels of activity. Collection of a sufficient quantity of the raw materials (typically of the order of two to four hundred kilograms) is first organised. The samples obtained by a sampling operation are generally dried, ground and homogenised. Aliquots are then taken at this stage and analysed to check the homogeneity of the bulk materials. Other preparation steps include aliquoting into bottles in amounts of about 25-100 g per bottle. To ensure long-term stability of the material, the sealed bottles are sterilised by gamma-ray irradiation (Co-60 at a dose of 2.5 megarads). A further control of homogeneity takes place after the materials have been distributed into bottles. Within-bottle and between bottle homogeneity is
Chapter5
236
determined separately, usually by determining 4~ 137Cs, 9~ and U. When this has been done, the material is announced in the AQCS Catalogue as an intercomparison material. Participants in such intercomparisons are provided with information about the material and special forms on which they are requested to report, for each element, up to six individual net results on a dry-weight basis, the sample weights used, information about the analytical method, and various other items. To preserve anonymity, each participant is assigned a code number, known only to himself and the AQCS programme, by which he is identified in the report that is subsequently prepared on the results of the intercomparisons. The number of participants in each intercomparison varies but at present is around fifty. A chronological list of materials for intercomparisons which have been organised by the AQCS programme during a nine-year period is given in Table 5.17. Table 5.17 IAEA intercomparison exercises involving radionuclides during the nine years: 1983-1992 Matrix
Level
IAEA code
(year)
Certified as RM
Soil
environmental environmental environmental environmental
AG-B-1 A-14 SD-N-1/2 Soil-6
(1983) (1983) (1983) (1983)
+ + + +
Sediment marine Fish flesh
environmental environmental
SD-N-2
(1983)
+
MA-B-3/RM
(1986)
+
Sediment, deep sea
environmental
SD-A-1
(1986)
+
Sediment, like Air-filter, simulated Milk powder
environmental artificial
SL-2 IAEA-083 IAEA-152
(1986) (1986) (1987)
+ + +
IAEA- 154
(1987)
+
IAEA-312 IAEA-313
(1988) (1988)
+ +
environmental elevated a
IAEA-314 IAEA-321 IAEA- 156
(1988) (1988) (1988)
+ + +
Alga, marine Milk powder Sediment, marine
Whey powder Soil Sediment, stream Sediment, stream Milk powder Clover Seaweeds, mediterranean
elevated a elevated a environmental environmental environmental
elevated a
IAEA-308
(1988)
+
Sediment, Baltic Sea
elevated a
Sea plant, posidonia oceanica Uranium ore, phosphate Tuna homogenate, Mediterranean
elevated a environmental
lAEA-306 lAEA-307
(1988) (1988)
+ +
Sediment, Pacific Ocean
elevated a
IAEA-364 IAEA-352 IAEA-368
(1989) (1989) (1990)
+ + +
Soil Grass Cockle flesh Sediment, marine
elevated a elevated a environmental environmental
IAEA-375
(1991-92)
-
IAEA-373 IAEA- 134 IAEA- 135
(1991-92) (1992) (1992)
-
natural
aContaminated with radioactive fallout from Chernobyl.
Measurements of Radioactivi~.,
237
The results submitted by the participants are in all cases evaluated by the AQCS programme. A specific feature of any intercomparison is that gross errors occur quite frequently and results differing by as much as two or three orders of magnitude may be reported by participating laboratories. Various approaches and criteria for the detection and rejection of the highest and the lowest values or outliers have been discussed in the literature. The analytical data received in intercomparison exercises by the AQCS programme are treated using two different methods in order to derive a consensus value, which is considered to be a reliable estimate of the true value. Applying the first method, four different criteria, namely Dixon' s test, Grubbs' test, the coefficient of dewness test and the coefficient of kurtosis test are used at a significance level of o~ = 0.05. If a laboratory mean for each element as single unweighted value was declared to be an outlier by any criterion, it is rejected and the whole procedure repeated until no more outliers could be identified. The remaining laboratory means are then combined in the usual way to provide estimates of the overall mean (consensus value) and its associated standard deviation, standard error and 95% confidence interval. The consensus values cannot automatically be accepted as recommended to certified values because their analytical validity usually requires a re-assessment in the light of additional analytical information such as concentration level, number of different analytical methods used, percent of outliers and other criteria. In practice, certified or recommended values are always based on the following requirements: data should be available from a certain number of participants and two or more different analytical methods; there should be no significant differences between the groups of accepted results; outliers should not exceed 20-30% of the submitted results. Depending on the extent to which the data satisfy such acceptance criteria, the consensus values are then assigned to one of the following conclusions: certified or recommended concentration, information value, or not recommended. The Agency's Analytical Quality Control Services (AQCS) programme provides mainly four types of materials: 9 materials which can be used in analytical laboratories working in the fields of nuclear technology and isotope hydrology. These include uranium ore reference materials and other substances of interest for nuclear fuel technology as well as stable isotope reference materials for mass spectrometric determination of isotope ratios in natural waters; 9 materials with a known content of uranium, thorium and/or transuranium elements or fission products for the determination of environmental radioactivity or control of nuclear safety; 9 materials for use in the determination of stable trace elements in environment, biomedical and marine research; 9 materials which can be used in analytical laboratories working in the fields of monitoring organic microcontaminants in the marine environment. Many countries practise national intercomparison programs. For example, the Japanese nationwide intercomparison program is based on the following:
238
Chapter 5
a. Comparison method Two methods of comparison, the "sample dividing method" and the "reference sample method", were adopted for comparing the results of radionuclide analysis. b. Item for analysis and measurement method Gamma spectrometry is used. Participating laboratories are requested to determine artificial radionuclides as 4~ 54Mn, 59Fe, 6~ 131I, 137Cs, ~44Ce, for the , 'reference sample method", but as 40K and 137Cs for the "sample dividing method". c. Samples and materials for intercomparison The environmental samples are soil, milk and crops. The reference samples are agar gel, alumina powder and liquid milk, which are all spiked with known radioisotopes. 5.5.2 Reference materials
All of the IAEA reference materials which are currently available have been certified on the basis of previously conducted intercomparison exercises. Natural matrix reference materials with certified values for the activities of various radionuclides are listed in Table 5.18. Some of the materials listed in Table 5.18 are the first "post-Chernobyl" natural matrix radionuclides reference materials that are internationally available. Including those reference materials available before Chernobyl, activities range for ~37Csfrom 0.8 Bq/kg (marine sediment, IAEA SD-N-2) to 3.7 kBq/kg (whey powder, IAEA-154). Ideally, there is a need for several reference materials which have a similar matrix type to the samples being analysed and which contain a concentration of the analyte representative of the whole working range that is of interest. Table 5.19 lists stream sediments and milk powder reference materials which reflect the fact that such materials have a different level of activity with practically the same matrix type. Reference materials for radioactivity measurements can also be obtained from the following specialised international or national organisations. 1. Central Bureau for Nuclear Measurements, Commission of the European Communities, Joint Research Centre, Geel (Belgium). 2. Office des Rayonnements Ionisants Commissariat ?a l'Energie Atomique BP 21, 91910, Gif-Sur-Yvette (France). 3. Commission d'Etablissement des Methodes d'Analyse Commissariat ?a l'Energie Atomique BP 6, 92265, Fontenay aux Roses (France). 4. AEA Fuel Services, Chemistry Division, Harwell Laboratory, Oxfordshire OX11 0EA (UK). 5. New Brunswick Laboratory, US Department of Energy 9800 South Cass Avenue, Argonne, IL 60439-4899 (USA). 6. All Union Foreign Economic Association "Techsnabexport", Staromonetniy Per. 26, 109180, Moscow (Russia).
Measurements of Radioactivi~
239
Table 5.18 IAEA Reference materials for measurements of natural and fallout radioactivity in environmental and food samples Matrix
Analytes
IAEA code
Sediment, lake
K-40, CS- 137
SL-2
Sediment, stream
Ra-226, Th, U
IAEA-313, IAEA, 314
Soil
Ra-226, Th, U
IAEA-312
Soil
Sr-90, Cs-137, Ra-226, Pu-239
Soil 6
Bone, animal
Sr-90, Ra-226
A-12
Clover
K-40, Sr-90, CS- 134, Cs- 137
IAEA- 156
Milk powder
K-40, Sr-90, Cs- 137
A- 14
Milk powder
K-40, Sr-90, Cs-134, Cs.137
IAEA-152
Milk powder
K-40, Sr-90, Cs- 134, Cs- 137
IAEA- 154
Whey powder
K-40, Sr-90, CS- 134, Cs- 137
IAEA- 154
Fish, flesh
K-40, Cs- 137
MA-B-3/RM
Seaweeds, Mediterranean
K-40, Ru- 106, Ag- 110m, Cs- 134, Cs- 1 3 7 , Pb-210, Th-228, Pu-238, Pu-239+240, Am-241
IAEA-308
Sea-plant, posidonia
K-40, Ru-106, Ag-110m, Cs-134, Cs-137, Ra-226, Pu-238, Pu-230+240, Am-241
IAEA-307
oceanica Sediment, marine
K-40, Cs- 137, Th-232, Pu-239+240
SD-N-2
Sediment, Pacific Ocean
Co-60, Sr-90, Cs-137, Pu-239+240
IAEA-367
Sediment, Pacific Ocean
Co-60, Eu-155, Pb-210, Ra-226, Pu-238, U-238, Pu-239-240
IAEA-368
Tuna homogenate, Mediterranean
K-40, Cs-137, Pb-210, Po-210
IAEA-352
Water, Pacific Ocean
Sr-90, Cs-137, Pb-210, Po-210
IAEA-352
Table 5.19 Reference materials of a similar matrix with different levels of analytes Analytes
Activity (Bq/kg)
Reference date
Matrix
Code
Ra-226
342 732
30.01.88 30.01.88
Stream sediment Stream sediment
IAEA-313 IAEA-314
Sr-90
1.5 3.3 7.7
31.08.87 01.01.90 31.08.87
Milk powder Milk powder Milk powder
A-14 IAEA-321 IAEA-152
Cs- 137
1.79 72.6 2159
31.08.87 01.01.90 31.08.87
Milk powder Milk powder Milk powder
A-14 IAEA-321 IAEA- 152
Chapter 5
240
The IAEA AQCS Programme provides three main types of material. 9 Materials that can be used in analytical laboratories working in the fields of nuclear technology and isotope hydrology. These include uranium ore reference materials and other substances relevant to nuclear fuel technology as well as stable isotope reference materials for mass spectrometric determination of isotope ratios in natural waters. 9 Materials with known contents of uranium, thorium and/or transuranic elements or fission products for the determination of environmental radioactivity or control of nuclear safety. 9 Materials for use in the determination of stable trace elements in environmental or biomedical research. Radiochemical methods such as neutron activation or isotope dilution analysis, are often used in the determination of such trace elements and constitute an important contribution of nuclear techniques to applied science (Strachnov et al., 1993). Table 5.20 lists the radionuclides referenced by IAEA, their activity, matrix, and sample code. Table 5.20 includes also materials of marine origin (Ballestra et al., 1992). The intercomparison samples cover a range of materials and contain radionuclides with very different levels. IAEA intercalibration exercises are conducted with the involvement of many laboratories. As an example, Fig. 5.52 shows the results of an intercomparison run for 13VCsdetermination in milk powder. Some laboratories had difficulties in determining
2800
2400
0oooo+ oooooo ~
2000
o~ e~ O0
r
0
1600
1200
800
400
36'
' 5' ' i 6 ' ' i 3 ' '34" '10 w '12' ' i 4 ' 29 37 4 398 6 28 35
'30' 6
'27' '26' '17' '31' ' 2' '11' 33 32 20 21 25 39A
' 9' 'I8'A '22 r 'l~J 1 23 15 3
LAB.CODE NO. Fig. 5.52. Results of IAEA-152 intercomparison 137Csdeterminations. Recommended v a l u e - 2065 Bq/kg; confidence interval - 1991-2143 Bq/kg.
241
Measurements of Radioactivity
Table 5.20 Radionuclides referenced by IAEA Ref. analyte
Activity or conc.
Confidence (Bq/kg)
Matrix interval
Reference date
Sample code
40K
391
379-405
1 Jan. 1989
IAEA-352
527 150 220 240 272 539 552 657 1381 1575•
510-543 141-161 189-226 211-269 252-299 510-574 563-569 637-676 1320-1456 1511-1644
Tuna homogenat e, Mediterranean Milk powder Sea-plant, Posidonia oceanica Sediment, marine Sediment, lake Fish flesh Milk powder Milk powder Clover Seaweeds, Mediterranean Hay powder
31 Aug. 1987 1 Jan. 1988 1 Jan. 1985 31 Jan. 1986 1 Jan. 1986 31 Aug. 1987 1 Jan. 1990 1 Aug. 1986 I Jan. 1988 31 Aug. 1987
A-14 IAEA-307 SD-N-2 SL2 MA-B-3/RN IAEA-152 IAEA-321 IAEA- 156 IAEA-308 IAEA- 154
9~
1.5 3.3 6.9 7.7 14.8 30.34 54.8
1.33-1.57 3.16-3.44 6.0-8.0 7.0-8.3 13.4-16.3 24.2-31.67 46.3-59.2
Milk powder Milk powder Hay powder Milk powder Clover Soil Bone, animal
31 Aug. 1987 1 Jan. 1990 31 Aug. 1987 31 Aug. 1987 1 Aug. 1986 30 Jan. 1983 15 Dec. 1981
A-14 IAEA-321 IAEA-154 IAEA- 152 IAEA-156 SOIL-6 A-12
106Ru
23 33.5
22-25 30.0-36.5
Seaweeds, Mediterranean Sea-plant, Posidonia oceanica
1 Jan. 1988 1 Jan. 1988
IAEA-308 IAEA-307
ll0mAg
20 5.1
1-2.27 4.8-5.5
Seaweeds, Mediterranean Sea-plant, Posidonia oceanica
1 Jan. 1988 1 Jan. 1988
IAEA-308 IAEA-307
1.6 1.6 15.5 132 764 1355
1.5-1.8 1.5-1.9 14.8-16.2 126-138 722-802 1295-14 17
Seaweeds, Mediterranean Sea-plant, Posidonia oceanica Milk powder Clover Milk powder Whey powder
1 Jan. 1988 1 Jan. 1988 1 Jan. 1990 1 Aug. 1986 31 Aug. 1987 31 Aug. 1987
IAEA-308 IAEA-307 IAEA-321 IAEA- 156 IAEA- 152 IAEA-154
2.7
2.5-2.8
1 Jan. 1989
IAEA-352
0.8 1.79 2.4 4.9 5.6 14.2 53.65 72.6 264 2159 3749
0.5-1.0 1.62-1.97 22-2.6 4.5-5.2 5.3-6.0 13.7-15.3 51.43-57.91 71.1-74.2 254-274 2503-22 09 3613-38 87
Tuna homogenate, Mediterranean Sediment, marine Milk powder Sediment lake Sea-plant, Posidonia oceanica Seaweeds, Mediterranean Fish flesh Soil Milk powder Clover Milk powder Hay powder
1 Jan. 1985 31 Aug. 1987 31 Aug. 1986 1 Jan. 1988 1 Jan. 1988 1 Jan. 1986 30 Jan. 1983 1 Jan. 1990 1 Aug. 1986 31 Aug. 1987 31 Aug. 1987
SD-N-2 A-14 SL-2 IAEA-307 IAEA-308 MA-B-3/RN SOIL-6 IAEA-321 IAEA- 156 IAEA- 152 IAEA- 154
134Cs
137Cs
continued
242
Chapter 5
Table 5.20 (continuation)
Ref. analyte
Activity or conc.
Confidence (Bq/kg)
Matrix interval
Reference date
Sample code
21opb
0.6
0.36-1.0
1 Jan. 1989
IAEA-352
73
66-75
Tuna homogenate, Mediterranean Seaweeds, Mediterranean
1 Jan. 1988
IAEA-308
2.2
1.7-27
Tuna homogenate, Mediterranean
1 Jan. 1989
IAEA-352
3.1 5.2 79.92 269 342 732
21-4.4 4.4-6.7 69.56-93-43 250-287 307-379 678-787
Sea-plant, Posidonia oceanica Bone, animal Soil Soil Sediment, stream Sediment, stream
1 Jan. 1988 15 Dec. 1981 30 Jan. 1983 30 Jan. 1988 30 Jan. 1988 30 Jan. 1988
IAEA-307 A- 12 SOIL-6 IAEA-312 IAEA-313 IAEA-314
228Th
25
2.2-3.6
Seaweeds, Mediterranean
1 Jan. 1988
IAEA-308
232Th
4.9
4.5-5.4
Sediment, marine
1 Jan. 1985
SD-N-2
0.017 0.025
0.016-0. 023 0.022-0. 028
Seaweeds, Mediterranean Sea-plant, Posidonia oceanica
1 Jan. 1988 1 Jan. 1988
IAEA-308 IAEA-307
1.04
0.962-1. 11
Soil
30 Jan. 1983
SOIL-6
8.8 0.50 0.72
6.51-4.0 0.46-0.52 0.66-0.7 9
Sediment, marine Seaweeds, Mediterranean Sea-plant, Posidonia oceanica
1 Jan. 1985 1 Jan. 1988 1 Jan. 1988
SD-N-2 IAEA-307 IAEA-307
0.036 0.17
0.030-0. 050 0.16-0.2 5
Sea-plant, Posidonia oceanica Seaweeds, Mediterranean
1 Jan. 1988 1 Jan. 1988
IAEA-307 IAEA-308
21~ 226Ra
238pu
239pu 239pu
241Am
aNote: The 232Th is in equilibrium with 228Ra and 228Th.
the activity level. This situation is rapidly improving with time" Table 5.21 shows the improvements in the quality of the work at the participating laboratories. This is also seen in Table 5.22 where the mean values and relative standard deviations of three intercomparison runs for the 9~ determination in simulated air filters are presented. The intercomparison exercises show a need for greater standardisation of the analytical techniques used for radionuclide determination. This is indicated in McGee (1992), where the bias and measurement errors in radioactivity data from four European radiation research laboratories were reported. Within the framework of the International Chernobyl Project, the IAEA's Seibersdorf Laboratories organised an intercalibration exercise (Cooper et al., 1992)
Measurements of Radioactivity
243
Table 5.21 Determination of
1983 1989
137Csin the same milk powder during intercomparisons in 1983 and 1989 Mean value (Bq/kg)
Rel. SD (%)
% outliers/lab.
2.08 1.70
45 19
10 0
Table 5.22 Determination of Sr-90 in simulated air filters during intercomparison runs in 1973, 1976, 1988
1973 1976 1988
Mean value
Rel. S.D.
178 Bq/filter 179.5 Bq/filter 231 Bq/filter
27% 17% 3%
Table 5.23 Comparison of performance of the two groups of laboratories: worldwide vs. Soviet Union Radionuclide
137Cs
134Cs 4~ 9~
Range of reported values for milk(H) (Bq/kg) Worldwide
USSR
469.3-2491.3 58.0-652.5 103.6-3650.0 5.53-8.54
175-3070 184.7-542.5 429-4959 1.43-68.8
among some of the laboratories which were involved in assessing the environmental contamination in the former USSR caused by the accident. The objective was to assess the reliability of the radioanalytical data for food and environmental samples, which were used to assess the doses. The initial study reference materials from the stocks of the IAEA' s Analytical Quality Control Service (AQCS) were re-labelled and submitted to 71 laboratories as blind samples in June and July of 1990. These natural matrix materials included samples of milk (containing two different levels of radioactivity), soil, air filters and clover. The concentrations of radionuclides (137Cs, 134Cs, 4~ 9~ 239pL1, 226Ra, 6~ 133Ba, 2~~ in these samples were known from previous intercalibration exercises. The overall range in performance was broad, which is as observed in previous international intercomparisons. This is illustrated in Table 5.23 where the results of the
244
Chapter 5
original IAEA intercomparison run (worldwide) and former Soviet Union laboratories, for high level (H) milk are presented. The Central Service for Protection against Ionising Radiation (SCPRI), a service of the French Ministry of Public Health, National Institute of Health and Medical Research, was nominated at the end of 1969 as the International Reference Centre (IRC) of the World Health Organisation for Radioactivity measurements. Four laboratories in the world have been officially designated as WHO collaborating laboratories. These laboratories are: 9 Radiation Protection Bureau in Ottawa (Canada). 9 National Institute of Radiation Protection in Stockholm (Sweden), 9 Environmental Monitoring and Support Laboratory (EPA) in Las Vegas (USA), 9 National Radiation Laboratory in Christchurch (New Zealand). At the present time, 28 laboratories from 17 countries are interested in the WHO-IRC Intercomparisons. Its program of intercomparison shows the following characteristics: 1. the radioactivity of the samples is the present environmental monitoring level; 2. generally, the samples present real radioactivity due to the fallout or releases of nuclear facilities; 3. the amount of the product provided allows several tests to be carried out; 4. standard materials can be provided; 5. a preliminary study of the results of each intercomparison is given to the participants as soon as possible. Table 5.24 shows concrete contents of Intercomparisons which WHO has carried out so far. The first column of Table 5.24 indicates periods when samples were sent to participating laboratories, the second their nature, the third, their numbers, the fourth, nuclides and stable elements to be measured and determined for the intercomparison purposes, and the fifth and last column, the scope of radioactivity levels in the samples. As demonstrated by this table, a wide variety of samples has been adopted since 1970, among which are liquid milk, animal bones, human bones, foods, low-level radioactive liquid waste, ground water, mineral water, fiver sediment, seaweed, pond water, fresh water fish, cereals, seawater, rain water, drinking water, soil, and vegetation. Concerning the general conditions of the intercomparison programs in progress and the results obtained, it can be noted that: 9 The IRC has diversified its program by introducing new categories of samples (waters from various origins, sediments, fish, seaweed, liquid waste, cereals, and soil, etc.) in which laboratories involved in environmental monitoring of nuclear power plants are interested. 9 With regard to the quality of the analyses, the situation presented in the preceding annual report has not evolved much. 9 The regular participation of different laboratories in the intercomparisons provides a comprehensive view of their technical capabilities and of the quality of their analytical work.
Measurements of Radioactivio'
245
Table 5.24 WHO IRC Intercomparisons Radioactivity level
Period of dispatch
Nature of the sample
No. of the Proposed sample determinations
June 1970
liquid milk
A010
9~
137Cs, Ca, K
9~ pCi/1
Feb. 1971
liquid milk
A338
9~
137Cs, Ca, K
9~ -20 pCi/l, 137Cs -20 pCi/1
June 1 9 7 1
animal bones
A504 A505
9~
Sr, Ca
9~ 9~
Feb. 1972
human bones
A806
9~
Sr, Ca
9~ <0.5 pCi/g-ash
Nov. 1972
dried total diet
B078
9~ 137Cs, Sr, Ca, K
9~ -40 pCi/kg-dry 137Cs -90 pCi/kg-dry
July 1973
liquid milk
B845
9~
9~ -20 pCi/1, 137Cs -30 pCi/1
April 1974
ground water low C140 level liquid waste C 141
3H
3H - 10,000 pCi/1 3H - 150,000 pCi/1
Feb. 1975
rain water spiked ground water
C617 C618
3H
3H - 1,000 pCi/l 3H -5,000 pCi/1
March 1976
mineral water
D256
gross beta, K nat. U, 226Ra
nat. U -20 ~tg/1, 226Ra -15 pCi/1
June 1976
river water
D402
3H, 9~
Sept. 1976
liquid milk
D499
9~
Dec. 1976
river sediment
D601
activation and fission products (54Mn, 58C0,
137Cs, Ca, K
l~
+l~
,
1255b ' 137Cs.." 137Cs, Ca, K
pCi/1, 137Cs-120
pCi/g-ash pCi/g-ash
3H <5,000 pCi/1 l~176 y 1,000 pCi/1 9~ 125Sb, 137Cs <50 pCi/1 9~ -10 pCi/1, 137Cs -15 pCi/1 5,000-20,000 pCi per kg-dry for each radionuclide
6~ ' 9~ ' 134Cs ' 137Cs...) March 1977
animal bones
D736 D737
90Sr, Sr, Ca
9~ 9~
July 1977
sea fish
D925
9~
pCi/kg-dry 134Cs -400 pCi/kg-dry 137Cs .-.4,000 pCi/kg-dry
134Cs, 137Cs, Sr, Ca, 9~
K Dec. 1977
marine sediment
E114
pCi/g-ash pCi/g-ash
9~ ' 95Zr+95Nb' l~ ' l~ +l~ ' l l~ '
1255b ' 137Cs ...
lO6Ru+106Ru' 144Ce+ 144pr < 10,000 pCi/kg-dry other nuclides <500 pCi/kg-dry
continued
246
Chapter 5
Table 5.24 (continuation) Period of dispatch
Nature of the sample
No. of the Proposed sample determinations
Radioactivity level
March 1978
liquid milk
E414
9~
9~ pCi/1
June 1 9 7 8
seaweed
E468
fission products
20-- 10,000 pCi/kg-dry according to the radionuclides
Oct. 1978
pond waters
E888 E889
3H
3H -5,000 pCi/1 3H -20,000 pCi/1
March 1979
low level liquid waste
F100
3H, 54Mn, 58Co, 6~
3H -2,5 ~tCi/1, 54Mn -800 pCi/l 58Co -50,000 pCi/1 58Co -2,500 pCi/1
June 1 9 7 9
liquid milk
F140
9~
9~ -40 pCi/1, 137Cs -250 pCi/1
Nov. 1979
fresh water fish
F290
U, 226Ra, 9~
March 1980
cereals (wheat)
F553
9~
June 1980
sea water
F712
fission products
Oct. 1980
rain water
F856
3H
3H -2,500 pCi/1
Feb. 1981
liquid milk
G041
9~ 137Cs,Ca, K
9~ pCi/1
June 1981
drinking water
G336
3H, 9~
3H <10,000 pCi/1 1o6 R u - 7 5 pC i/1 9~ pCi/1
Oct. 1981
soil
G477
9~ 137Csnatural radionuclides
Feb. 1982
total diet
G660
9~
June 1 9 8 2
vegetation
G730
7Be, 9~
~37Cs,Ca, K
137Cs 137Cs
137Cs
l~
pCi/1, 137Cs -30
U -1,500 ~tg/kg-dry 226Ra - 3,500 pCi/kg-dry 9~ -6,000 pCi/kg-dry 137Cs - 1,500 pCi/kg-dry 9~
137Cs-20 pCi/kg-dry
total activity -5,000 pCi/l
pCi/l,
137Cs-300
9~ -200 pCi/kg-dry 137Cs -500 pCi/kg-dry
137Cs, Ca, K, nat. U 9~ Bq/kg-dry (-10 pCi/kg-dry) 137Cs-0.4 Bq/kg-dry (-10 pCi/kg-dry) 137Cs, Ca, K
9Be -200 Bq/kg-dry (-5000 pCi/kg-dry) 9~ Bq/kg-dry (-500 pCi/kg-dry) 137Cs -5 Bq/kg-dry (-100 pCi/kg-dry)
Measurements of Radioactivit3,
247
Period of dispatch
Nature of the sample
No. of the sample
Proposed determinations
Dec. 1982
mineral water
G972 G973
nat
Feb. 1983
liquid milk
H071
9~
June 1983
sea fish
H264
9~ 137Cs,Ca, K, Sr
9~ Bq/kg-dry (-50 pCi/kg-dry) 13VCs-20 Bq/kg-dry (-500 pCi/kg-dry)
Feb. 1984
river sediment
H519
activation and fission products (54Mn' 58Co' 6~ ' 9~ '
5-500 Bq kg -~ dry, according to the radionuclides
U, 226Ra
rain water ground water
40 PM 300 3H 41 P 300 3H
Feb. 1985
liquid milk
42 L 300
U 50 - 100 ~tg/1 2Z6Ra -0.4 Bq/1 ( -10 pCi/l)
137Cs,Ca, K
134Cs' 137Cs' l~ June 1984
Radioactivity level
9~ Bq/1 (~pCi/1) 137Sr-0,3 Bq/1 (~pCi/1)
) ~200 Bq 1-l ~10,000 Bg 1-1
3H 9~
~100 Bq 1-1 -0.3 Bg 1-l ~0.6 Bq 1-1
137Cs May 1985
low level 43 E 300 radioactive liquid effluent
3H gamma emitters (54Mn, 58Co, 6~ 137Cs)
3H ~5• 105 Bq 1-I 54Mn - 1x 102 Bq 1-1 134Cs, 58C0 ~2x103 Bq 1-l 6~ ~2x103 Bq 1-l 134Cs -lxl02 Bq 1-1 137Cs-lxl02 Bq 1-l
Feb. 1986
urine
44 UR 300 3H
-500,000 bq 1-1
May 1986
aquatic plants
45 V 300
10 --2,000 Bq kg -l dry, according to the radionuclides
April 1987
liquid milk
46 L 3000 9~
activation and fission products (54Mn, 58Co, 6~ 9~ 137Cs, l~ natural radionuclides
134Cs 137Ca
0 -1 Bq 1-l 0 -50 Bq 1-1 0 - 1 0 0 Bq 1-1
Oct. 1987
shelled hazelnuts
47 V 300
134Cs 137Cs
0 -200 Bq kg -1, according to the radionuclides
April 1988
dry may blossom
48 V 300
134Cs
0 -5,000 Bq 1-l
137Cs Oct. 1988
mineral water
49 SH 300 226Ra
<2 Bq 1-1 5 -100 ~tg 1-1
continued
Chapter 5
248
Table 5.24 (continuation) Radioactivity level
Period of dispatch
Nature of the sample
No. of the Proposed sample determinations
April 1 9 8 9
radioactive effluent
50 E 300
3H 54Mn, 58C0, 6~ 134Cs, 137Cs
Nov. 1989
herba drosera longifolia
51 V 300 52 V 300
228Th,228Ra, 232Th
Aug. 1990
river sediments
53 SD 300 activation and fission products
300 Bq kg -1 dry
Dec. 1990
seaweeds
54 AL 300 transuranium elements
5 Bq 1-l
May 1991
river sediments
55 SR 300 natural radionuclides
103 -- 106 Bq 1-1 124Sb, 0 "10 4 Bq 1-1 for each nuclide 5x 102 Bq kg-1 dry for each nuclide
REFERENCES Abuzwida, M., Abouzreba, S., Almedhem, B., Zolotarev, Y.A. and Komarov, N.A., Selective liquid scintillation method of plutonium or-spectrometry. J. Radioanal. Nucl. Chem., 111 (1987) 3-10. Albrecht, E. and Kaul, Continuous registration of airborne radioactivity in nuclear operations. IAEA, Vienna, 1967. Amoudry, F. and Burger, P., Determination of the 239pu/24~ isotopic ratio by high resolution alpha spectrometry. Nucl. Instrum. Meth., 233 (1984) 360. Amoudry, F. and Burger, P., Nucl. Instr. Meth. Phys. Res., 223 (1984) 360. Arnold, D., Jagielak, J., Kolb, W., Pietruszewski, A., Wershofen, H. and Zarucki, R., Practical experience in and improvements to aerosol sampling for trace analysis of airborne radionuclides in ground level air. PTB-Report PTB-Ra-34, Physikalishc-Technische Bundesanstalt, Braunschweig, Germany, ISSN 0341-6747, ISBN 3-89429-436-1. Baeza, A., Corvo, G., delRio, M., Miro, C. and Paniagua, J.M., A program ESPEC for the analysis of gamma spectra of environmental samples. Appl. Radiat. Isot., 43 (1992) 833-839. Ballestra, S., Lopez, J.J., Gastand, J., Parsi, P., Vas, D. and Noshkin, V., Twenty years of IAEA-ILMR radionuclide intercomparison programmes. Appl. Radiat. Isot., 43 (1992) 161-173. Baumann, H., Solid sampling with inductively coupled plasma-mass spectrometryma survey. Fresenius J. Anal. Chem., 342 (1992) 907. Baxter, M.J., Burrell, J.A., Crews, H.M. and Massey, R.C., Aluminum levels in milk and infant formulae. Food Additives Contam. 8(5) (1991) 653-660. Baxter, M.J., Burrell, J.A., Crews, H.M., Massey, R.C. and McWeeny, D.J., A procedure for the determination of lead in green vegetables at concentrations down to 1 ~tg/kg. Food Additives Contam. 6(3) (1989) 341-349. Baxter, M.J., Burrell, J.A., Crews, H.M., Smith, A. and Massey, R.C., Lead contamination during domestic preparation and cooking of potatoes and leaching of bone-derived lead on roasting, marinating and boiling beef. Food Additives Contam., 9(3) (1992) 225-235. Beals, D.M., Determination of Technetium-99 in aqueous samples by isotope dilution inductively coupled plasma-mass spectrometry. Proceedings of 3rd Annual Conference on Nuclear and Radiochemistry, Vienna, Sept. 1992.
Measurements of RadioactiviO'
249
Beauchemin, D., A comparison of lCP atomic spectrometry techniques. Spectroscopy, 7(7) (1992) 12. Bell, A.G., Am. J. Sci., 20 (1980) 305. Benkens, R.P., Radiocarbon, 32 (1990) 335-339. Bernabee, R.P., Percival, D.R. and Hindman, F.D., Liquid-liquid extraction separation and determination of plutonium and americium. Anal. Chem., 52 (1980) 2351-2353. Bertin, E.P., Principles and Practice of X-ray Spectrometric Analysis. Plenum, New York, 1978. Birks, J.B., The Theory and Practice of Scintillation Counting. Pergamon Press, London, 1964, 662 pp. Bland, C.J., Truffy, J. and de Bruyne, T., Deconvolution of alpha-particle spectra to obtain plutonium isotopic ratios. Appl. Radiat. lsot., 43 (1992) 201-209. Bolshov, M.A., Zybin, A.V. and Smirenkim, I.I., Atomic fluorescence spectroscopy with laser excitation. Spectrochim. Acta, 36B, (1988) 1143-1152. Bortels, G. and Collaers, P.J., Analytical functions for fitting peaks in alpha-particle spectra. Int. J. Appl. Radiat. Isotop., 38 (1987) 831. Brodzinski, R.L. and Perkins, R.W., Drum assayer for measuring radionuclides in foods. ICRM, IAEA, Vienna, 1992. Brost, R.A. Feichter, J. and Heimann, M., Three-dimensional simulation of 7Be in a global climate model. J. Geophys. Res., 96 (1991 ) 22423-22445. Brown, R.M., Long, S.E. and Pickford, C.J., The measurement of long lived radionuclides by non-radiometric methods. Sci. Total Environ., 70 (1988) 265-274. Budnitz, R.J., Plutonium: A Review of Measurement Techniques for Environmental Monitoring, Rep. LBL-2039, Lawrence Berkeley Lab., CA (1973). Budnitz, R.J., Tritium instrumentation for environmental and occupational monitoring--A review. Health Phys., 26 (1974) 165. Bulman, R.A. and Cooper, J.R., (ed.), Speciation of Fission and Activation Products in the Environment. Elsevier, London and New York, 1985. Burton, W.M. and Stewart, N.G., Use of long lived radioactivity as an atmospheric tracer. Nature, 186 (1960) 584-589. Bush, E., A double ratio technique as an aid to selection of sample preparation procedures in liquid scintillation counting. Int. J. Appl. Radiat. Isot., 19 (1968)447-452. Carmon, B. and Dyer, A., Assay for 9~ and 89Sr in aqueous nitrate solution in the presence of '37Cs. J. Radioanal. Nucl. Chem., 98 (2) (1986) 265-273. Cartwright, B.G. and Shirk, E.K., Nucl. Instr. Meth., 153 (1978)457--460. Chase, G.D. and Rabinowitz, J.L., Principles of Radioisotope Methodology, 3rd edn. Burgess Publishing Co., Minneapolis, 1967, 633 pp. Chieco, N.A. Bogen, D.C. and Knutson, E.O. (eds.), EML Procedures Manual, 27th edn., 1992. Chieco, N.A., Bogen, D.C. and Knutson, E.Q. (eds.), Environmental measurements laboratory (EML) Procedures Manual, HASL-300-Ed. 27-vo1.1 (USDOE Environmental Measurements Lab., New York, NY (USA) (1990) 783 p. Chu, K.D. and Hopke, P.K., APCA paper No. 85-85.5 Pittsburgh, PA, 1985. Chu, N., Plutonium determination in soil by leaching and ion exchange separation. Anal. Chem., 43 (1971) 449-452. Cohen, B.L. and Nason, R., Health Phys., 50 (1986) 457-463. Colles, M.J., Geddes, N.R. and Mehdizadeh, E., Contemp. Phys., 20 (1979) 11. Cooper, E.L., Valkovic, V., Strachnov, V., Dekner, R. and Danesi, P.R., Results of the intercalibration study of laboratories involved in assessing the environmental consequences of the Chernobyl accident. Appl. Radiat. Isot., 43 (1992) 149. Countess, R.J., Health Phys., 31 (1970) 445-456. Cox, R.J., Pickford, C.J. and Thompson, M., Determination of Iodine-129 in vegetable samples by inductively coupled plasma mass spectrometry. J. Anal. Atmos. Spectrom., 7 (1992) 635. Crain, J.S. and Gallimore, D.L., Inductively coupled plasma-mass spectrometry of synthetic elements: 99Tc. Appl. Spectros., 46(3) (1992) 547. Crespo, M.T., Gascon, J.L. and Acena, M.L., Procedures for the assay of alpha-particle emitters in water samples. Appl. Radiat. Isot., 43 (1992) 19-28.
250
Chapter 5
Crews, H.M., Baxter, M.J., Bigwood, T. et al., Lead in feed incident--multielement analysis of cattle feed and tissues by ICP-MS and co-operative quality assurance scheme for lead in milk. Food Additives Contam., 9(4) (1992) 365-378. Crews, H.M., Dean, J.R., Ebdon, L. and Massey, R.C., Application of HPLC-ICP-MS to the investigation of cadmium speciation in pig kidney following cooking and in vitro gastro-intestinal digestion. Analyst, 114 (1989) 895-899. Crews, H.M., Dean, J.R., Ebdon, L. and Massey, R.C., The use of ICP-MS in trace element speciation studies. Paper no. E 1.5, presented at CSI XXV, Toronto, Canada, June 21-26, 1987. Crews, H.M., Massey, R., McWeenny, D.J. and Dean, J.R., Trace element speciation in food: a combined anzymolysis-SEC-ICP-MS approach. J. Res. Nat. Bur. Stand., 93(3) (1988) 349-350. Currie, L.A., Anal. Chem., 40 (1968) 586. Dean, J.R., Ebdon, L., Crews, H.M. and Massey, R.C., Characteristics of flow injection-ICP-MS for trace metal analysis. J. Anal. Atom. Spectrom., 3 (1988) 349-354. Dean, J.R., Ebdon, L. and Massey, R., Selection of mode for the measurement of lead isotope ratios by ICP-MS and its application to milk powder analysis. J. Anal. Atom. Spectrom., 2 (1987) 369-374. Dean, J.R., Munro, S., Ebdon, L., Crews, H.M. and Massey, R.C., Studies of metalloprotein species by directly coupled HPLC-ICP-MS. J. Anal. Atom. Spectrom., 3 (1987) 607-610. Deberin, K. and Hermer, R.G., Gamma- and X-ray Spectrometry with Semiconductor Detectors. North-Holland, Amsterdam, 1988. Decker, K.M. and Sanderson, C.G., Appl. Radiat. lsot., 42 (1992) 323. Denoyer, E.R. and Stroh, A., Expanding ICP-MS capabilities using flow injection. Am. Lab., 74 (Feb. 1992). Diamond, R.M., Street, Jr., K. and Seaborg, G.T., An ion-exchange study of possible hybridised 5f bonding in the actinides. J. Am. Chem. Sci., 76 (1954) 1461. Dryak, P., Kovar, P., Plchova, L. and Suran, J., Corrections for the Marinelli geometry. J. Radioanal. Nucl. Chem. Lett., 135 (1989) 281-292. Dua, S.K., Kotrappa, P. and Gupta, P.C., Health Phys., 45 (1983) 152-157. Eagles, J. Fairweather-Tait, S.U., Mellon, F.A. et al., Comparison of fast-atom bombardment, thermal ionisation and ICP-MS for the measurement of 64Zn/67Zn stable isotopes in human nutrition studies. Rapid Comm. Mass Spectrom., 3(6) (1989) 203-205. Eakins, J.D., et al., Analytical procedure used by the bioassay section at AERE Harwell, Rep. AERE-AM 103, UK Atomic Energy Authority, Harwell Lab., Oxfordshire (1986). Elmore, D. and Philips, F.M., Science, 236 (1978) 543. Erdman, C.A., Reynolds, A.B., Radionuclide behaviour during normal operation of liquid-metal-coolant fast breeder reactors, Part I: Production. Nucl. Safety, 16 (1975) 43. Fairweather-Tait, S.F., Fox, T.E., Wharf, S.G., Eagles, J., Crews, H.M., Massey, R.C., Apparent zinc absorption byrats from foods labelled intrinsically and extrinsically with 67Zn. Br. J. Nutr., 66 (1991) 65-71. Fantini, M. and Richard, G., Solid State Nuclear Track Detectors. Pergamon Press, Oxford, 1981, pp. 193-196. Feely, H., Helger, I., Juzdan, Z., Klusek, C., Larsen, R., Leifer, R., Sanderson, C. and Dreicer, M., Fallout in the New York Metropolitan area following the Chernobyl accident. J. Environ. Radioact., 7 (1988) 177-191. Feely, H.W., Larsen, R.J. and Sanderson, C.G., Factors that cause seasonal variations in beryllium-7 concentrations in surface air. J. Environ. Radioactivity, 9 (1989) 223-249. Feely, H.W.,, Friend, J.P., Krey, P.W. and Russel, B.A., Flight data and results of radiochemical analysis of filter samples collected during 1961 and 1962. Health and Safety Laboratory, New York, Report No. HASL- 153. Feichter, J., Brost, R.A. and Heimann, M., Three-dimensional modelling of the concentration and deposition of 2'~ aerosols. J. Geophys. Res., 96 (1991) 22447-22460. Ferreri, G. and Surbeck, H., in: Proc. Second Int. Meeting on Low-Level air Radioactivity Monitoring, Madralin, near Warsaw, Poland, 14-18 Feb. 1994. Ferreri, G. and Surbeck, H., Low level air radioactivity measurements in Switzerland, in: Proc. Second Int.
Measurements of Radioactivi~'
251
Meeting on Low-level Air Radioactivity Monitoring, Manderlin, near Warsaw, 14-18 Feb. 1994. Fews, A.P. and Henshaw, D.L., Nucl. Instr. Methods in Phys. Res, 223 (1984) 400. Fox, B.W., Techniques of Sample Preparation for Liquid Scintillation Counting. North Holland, Amsterdam, 1976. Friedlander, G. Kennedy, J.W., and Miller, M., Nuclear and Radiochemistry, 3rd edn. Wiley, New York, 1964. Frolov, E.A., A precision facility for measuring alpha-particle energies and flux density. Appl. Radiat. Isot., 43 (1992) 211-222. Fukai, R., Holm, E. and Ballestra, S., Modified procedure for plutonium and americium analysis of environmental samples using solvent extraction, Rep. No. 187, International Laboratory of Marine Radioactivity, IAEA, Monaco (1976). Gilmore, G. and Hemingway, J.D., Practical Gamma-ray Spectrometry. Wiley, New York, 1995. George, A.C., Health Phys., 64 (1984) 867-872. George, A.C., Scintillation flasks for the determination of low levels of Radon, in: Proc. of Ninth Midyear Health Physics Symp., Denver, CO. Gladney, E.S., Moss, W.D., Gautier, M.A. and Bell, M.G., Determination of U in urine: Comparison of ICP-mass spectrometry and delayed neutron assay. Health Phys., 57(1) (1989) 171. Gonzales, A.J., Bennett, B.G. and Geoffrey, A.M., Mission report: Radiological accident at Tomsk-7, Russian Federation, 6 April 1993, IAEA No. Tomsk-7 sr/2653 (1993). Gove, H.E., Litherland, A.E., Elmore, D. (eds.), Proceedings of the IV International Symposium on Accelerator Mass Spectrometry, Niagara-on-the-Lake, Ontario, Canada 1987, in: Nucl. Instrum. Methods Phys. Res. B, 29 (1987) 1. Greenberg, A. et al., Standard methods for the examination of water and waste-water, 15th Edition, Section 703 (1981). American Public Health Assn./American Water Works Assn./Water Pollution Control Federation Washington, D.C. Gunnick, R., Plutonium isotopic analysis of nondescript sample by gamma-ray spectrometry. In: W.S. Lyon (ed.), Proceedings of the 25th Conference on Analytical Chemistry in Nuclear Technology. 23 Ann Arbor Science Publishers, Gatlinburg, TN, USA (1982). Hampson, B. and Tennant, D., Simultaneous determination of actinide nuclides in environmental materials by solvent extraction and alpha spectrometry. Analyst, 98 (1973) 873-885. Harley, E.P., Krey, P.W. and Volchok, H.L., Global inventory and distribution of fallout plutonium. Nature, 241 (1973) 444. Harley, J.H. (ed.), HASL Procedures Manual HSL-300. N. York, Atomic Energy Commission, 1972, 602 PP. Heimlich, M., Beeley, P.A. and Page, J.A., GAANAL-PC: A program for gamma-ray spectrum analysis using a microcomputer. J. Radioanal. Nucl. Chem., 132 (1989) 281-291. Hieftje, G.M., Towards the next generation of plasma source mass spectrometers. J. Anal. Atmos. Spectrom., 7 (1992) 783. Hindman, F.D., Actinide separations for o~-spectrometry using neodymium fluoride coprecipitation. Anal. Chem., 58 (1986) 1238-1241. Hino, Y. and Kawada, Y., Nucl. Instr. Meth., A286 (1990) 543. Hoffman, D.C., Lawrence, F.O., Mewherter, J.L. and Rourke, F.M., Nature, 234 (1971) 132-134. Holm, E., Ballestra, S. and Fukai, R., A method for ion-exchange separation of low levels of americium in environmental materials. Talanta, 26 (1979). Holm, F., Oregioni, B., Vas, D., Petterson, H., Rioseco, J. and Nilsson, U., J. Radioanal. Nucl. Chem., 138 (1990) 111. Hopke, P.K., Health Phys., 57 (1989) 39-42. Horlick, G., Plasma-source mass spectrometry for elemental analysis. Spectrometry 7(1) (1992) 22. Horrocks, D.L., Applications of Liquid Scintillation Counting. Academic Press, New York, 1974. Horwitz, E.P., Dietz, M.L., Nelson, D.M., LaRosa, J.J. and Fairman, W.D., Concentration and separation of actinides from urine using a support bifurcational organophosphorus extractant. Anal. Chim. Acta, 238 (1990) 263-271. Houk, R.S. and Thompson, J.J., Trace metal isotopic analysis of microlitre solution volumes by ICP-MS.
252
Chapter 5
Biomed. Mass. Spectrom., 10 (1983) 107-112. Hubbel, J.H., Photon mass attenuation and energy-absorption coefficients from 1 keV to 20 MeV. Int. J. Appl. Radiat. Isot., 33 (1982) 1269-1290. Hulmston, P. and Hutton, R.C., Analytical capabilities of electrothermal vaporisation-inductively coupled plasma-mass spectrometry. Spectroscopy, 6( 1) (1991) 35. Hurst, G.S. and Payne, M.G., Principles and Applications of Resonance Ionisation Spectroscopy. Adam Higler, Briston, 1988. Hursthouse, A.S., Baxter, M.S., MxKay, K. and Livens, F.R., Evaluation of methods for the assay of Neptunium and other long-lived actinides in environmental samples. J. Radioanal. Nucl. Chem., Art., 157(2) (1992) 281. IAEA, Tomsk-7 caused minimal radiation hazards. Nuclear Waste News, 13 (1993) 157. IAEA, International Atomic Energy Agency, IAEA-TECDOC-619. International Atomic Energy Agency, Vienna (1991). IAEA Technical Reports series no. 295, Measurement of Radionuclides in Food and the Environment--A Guidebook. International Atomic Energy Agency, Vienna (1989). IAEA, Low level tritium measurement, IAEA-TECDOC-246, International Atomic Energy Agency, Vienna ( 1981). IAEA, Reference Methods for Marine Radioactivity Studies, Technical Reports series No. 118, International Atomic Energy Agency, Vienna (1970). ICRP 65 ICRP Publication 65. Protection against Radon-222 at home and at work, 1994. ICRP 66 ICRP publication 66. Annals of the ICRP 24(1/3), 1994. Igarashi, Y., Kawamura, H., Shiraishi, K. and Takaku, Y., Determination of Thorium and Uranium in biological samples inductively coupled plasma mass spectrometry using internal standardisation. J. Anal. At. Spectrom., 4 (1989) 571. International Electrotechnical Commission, Geneva, Switzerland, IEC Publication No. 697, 1981. Irlweck, K. and Veselsky, J.C., The electrodeposition of plutonium at tracer concentrations from a mixed NH~C1/NH4I system. Int. J. Appl. Radiat. Isot., 26 (1975) 481-483. James, M.R., Wayman, T., Stellmacher, K. and Hall, R.J.B., The determination of plutonium in urine by electro-thermal vaporisation inductively coupled mass spectrometry (ETV-ICP-MS), presented at 3rd Surrey Conference on Plasma Source Mass Spectrometry, Manchester, England, July 1989. Janghorbani, M., Davis, T.A. and Ting, B.T.G., Measurement of stable isotopes of bromine in biological fluids with ICP-MS. Analyst, 113 (1988) 405-411. Jarvis, K.E., Gray, A.L. and Houk, R.S., Handbook of Inductively Coupled Plasma Mass Spectrometry. Chapman and Hall, New York, 1992, p. 36. Jaworowski, Z. and Kownacka, L., Lead and radium in the lower stratosphere. Nature, 263 (1976) 303-304. Jiang, F.S., Lee, S.C., Bakhtiar, S.N. and Kuroda, P.K., Determination of thorium, uranium and plutonium isotopes in atmospheric samples. J. Radioanal. Nucl. Chem., 100 (1986) 65-72. Johansson, J.A.E. and Campbell, J.L., PIXE: A Novel Technique for Elemental Analysis. Wiley, 1988. Johns, F.B., Hahn, P.R., Thome, D.J. and Bretthauer, E.W. (eds.), Radiochemical analytical procedures for analysis of environmental samples, Rep. EMSL-0539-17. Environmental Monitoring and Support Lab., Las Vegas, NV (1979). Johns, F.B., Handbook of Radiochemical Analytical Methods, Rep. EPA-680/4/75-001. National Environmental Research Center, Las Vegas, NV (1975). Jones, K.W. and Gordon, B.M., Trace element determinations with synchrotron induced x-ray emission. Anal. Chem., 61 (1989) 341A. Kanish, G., Ruhle, H. and Schelenz, R. (eds.), Messan leitungen ftir die Uberwachung der Radioaktivit~it in der Umwelt, Bundesministerium des Innern, Karlsruhe (1984). Karol, I.L., Calculation of planetary distributions of radon and its long-lived daughter concentrations in the troposphere and lower stratosphere. Tellus, 22 (1970) 219-227. Kathren, R.L., Radioactivity in the EnvironmentwSources, Distribution, and Surveillance. Harvard Academic Publishers, 1984. Katz, J.J., Seaborg, G.T. and Morss, L.R., The Chemistry of the Actinide Elements, 2nd edn., Chapman
Measurements of Radioactivi O,
253
and Hall, New York, 1986. Kawada, Y., Nucl. Instr. Meth., 98 (1972) 21. Keder, W.E., Extraction of tetra- and hexavalent actinides from hydrochloric acid by tri-n-octylamine. J. Inorg. Nucl. Chem., 24 (1962) 561. Keder, W.E., Sheppard, J.C. and Wilson, A.S., The extraction of actinide elements from nitric acid solutions by tri-n-octylamine. J. Inorg. Nucl. Chem., 12 (1960) 327. Kilins, L.R., Litherland, A.E., Rucklidge, J.C. and Baba, N., Accelerator mass-spectrometric measurements of heavy long-lived isotopes. Appl. Radiat. Isot., 43 (1992) 279-287. Kim, C.-K., Oura, Y., Takaku, Y., Nitta, H., Igarashi, Y. and Ikeda, N., Measurement of 24~ ratio by fission track method and inductively coupled plasma mass spectrometry. J. Radioanal. Nucl. Chem. Lett., 136(5) (1989a) 353. Kim, C.-K., Seki, R., Morita, S., Yamasaki, S., Tsumura, A., Takaku, Y., Igarashi, Y. and Yamamoti, M., Application of a high resolution inductively coupled plasma mass spectrometer to the measurement of long lived radionuclides. J. Anal. Atomic Spectrom., 6 (1991) 205-209. Kim, C.-K., Takaku, A., Yamamoto, M., Kawamura, H., Hiraishi, K., Igarashi, Y., Igarashi, H., Takayama, H. and Ikeda, N., Determination of 237Npin Environmental samples using inductively coupled plasma mass spectrometry. J. Radioanal. Nucl. Chem. Art., 132(1) (1989b) 131. Kim, J., Stumpe, R. and Klenze, R., Laser induced photoacoustic spectroscopy for the speciation of transuranic elements in natural aquatic systems. Topics Curr. Chem., 157 (1990) 131-179. Kinney, J.B. and Staley, R.H., Annu. Rev. Mater. Sci., 12 (1982) 295. Kirkbright, G.F. and Castleden, S.L., Chem. Ber., 16 (1980) 661. Kleinberg, J. and Smith, H.L. (eds.), Collected radiochemical procedures, Rep. LA-1721, 4th edn., Los Alamos national Lab., Los Alamos, NM (1982). Knoll, G.F., Radiation Detection and Measurement. Wiley, New York, 1979, 816 pp. Knoth, J. and Schwenke, H., An x-ray fluorescence spectrometer with totally reflecting sample support for trace analysis at the ppb level. Fres. Z. Anal. Chem., 291 (1978) 200-204. Konno, T. and Suguro, H., A method for the determination of tritium in environmental water samples by electrolytic enrichment. Radioisotopes 35 1 (1986) 1-8. Koppenaal, D.W., Atomic mass spectrometry. Anal. Chem., 64 (1992) 320R. Korkisch, J., Handbook of Ion Exchange Resins, Their Application to Inorganic Analytical Chemistry, Vols. 1-6. CRC Press, Boca Raton, FL, 1989. Kotrappa, P., Dempsey, J.C., Hickley, J.R. and Steiff, L.R., Health Phys., 54 (1988) 47-56. Kownacka, L., Jaworowski, Z., Baranski, A., Suplinska, M., Lewitowicz, J., Wesolek, R. and Romicki, Z., Measurements of concentration of 9~ '~ '258b, '37Cs, "~Ce, 21~ Z26Ra, U and stable Pb in the tropospheric and lower stratospheric air from 1973 to 1984, Central Laboratory for Radiological Protection, Warsaw, Report No. CLOR- 117/D (1985). Kownacka, L., Jaworowski, Z., Suplinska, M., Vertical distribution and flows of lead and natural radionuclides in the atmosphere. Sci. Total Environ., 91 (1990) 199-221. Kownacka, L., Vertical distribution of fission products, 226Raand Pb in the atmosphere, Institute of Nuclear Research, Warsaw, Ph.D. thesis, Report INR- 1895/IA/D/A (1980). Kreuzer, L.B., J. Appl. Phys., 42 (1971) 2934. Krieger, H.L. and Whittaker, E.L., Prescribed procedures for measurement of radioactivity in drinking water, Rep. EPA-600/4032, Environmental Monitoring and Support Lab., Cincinnati, OH (1980). Krieger, H.L. and Whittaker, E.L., Prescribed procedures for measurement of radioactivity in drinking water, Rep. EPA-600/4-80-032, Environmental Monitoring and Support lab., Cincinnati, OH (1980b). Lain, J.W. and McLaren, J.W., Use of aerosol processing and nitrogen-argon plasmas for reduction of oxide interference in inductively coupled plasma mass spectrometry. J. Anal. At. Spectrom., 5 (1990) 419. Larsen, R. and Juzdan, Z., Radioactivity at Barrow and Mauna Loa following the Chernobyl accident, in: Geophysical Monitoring for Climatic Change, No. 14, Summary Report 1985, pp. 130-133, December (1986). Larsen, R.J. and Sanderson, C.G., EML surface air sampling program, 1989 data, USDOE Report-541, August (1991) Available from NTIS, Springfield, VA.
254
Chapter 5
Larsen, R.J., Haagenson, P. and Reiss, N., Transport process associated with the initial elevated concentrations of Chernobyl radioactivity in surface air in the U.S.J. Environ. Radioactivity, 10 (1989) 1-18. Larsen, R.J., Letter to Editor: Global decrease in Beryllium-7 in surface air. J. Environ. Radioactivity, 18 (1993) 85-87. Larsen, R.J., Sanderson, C.G. and Kada, J., EML surface air sampling program, 1990-1993 data, Report EML-572 (1995) pp. 4-247. Larsen, R.J., Sanderson, C.G., Lee, H.N., Decker, K.M. and Beck, H.L., Letter to Editor: Fission products detected in Alaska following the Tomsk-7 accident. J. Environ. Radioactivity, 23 (1994) 205-209. Larsen, R.J., Sanderson, C.G., Rivera, W. and Zamichieli, M., The characterisation of radionuclides in North America and Hawaiian surface air and deposition following the Chernobyl accident, USDOE Report EML-460, pp. 1-104, October (1986). Larsen, R.J., Submicrometer aerosol collection characteristics of Dynaweb DW7301L and Freudenberg FA2311 filter media, Abstract Book, American Association for Aerosol Research, 1990 Annual Meeting, pp. 266, June 18-22 (1990). Lederer, C.M. and Shirley, V.S., Table of Isotopes, 7th edn. Wiley, New York, 1978. Lee, H.N. and Feichter, J., An intercomparison of wet precipitation scavenging schemes and the emission rates of 222Rn for the simulation of global transport and deposition of -'~~ J. Geophys. Res., 100 (1995) 253. Lee, H.N., Feichter, J. and Rehfeld, S., Simulation of global transport and deposition of 222Rn and 2'~ Proceedings of Topical Meeting on Environmental Transport and Dosimetry of the American Nuclear Society, Charleston, S.C., September 1-3 (1993a). Lee, H.N., Larsen, R.J. and Sanderson, C.G., Tomsk-7 debris at BRW: Detection and transport, in: Climate Monitoring and Diagnostic Laboratory, No. 21, Summary Report 1992, pp. 104-105, December (1993b). Lepp~ineu, A., Ground-level air sampling in Finnish centre for radiation and nuclear safety, Proc. Second Int. Meeting on Low-Level air Radioactivity Monitoring, Madralin, near Warsaw, Poland, 14-18 Feb. 1994. Lethokov, V.S., Laser Photoionization Spectroscopy. Academic Press, London (1987). Lieberman, R., (Ed.), Radiochemistry procedures manual, Rep. EPA 520/5/-84-006, Eastern Environmental Radiation Facility, Montgomery, AL (1984). Lieberman, R., Moghissi, A.A., Low-level counting and liquid scintillationmlI Applications of emulsions in tritium counting. Int. J. Appl. Radiat. Isot., 21 (1970) 319-327. Litherland, A.E., Allen, K.W. and Hall, E.T., (Eds.), Proc. of the Royal Society (1987). Litherland, A.E., Fundamentals of accelerator mass spectrometry. Phil. Trans. R. Soc. Lond., A323 (1987) 5-21. Livens, F.R. and Singleton, D.L., Evaluation of methods for the radiometric measurement of Americium241 in environmental samples. Analyst, 114 (1989) 1097. Livingston, H.D., Mann, D.R. and Bowen, V.T., Analytical procedures for transuranic elements in seawater and marine sediments. In: Analytical Methods in Oceanography, Advances in Chemistry Series 147. American Chemical Society, 1975. Lucas, H.F., Rev. Sci. Instr., 28 (1957) 258-259. Lyamshev, L.M. and Sedov, L.N., Sov. Phys. Acoust., 27 (1981) 4. Machta, L., Some aspects of the USA fallout programme. Meteorological aspects of atmospheric radioactivity, World Meteorological Organisation, Geneva, Tech. Note No. 68, pp. 155-177 (1965). Martell, E.A. and Moore, H.E., Tropospheric residence times: A critical review. J. Rech. Atmos., 8 (1974) 903-910. McGee, E.J., Colgan, P.A., Keatinge, M., Herrill, A.D., Kennedy, V.H., Johanson, K.J., Aarkrog, A. and Nielsen, S.P., Bias and measurement errors in radioactivity data from four European Radiation Research Laboratories. Analyst, 117 (1992) 941. McKlveen, J.W. and McDowell, W.J., Nucl. Instr. Meth. Phys. Res., 223 (1984) 372. McMahon, A.W., An intercomparison of non-radiometric methods for the measurements of low levels of radionuclides. Appl. Radiat. Isot., 43 (1992) 289-303.
Measurements of Radioactivit3,
255
McMahon, W., A critical intercomparison of techniques for the determination of low levels of long lived radionuclides. AERE R 13617, Harwell Laboratory, Oxfordshire, U.K. (1989). Millies-Labroix, J.-C., Bourlat, Y., Mosnier, R., Airborne radioactivity measurements in Montlhery (France), in: Proc. Second Int. Meeting on Low-Level Air Radioactivity Monitoring, Madralin, near Warsaw, Poland, 14-18 Feb. 1994. Momoshima, N., Nagasato, Y. and Takashima, Y., A sensitive method for the determination of tritium in urine. J. Radioanal. Nucl. Chem. Lett., 107 (1986) 353-359. Monitoring and Assessment Research Center, United Kingdom United Nations Environmental Programme, Environmental data report. Basil Blackwell Ltd. (1987). Monz, L., Hohmann, R., Kluge, H.-J., Kunze, S., Lantzsch, J., Otteu, E.W., Passler, G., Senne, P., Stenner, J., Stratmann, K., Wendt, K. and Zimmer, K., Fast, low-level detection of strontium-90 and strontium-89 in environmental samples by collinear resonance ionisation spectroscopy. Spectrochim. Acta, 48B (1993) 1655-1671. Moore, H.E., Poet, S.E. and Martell, E.A., Size distribution and origin of lead-210, bismuth-210 and polonium-210 on airborne particles in the troposphere. In: T.F. Gesell and W.M. Lowder (Eds.), Natural Radiation Environment III. Technical Information Center, U.S. Dept. of Energy, Springfield, Virginia, pp. 415-429 (1980); Pierson, D.H., Cambray, R.S. and Spicer, G.S., Lead-210 and polonium-210 in the atmosphere, Tellus, 18 (1966) 427-433. Morita, S., Kim., C.-K., Takaku, Y., Seki, R. and Ikeda, N., Determination of Technetium-99 in environmental samples by inductively coupled plasma mass spectrometry. Appl. Radiat. Isot., 42(6) (1991) 531. Morita, S., Tobita, K. and Kurabayashi, M., Determination of Technetium-99 in environmental samples by inductively coupled plasma mass spectrometry. Radiochim. Acta 63 (1993) 63-67. Mouchel, D. and Wordel, R., Measurement of low-level radioactivity in environmental samples by gamma-ray spectrometry. Appl. Radiat. Isot., 43 (1992) 49-59. Muller, R.A., Radioisotope dating with a cyclotron. Science, 196 (1977) 489. Munro, S., Ebdon, L. and McWeeny, D.J., Applications of ICP-MS for trace determinations in foods. J. Anal. Atom. Spectrom. 1 (1986) 211-219. National Council of Radiation Protection: NCRP-Report No. 58, A Handbook of Radioactivity Measurements Procedures, 2nd edn., (1985). Neame, K.D. and Howewood, C.A., Introduction to Liquid Scintillation Counting. Butterworth, London, 1974, pp. 67-71. Nicholson, S., Sanders, T.W. and Blaine, L.M., The determination of low levels of Technetium-99 in environmental samples by inductively coupled plasma-mass spectrometry. Presented at 3rd Surrey Conference on Plasma Source Mass Spectrometry, Manchester, England (July, 1989). Northcliff, L.C. and Shilling, R.F., Nucl. Data Tables, A7 (1970) 257. Norton, E.F., Chemical yield determination in radiochemistry, Nuclear Science Series, N.A.S.-NS 3111 (1967). Olesik, J.W., Elemental analysis using ICP-OES and ICP/MS--An evaluation and assessment of remaining problems. Anal. Chem., 63 (1991) 12A. Olsson, O., Holm, E. and Botter-Jensen, L., Appl. Radiat. Isot., 43 (1992) 77-82. Owen, L.M.W., Crews, H.M., Hutton, R.S. and Walsh, A., Preliminary study of metals in proteins by HPLC-ICP-MS using multi-element time-resolved analysis. Analyst, 117 (1992) 649-655. Owen, L.M.W., Crews, H.M. and Massey, R.C., Aluminium in tea: SEC-ICP-MS speciation studies of infusions and simulated gastro-intestinal digests. Chem. Spec. Bioavail., 4(3) (1992) 89-96. Pao, H.Y., Optoacoustic Spectroscopy and Detection. Academic, New York, 1977. Park, T.S. and Jeon, W.J., Measurement of radioactive samples in Marinelli beakers by gamma-ray spectrometry. J. Radioanal. Nucl. Chem., 193 (1995) 133-144. Patel, C.K.N. and Tam, A.C., Rev. Mod. Phys., 53 (1981) 517. Perkins, R.W. and Thomas, C.W., Worldwide Fallout. In: Transuranic Elements in the Environment (W.C. Hanson, ed.) DOE/TIC-22800, U.S. Department of Energy, NTIS, Springfield, VA, 1980. Pin, C., Lacombe, S., Telouk, P. and Imbert, J.-L., Isotope dilution inductively coupled plasma mass spectrometry: a straightforward method for rapid and accurate determination of uranium and thorium
256
Chapter 5
in silicate rocks. Anal. Chim. Acta, 256 (1992) 153. Pollard, P.M., McMillan, J.W., McCullough, T., Bowen, J.L., Buckley, L.C. and Arthur, J., An atomic absorption spectrophotometry system for the analysis of radioactive waste leachates. Harwell Report AERE R 11499 (1986). Porstendorfer, J. and Mercer, T.T., Health Phys., 15 (1979) 191-199. Price, W.J., Nuclear Radiation and Detection. McGraw-Hill Book, Co., New York, 1964. Pritchard, H.M. and Marien, K.A., Health Phys., 48 (1985) 797-803. Prospero, J.M., Savoie, D.L., Saltzman, E.S. and Larsen, R.J., Impact of oceanic sources of biogenic sulphur on sulphate aerosol concentrations at Mawson, Antarctica. Nature, 350 (1991) 221-223. Purser, K.H. et al., An attempt to detect stable N- ions from a sputter ion source and some implication of the results for the design of tandems for sensitive carbon analysis. Rev. Phys. Appl., 12 (1977) 1487. Rama and Honda, M., Natural radioactivity in the atmosphere. J. Geophys. Res., 66 (1961) 3227-3231. Rangarajan, C., Gopalakrishnan, S., Chandrasekaran, V.R. and Eapen, C.D., The relative concentrations of radon daughter products in surface air and the significance of their ratios. J. Geophys. Res., 80 (1975) 845-848. Rangarajan, C., Gopalakrishnan, S.S. and Eapen, C.D., Global variation of lead-210 in surface air and precipitation. Health and Safety Laboratory, New York, Report No. HASL-298, pp. I63-I82 (1976). Regan, J.G.T. and Tyler, J.F.C., The determination of Sr-90 and Sr-89 in water without separation of strontium from calcium. Analyst, 101 (1976) 32-38. Rehfeld, S. and Heimann, M., Three-dimensional atmospheric transport simulation of radioactive tracers -"~ 7Be, '~ and 9~ Max-Planck-lnstitut ftir Meteorologie, Report No. 144 (1994). Riglet, C., Provitina, O., Dautheribes, J.-L. and Revy, D., Determination of traces of Neptunium-237 in enriched Uranium solutions using inductively coupled plasma mass spectrometry. J. Anal. At. Spectrom., 7 (1992) 923. Robertson, D.E., Smith, M.R., Doppenaal, D.W., Kiddy, A.R., Strebin, R.S., Concentrations and behaviour of '291,99Tc, and '~C in low-level radioactive wastes from commercial nuclear power stations. Waste Manage., 2 (1991 ) 287. Rosencwaig, A., Adv. Electron. Electron Phys., 46 (1978) 207. Rosencwaig, A., Photoacoustics and photoacoustic spectroscopy. Chem. Anal., Vol. 57, New York, 1980. Ross, R.R., Nayce, J.R., Lardy, M.M., Inductively coupled plasma-mass spectrometry: An emerging method for analysis of long-lived radionuclides. Radioactivity Radiochem., 4 (1993) 24-37. Sakanoue, M., Transuranium nuclides in the environment. Radiochim. Acta, 42 (1987) 103. Samuellson, C., in: P.K. Hopke (ed.), ACS Symp. Series 331, Chapter 16. American Chemical Society, Washington, DC, 1987. Sanderson, C.G., An evaluation of commercial IBM PC software for the analysis of low level environmental gamma-ray spectra. Environ. Int., 14 (1989) 379-384. Sanderson, C.G., Latner, N. and Larsen, R.J., Environmental gamma-ray spectroscopy at remote sites with satellite data transmission. Nucl. Instr. Meth. Phys. Res. Sect. A, (1994) 271-277. Savoie, D.L., Prospero, J.M., Larsen, R.J., Huang, F., Izaguirre, M.A., Huang, T., Snowden, T.H., Custals, L. and Sanderson, C.G., Nitrogen and sulphur species in Antarctic aerosols at Mawson, Palmer Station, March (King George Island). J. Atmos. Chem., 17 (1993) 95-122. Savoie, D.L., Prospero, J.M., Saltzman, E.S. and Larsen, R.J., Nitrogen and sulphur species in aerosols at Mawson, Antarctica and their relationship to natural radionuclides. J. Atmos. Chem., 14 (1992) 181-204. Schmit, J.-P., Youla, M. and GElinas, Y., Multi-element analysis of biological tissues by inductively coupled plasma mass spectrometry. Anal. Chim. Acta, 249 (1991) 495. Scott, T.G. and Reynolds, S.A., Determination of plutonium environmental samples--II: Procedures. Radiochem. Radioanal. Lett., 23 (1975) 275-281. Sekine, K., Imai, T. and Kasai, A., Liquid-liquid extraction and sequential determination of plutonium and americium in environmental samples by alpha spectrometry. Talanta, 34 (1987) 567-570. Shaw, T.J. and Francois, R., A fast and sensitive ICP-MS assay for the determination of 23~ in marine sediments. Geochim. Cosmochim. Acta, 55 (1991) 2075. Shiraishi, K., Takaku, Y., Yoshimizu, K., Igarashi, Y., Masuda, K., Mclnroy, J.F. and Tanaka, G.,
Measurements of Radioactivi~
257
Determination of Thorium and Uranium in total diet samples by inductively coupled plasma mass spectrometry. J. Anal. At. Spectrum., 6 (1991) 335. Shuzhong, Z., Shaoxian, L. and Zhenhong, X, Rapid Determination of 9~ by TBP Extraction. IAEA CRM, Vienna, 1991. Sill, C.W. and Olson, D.G., Anal. Chem., 42 (1970) 1596. Smith, F.G., Wiederin, D.R. and Houk, R.S., Argon-Xenon plasma for alleviating polyatomic ion interferences in inductively coupled plasma mass spectrometry. Anal. Chem, 63 (1991) 1458. Smith, M.R., Wyse, E.J. and Koppenaal, D.W., Radionuclide detection by inductively coupled plasma mass spectrometry: a comparison of atomic and radiation detection methods. J. Radioanal. Nucl. Chem., 160 (1992) 341. Somoano, R.B., Angew Chem. Int., 17 (1978) 238. Spitz, H. and Wrenn, M.E., HASL-325 (1977) 48-64. Steinkopff, Th., Fay, B., Blaab, H., Jacobsen, I. and Kiesewetter, W., The surveillance of radioactivity in the atmosphere by the Deutscher Wetterdienst, in: Proc. Second Int. Meeting on Low-Level Air Radioactivity Monitoring, Madralin, near Warsaw, Poland, 14-18 Feb. 1994. Stevenson, P.C. and Nervik, W.E., The radiochemistry of the rare earths, Scandium, Yttrium and Actinium, US Report NAS.NS 3020 (1961). Strachnov, V., Valkovic, V., LaRosa, J., Dekner, R. and Zeisler, R., IAEA Programme of natural matrix reference materials for the determination of radionuclides. Fresenius J. Anal. Chem., 345 (1993) 169-173. Sugihara, T.T., Low-level radiochemical separations. Nucl. Sci. Ser., NAS-NS 3103 (1961). Sun, X.F., Ting, B.T.G., Zeisel, S.H. and Janghorbani, M., Accurate measurement of stable isotopes of lithium by ICP-MS. Analyst, 112 (1987) 1223-1228. Szabo, A.S., Method for calculation of 9~ activity in environmental samples. Radiochem. Radioanal. Lett., 54/5 (1982) 301-308. Tam, A.C., In: Kliger, D.S. (ed.) Ultrasensitive Laser Spectroscopy. Academic, New York, 1983. Tam, A.C., Rev. Mod. Phys., 58 (1986) 381. Tanaka, E., Nucl. Instr. Meth., 13 (1961) 43. Thomas, C.W., A rapid procedure for the measurement of the transuranic elements and thorium. IAEA CRM, Vienna. 1991. Thompson, J.J. and Houk, R.S., ICP-MS detection for multi-element flow injection analysis and elemental speciation by reversed-phase liquid chromatography. Anal. Chem. 58 (1986) 2541-2548. Ting, B.T.G. and Janghorbani, M., ICP-MS applied to isotope analysis of iron in human faecal matter. Anal. Chem. 58 (1985) 1334-1340. Ting, B.T.G., Mooers, C.S. and Janghorbani, M., Isotopic determination of selenium in biological materials with ICP-MS. Analyst, 114 (1989) 667-674. Tye, C. and Mennie, D., Actinide analysis with a novel high sensitivity ICP-MS. International Laboratory News, December 1994, p. 8. UNSCEAR, Ionising radiation: Sources and biological effects. United Nations Scientific Committee on the Effects of Atomic Radiation (1993). Vajda, N., Ghods-Esphahani, A. and Danesi, P.R., Determination of radiostrontium in soil samples using a crown ether. IAEA CRM, Vienna, 1991. Vapirev, F.I. and Hristova, A.V., Rapid correlation assessment of Sr in fallout by gamma-spectroscopy. IAEA CRM, Vienna, 1991. Volchok, H. and de Planque, G. (eds.), EML Procedures Manual, Rep. HASL-300, 26th edn, Environmental Measurements Lab., New York (1983). Watt, D.E. and Ramsden, D., High Sensitivity Counting Techniques. Pergamon Press, New York, 1964, 105. Wellisch, E.M., Phil. Mag., 6 (xxvi) (1913) 623-635. Wershofen H. and Arnold, D., Radionuclides in ground-level air in Braunschweig--a report of the PTB trace survey station for 1993 and 1994. PTB-6.33-95-3, Braunschweig, Germany, December 1995. Wershofen, H. and Arnold, D., Die Aktivit~itskonzentrationen 7-strahlender Radionuklide im Staub der bodennahen Luft von Braunschweig und Berlin 1992. Report PRB-6.301-93-1, Physikalisch-
258
Chapter 5
Technische Bundesanstalt, Bundesallee 100, D-38116 Braunschweig. Wessman, R.A., Leventhal, L., Historical review of transuranic analytical methodology, NVO-1978. West, G.A., Barrett, J.J., Siebert, D.R. and Reddy, K.V., Rev. Sci. Instrum., 54 (1983) 797. Wiederin, D.R., Smith, F.G. and Houk, R.S., Direct injection nebulization for inductively coupled plasma mass spectrometry. Anal. Chem., 63 (1991) 219. W61fli, W., Polach, H.A. and Andersen, H.H. (eds.), Proceedings of the III International Symposium on Accelerator Mass Spectrometry, Ztirich, Switzerland (1984), in Nucl. Instrum. Methods Phys. Res. B, 5 (1984) 91. Wong, K.M., Brown, G.S. and Noshkin, V.E., A rapid procedure for plutonium separation in large volumes of fresh and saline water by manganese dioxide coprecipitation. J. Radioanal. Chem., 42 (1975) 7-15. Wong, K.M., Radiochemical determination of plutonium in sea water, sediments and marine organisms. Anal. Chim. Acta, 56 (1971) 355-364. World Health Organisation, Methods of Radiochemical Analysis. Geneva, 1966. Wrenn, M.E., Spitz, H. and Cohen, N., IEEE Trans. Nucl. Sci. NS-22, (1975) 645-648. X-ray and gamma-ray standards for detector calibration, IAEA-TECDOC-619, 1991. Yamamoto, M., Chatani, K., Komura, K. and Ueno, K., Development of alpha-ray spectrometric techniques for the measurement of low-level ~37Npin environmental soil and sediment. Radiochim. Acta, 47 (1989) 63. Yamato, A., J. Radioanal. Chem., 75 (1982) 265. Zharov, V.P., In: Letkhov, V.S., (ed.) Laser Analytical Spectrochemistry. Adam Hilger, Bristol, 1985. Ziegler, J.F., Helium Stopping Powers and Ranges in all Elements. Pergamon Press, New York, 1977.
259
CHAPTER 6
Radiation Safety
6.1 INTRODUCTION It has been recognised since early studies on X-rays and radioactive minerals that exposure to high levels of radiation can cause clinical damage to the tissues of the human body. In addition, long-term epidemiological studies of populations exposed to radiation, especially the survivors of the atomic bombing of Hiroshima and Nagasaki in Japan 1945, have demonstrated that exposure to radiation also has a potential for the delayed induction of malignancies. It is, therefore, essential that activities involving radiation exposure, such as the production and use of radiation sources and radioactive materials, and the operation of nuclear installations and management of the radioactive waste they produce, be subject to certain standards of safety in order to protect persons exposed to radiation. Ionising radiation and radioactive substances are natural and permanent features of the environment, and thus the risks associated with radiation in all its forms can only be restricted, not eliminated entirely. Additionally, the use of man-made radiation is widespread. Sources of ionising radiation are essential to modern health care: disposable medical supplies sterilised by intense radiation have been central to combating disease; radiology is a vital diagnostic tool; and radiotherapy is commonly part of the treatment of malignancies. The use of nuclear energy and applications of its byproducts, i.e. ionising radiation and radioactive substances, continue to increase around the world. Nuclear techniques are in growing use for industry, agriculture, medicine and many fields of research, benefiting hundreds of millions of people and giving employment to millions of people in the related occupations. Irradiation is used around the world to preserve foodstuffs and reduce wastage, and sterilisation techniques have been used to eradicate disease-carrying insects and pests. Industrial radiography is in routine use, for example to examine welds and detect cracks and help prevent the failure of engineered structures. The acceptance by society of risks associated with radiation is conditional on the benefits to be gained from the use made of radiation. Nonetheless, the risks must be restricted and protected against by the application of radiation safety standards.
260
Chapter6
Exposure to radiation can cause detrimental health effects. At large doses, radiation effects such as nausea, reddening of the skin or, in severe cases, more acute syndromes are clinically expressed in exposed individuals within a short period of time after the exposure; such effects are called "deterministic" because they are certain to occur if the dose exceeds a threshold level. Radiation exposure can also induce effects such as malignancies, which are expressed after a latency period and may be epidemiologically detectable in a population; this induction is assumed to take place over the entire range of doses without a threshold level. Hereditary effects due to radiation exposure have been statistically detected in other mammalian populations and are presumed to occur in human populations also. These epidemiologically detectable effects malignancies and hereditary e f f e c t s ~ a r e termed "stochastic" effects because of their random nature. Deterministic effects due to radiation exposure are the result of different processes, mainly cell killing and delayed cell division, which can, if extensive enough, impair the function of the exposed tissue. The severity of a particular deterministic effect in an exposed individual increases with the dose above the threshold for the occurrence of the effect. Stochastic effects may occur if an irradiated cell is modified rather than killed. Modified cells may, after a prolonged delay, develop into a cancer. The body's repair and defence mechanisms make this a very improbable outcome at small doses; nevertheless, there is no evidence of a threshold dose below which cancer cannot result. The probability of occurrence of cancer is higher for higher doses, but the severity of any cancer that may result from irradiation is independent of the dose. Also, if a germ cell whose function is to transmit genetic information to progeny is damaged owing to radiation exposure, it is conceivable that hereditary effects of various types may develop in the descendants of the exposed person. The likelihood of stochastic effects is presumed to be proportional to the dose r e c e i v e d ~ a l s o without a dose threshold. In addition to the aforementioned health effects, other health effects may occur in infants due to exposure of the embryo or foetus to radiation. These effects include a greater likelihood of leukaemia and, for exposure above various threshold dose values during certain periods of pregnancy, severe mental retardation and congenital malformations. Measures for protection against exposure to an external radiation source are as follows: 1. Keep a distance from a radiation source. 2. Place a shield between a radiation source and a worker. 3. Make the exposure time short. In many cases, the use of only one of the above measures would be insufficient for protection against exposure and so all three measures are combined to achieve effective protection. If an adequate shielding material is not available, the distance and time are so set as to make up for it. There are two reasons for keeping a distance from a radiation source. One reason is the geometrical one. For this reason, the exposure rate decreases with distance from a radiation source. The rate of reduction differs depending on the geometric layout of the
Radiation Safet3,
261
source and surrounding materials. However, when the radiation source can be taken as a point, the exposure rate reduces in inverse proportion to the square of a distance from the source (the inverse square law). Another reason is the absorption of radiation by air. This can be expected with alpha-rays and beta-rays, but cannot be expected with X-rays and gamma-rays, whose capability of penetration is large.
(i) Alpha-rays Alpha-rays are completely absorbed by a several centimetres thick air layer. However, many alpha-radionuclides also emit gamma-rays; therefore, care must be taken against external exposure to such gamma-ray.
(ii) Beta-rays Beta-rays have larger capability in penetrating power than alpha-rays. The range in the air depends on the energy of the beta-rays. The range of ~4C and 35S, whose maximum energies are 0.16 MeV or so, is about 30 cm. The ranges of 32p (1.71 MeV) and 90y (2.28 MeV) are about 6 m and 8 m, respectively. The absorbed dose rate D (rad/h) imparted by beta particles at a certain distance from a part radiation source which emits beta-rays of a relatively high energy exceeding 0.8 MeV such as 32p and 90y is represented as a rule-of-thumb by the expression given below: O =
3•
A d ~-
(6.1)
where A is the radioactivity (Ci) of a source and d is a distance (cm) from a radioactive material. The expression does not contain the term of energy of a beta particle. This is because the dose rate hardly changes with energy in the range of 0.8-3 MeV as shown in Fig. 6.1. Figure 6.2 shows, for several beta-ray emitting radionuclides, the relationship between the absorbed dose rate per unit radioactivity and a distance from a radiation source.
(iii) Gamma-rays Gamma-rays follow the inverse square law over a wide range of exposure rate and distance. Suppose that a point gamma ray source is uniformly emitting Sphotons of energy E (MeV) per second in every direction. The photon fluence rate 9 (photons/ cm2/s) at a distance of d cm from the source is represented by S =~ 4/l;d 2
(6.2)
If a gamma-ray of 1 roentgen (R) transfers energy of about 87 ergs or 5.4x 107 MeV to 1 gram of air (870 mrad = 8.7 mGy), the exposure rate X (R/h) at a point 1 cm off the source is expressed by
Chapter 6
262
IN VACUUM 1000~ 8 7
~
5
--
4
-
I
IN AIR I
'I" I
"1
I
....
I
"
I
I .....
I
I
I
''
'i'"
!
I
I
_
i i
v
i
~,
10 cm
1000 9 8 7 6 5 4
2
loo
9
8 7 6 5 4
lm ._____--------
/
/ /: t
/[ IN VACUUM
'
!
' ~
AIR
BETA ENERGY, E~.~ / MeV
Fig. 6.1. D o s e rate f r o m a 1 Ci p o i n t s o u r c e of b e t a radiation.
RadiationSafeO,
263 100 50
(2mCi) t.Z
10
5
r~
9
0.5
90Sr +90y
(2mCi) 54,2.27)
35S / (0.167)
0
r~
<
0.1 147pm
(0.23)
0.05
0.01 t~_t...._t..a__~_x~ 2 5 10
20
50
100
200
500
DISTANCE FROM A BETA - RAY SOURCE / cm Fig. 6.2. Distance from a point beta-ray source vs. beta-ray absorbed dose rate in air. [Dose rate per mCi. However, for 9~ + 90y, the dose rate from a 2 mCi source (1 mCi 9~ and 1 mCi 90y)]. The parenthesized values are maximum energies of beta-ray in MeV.
= 4.10 x 10-3SE(I a en )air
(6.3)
where (~en)air is the linear energy absorption coefficient of air. Now, suppose that n photons are emitted in a single disintegration. With the yield of photon i taken as Pi and energy as E i (MeV), the exposure rate X (R/h-dps) at a point 1 cm off a point gamma-ray source of 1 dps (Bq) is represented by 2 = 5.27 x 10 -6 ~PiEi
(~t en )air, i
(6.4)
i=1
where (~ten)air, i is the mass energy absorption coefficient of air for the ?-ray whose energy is E i MeV and P is the density of air (1.293x 10 -3 g/cm3). Exposure rate .~ at a point 1 m distant from a radiation source of 1 Ci is usually called exposure rate constant 1-"
264
Chapter 6
(R-m2/(h.Ci)). Exposure rate X(R/h) at a distance d (m) from a gamma-ray source of radioactivity A (Ci) can be approximated by using this exposure rate constant F as follows: = 1-'A d2
(6.5)
What is represented by eq. (6.5) is called the inverse square law. In handling a radioactive material, it is basically important not to hold a radiation source directly with the hand. Handle it by remote operation as much as possible. Use forceps, tongs or similar devices. In handling radioactive materials of curie level or higher, remote controlling devices such as a master-slave type manipulator must be used. Shielding is the most positive method of protection against exposure. This method reduces the radiation dose rate by placing a radiation-absorbing material between a radiation source and a worker. The kind and thickness of an absorbing material should be selected depending on the type, energy and intensity of a radiation involved. In handling an alpha radionuclide, shielding by keeping several centimetres away from a radiation source is not necessary because alpha particles are completely absorbed in the air layer. Even when the skin surface is contaminated with an alpha radionuclide, alpha particles can hardly penetrate to the sensitive layer, basal cell layer, or epidermis lying 50-100 ~tm apart beneath the skin surface. By wearing rubber gloves (usually about 0.25 mm thick), a hand can be completely protected against alpha radiation. The range of a beta particle is usually represented by a distance within aluminium. The simple way to estimate the range expressed in g/cm 2 is to take it as about half of the maximum beta-ray energy expressed in MeV. The range of a beta particle from 8Li, which has the maximum energy, 13 MeV, is about 7 g/cm 2. This range is equivalent to an aluminium (= 2.7 g/cm 3) thickness of about 2.6 cm. The ranges of commonly used beta-radionuclides 32p and 90y are about 0.75 g/cm 2 (equivalent to an aluminium thickness of about 0.28 cm) and 1.0 g/cm 2 (equivalent to an aluminium thickness of about 0.37 cm), respectively. Considering that the energy of beta particle emitted from 8Li is exceptionally high, this means that most of the beta rays can usually be shielded with an aluminium plate less than 1 cm thick. When beta rays pass through the material, an X-ray is produced by the so-called Bremsstrahlung process. Bremsstrahlung radiation increases as the atomic number of a shielding material or the energy of beta particles increases. Therefore, usually, not the aluminium plate but the plastic plate (- 1 g c m -3, transparent, 1.0-1.5 cm thick) is used for shielding beta rays. For beta radiation sources of high radioactivities exceeding 3 mCi (= 100 MBq), the X-ray produced in plastics by Bremsstrahlung must also be shielded. In such a case, high atomic number material such as lead is used in addition to the plastic plate. The relationship between the maximum energy and the maximum range of beta rays in various materials is shown in Fig. 6.3.
Radiation SafeO,
265 I
I
I
I
~
I
I I I
I
10 ~
10-'
9PER LEAD
1 0 .2
1 0 -3
0.2
0.4
1
2
4
MAXIMUM ENERGY OF BETA RAY / MeV Fig. 6 . 3 . Beta-ray shielding.
When the skin of the hand or other body part is contaminated with a beta radionuclide during work, the beta-ray reaches the tissue, basal cell layer, beneath the insensitive layer of epidermis, unlike in the case of an alpha radionuclide. When a beta radionuclide of a high energy is to be handled, it is desirable to wear protective goggles for protecting the eye lenses, which are located at 0.2-0.3 cm below the eye surface, from the radiation.
266
Chapter 6
When a positive electron-emitting radionuclide such as 22Na and 18F is to be handled, it is necessary to prepare a shield against two 0.51 MeV photons produced by annihilation of the positive electrons. Let us discuss in some detail the situation with a collimated beam of X-rays and gamma-rays. When photons pass through a shield of x cm thickness, the relationship between the intensity of the beam emerging from the shield, I (photons/cm2.s), and the intensity of the beam incident on the shield, I 0 (photons/cm2-s), is expressed as follows: I = loe -"x
(6.6)
where ~t is the linear attenuation coefficient (cm-~). As ax increases, the shielding effect increases. The value of ~t varies depending on the energy of the photon, and the atomic number and density of a shielding material. In addition to ~t, the half-value thickness is commonly used to represent the shielding effect. The half-value thickness, HVT (cm), is expressed with ~t is follows: HVT = In 2/~t = 0.693/~t
(6.7)
Expression (6.6) is rearranged with HVT as follows: I = (1/2)" I o
(6.8)
where n = x/HVT, and n indicates how many times the shield thickness is as large as the half-value thickness. By using a shield 10 times as thick as HVT, the gamma-ray intensity reduces to 1/1000 of the initial intensity. The tenth-value thickness, TVT, is also used in the calculation of shielding. As in eq. (6.7), TVT is defined as follows: TVT = In 10/~t = 2.303/~t.
(6.9)
TVT is 3.32 times HVT. With the shield thickness x being n' times TVT, expression (6.6) is rearranged as follows: I = (1/10)" I o.
(6.10)
This expression tells us that the gamma-ray intensity reduces in accordance with the decimal notation. For example, at twice and three times as large a shield thickness as TVT, the gamma-ray intensity reduces to 1/100 and 1/1000 of the incident intensity, respectively. This description above deals with the narrow beam of penetrating gamma-ray in which radiation from scattering can be neglected. In many actual cases, however, gamma-ray is of a broad beam. And in such cases, scattering must be taken into consideration and the following expression is used for calculation of shielding: I = l o B e -"~
(6.11)
Radiation SafeO,
267
where B is a correction factor taking into account an increase in intensity caused by the scattered radiations re-entering the emergent beams. This factor is known as build-up factor. The build-up factor is related to energy of gamma-ray, kind of shielding materials, thickness and shape of shield and a spread of flux of radiation. In practice, however, an equation expressing the relationship between the product of shield thickness and linear attenuation coefficient, and the build-up factor is used for calculating the build-up factor at a certain shield thickness. One of the typical approximations is as follows: B = 1 + ~x
(6.12)
Figure 6.4 shows the attenuation of gamma rays through steel, taking into account the build-up factor. For shielding against gamma-rays, materials of a high atomic number such as lead, steel and concrete are used as shielding materials. Concrete is classified into normal concrete (approx. 2.3 specific gravity) and heavy concrete (approx. 3.4 specific gravity). Normal concrete is one of the most popular building materials. Therefore, it is useful in the case where the building itself must perform the shielding function. When lead bricks or concrete blocks are used as shielding materials, shielding must be done as near radiation sources as possible and the shielding effect must be confirmed with a survey meter. Care must be taken in regard to leaking and scattering rays through clearances between piled bricks and blocks. A lead container is usually used to carry radioactive materials. The use of a shield to attenuate neutron beams means to reduce the neutron energy to a level of easy absorption. Heavy elements such as steel and lead are effective to reduce the neutron energy of over one MeV. After making neutron energy transfer by inelastic collision with such nuclei, neutrons are further slowed down by collision with light elements such as hydrogen. The role of the latter light elements is to reduce the neutron energy down to the level enabling neutron capture, i.e. thermal energy level (0.025 eV). In this connection, both gamma rays emitted by inelastic scattering caused by collision with heavy elements and gamma rays emitted by neutron capture when a neutron is absorbed into a light element must be taken into consideration. Water and paraffin wax containing much hydrogen are used to reduce the energy of a fast neutron. For example, a 25 cm thick paraffin wax can attenuate a fast neutron of 1 MeV by about a factor of 10. One mm thick cadmium sheet is adequate to absorb thermal neutrons. The concept of an attenuation coefficient and half-value thickness used in shielding against gamma rays can also be applied to shielding against neutron beams. It must be noted that the labyrinth does not have such a large shielding effect on neutron beams as it does on gamma rays. In a place of uniform dose rate, the exposure dose is proportional to the work time. To reduce the exposure dose of a worker, the work time must be reduced. The measures to achieve this end are as follows: (a) making a careful work plan in advance to do work
Chapter 6
268
10 -I
10 .2
Z
9 r./3 rae~
10 -3 Z [...,
6~
10 -4 J92Ir\
\~37Ce
10 .5
10 .6 0
5
10
15
20
25
30
T H I C K N E S S OF S T E E L / cm Fig. 6.4. Transmission of broad-beam gamma rays through steel (density: 7800 kg m-3).
efficiently; (b) carrying out a cold run in advance to become familiar with work procedure; and (c) using automatic devices and instruments. The protective measure of reducing work time is a measure to be taken only when the radiation level concerned is definitely known by measurement. This measure must be used only when protection cannot be secured by the methods of "distance" and "shielding". During work under a high dose rate, a worker must carry an alarm meter to control the work time strictly.
Radiation SafeO,
269
6.1.1 Internal contamination
Internal exposure results from intake of an unsealed radioactive material to the human body. Protection against internal exposure, therefore, can be achieved by preventing a radioactive material from being taken into the body. Entry routes of a radioactive material are the following: 1. Inhalation: entry through respiratory organs. 2. Ingestion: entry through gastrointestinal tract. 3. Direct absorption: absorption through intact or cut skin. Mathematical models are used to describe the various processes involved in the internal deposition and retention of radionuclides and the associated radiation doses received by various organs and tissues in the body. In models of this type, the body is viewed as a series of compartments into which the radioactive materials enter and exit at various rates, ultimately being removed from the system by some form of excretion, by radioactive decay, or both. Figure 6.5 represents a schematic view of a compartmentalised model of the body showing routes of entry of the materials (i.e., intake), compartments (e.g., organs), and excretion routes. INGESTION
I [
q
INHALATION
WOUNDS, SKINABSORPTIO~N
RESPIRATORY
SITEOF ENTRY
MECHANICAL ,~1 DISSOLUTAND CLEARANCE~ R P T I O N
BLOOD
( Fig. 6.5. Schematic representation of routes of entry, metabolic pathways and possible bioassay samples for internally deposited radionuclides.
270
Chapter 6
After a radionuclide intake has occurred, any radionuclide remaining within the organ of intake will irradiate that organ or tissue. The rate at which the radionuclide leaves the site of intake by dissolution and absorption into the blood depends on the physical and chemical properties of the materials present. The term "uptake" is used to describe the quantity of the radionuclides that is absorbed into the blood after an intake has occurred. After absorption, portions of the radionuclide will be deposited in, and irradiate other organs and tissues as illustrated in Fig. 6.5. Fractions of the radionuclide uptake are excreted by various routes, such as in urine or faeces. A portion of radionuclide deposited in an organ may be recycled (i.e., returned to the blood) so that it is again available for deposition in an organ or excretion from the body. A variety of in-vitro bioassay samples can be collected from an individual for the purpose of ascertaining whether an internal radionuclide deposition has occurred previously. Samples that may be collected and analysed for such a purpose include urine, faeces, hair, teeth, breath, tissue etc. Each of these samples provides an indirect measure of the internal radionuclide deposition because the organ and tissues of main concern related to the radionuclide are not being sampled. Thus, proper interpretation of these results requires knowledge of the relationships between the presence of a radionuclide in the various bioassay samples and organ radionuclide burdens of interest. Analysis of urine samples for excreted radionuclides is the most prevalent in-vitro bioassay procedure in use today. It is particularly useful for radionuclides that enter the body in a relatively transferable (i.e. soluble) form. When a radionuclide in such a form reaches the bloodstream, fractions will be deposited in various body organs and the remainder excreted, predominantly in the urine. The general sampling practice is to make periodic 24-h collections of an individual's urinary excretion. Presence of the radionuclide of interest at greater than the estimated background level indicates that some internal deposition may have occurred. Repetitive sampling provides a means of verifying the positive result and determining the average rate at which the excretion of the radionuclide is occurring at the time. Collection and analysis of daily faeces samples is another means of obtaining an indirect assessment of a possible internal radionuclide if the follow-up samples confirm the original observation, the overall results can be used to estimate the body burden existing at the time of the bioassay sampling, and if the time of exposure is known or can be estimated the initial body burden can also be estimated. This process is illustrated schematically in Fig. 6.6. The interpretation of bioassay results requires knowledge of the relationship between radionuclide excretion in the particular type of bioassay sample and the body burden or content existing at the time the samples were collected. Figure 6.6.a illustrates the retained body burden estimated from the analysis of bioassay sample at time t and the mathematical relationships from the excretion model. The initial body burden value can then be estimated using the retained body burden estimate and an assumed whole-body retention function for the particular radionuclide and its physical and chemical form.
Radiation Safety
271
;xd
Ca)
.d
Estimated Initial Burden
Z
o k~
9
[..
~ ~
Assumed Whole-Body l ~ R e t e n t i ~ Function
z
o' 9
Body Burden Estimated . at Time o f . hng
< 0
(b) z 9
Fz
oz [.<
t~et
9
.d
TIME AFTER EXPOSURE
Fig. 6.6. Schematic representation of a general approach that can be used to interpret the results of an in-vitro bioassay sample. The radionuclide quantities in (a) the whole body or (b) tissues or organs are expressed in logarithms because the retention curves, when expressed in such a manner, are often straight lines.
D u e to various biological, chemical and physical processes, the initial body burden computed above is distributed among several different organs and tissues. With the use of an appropriate metabolic model, the distribution of the initial body burden among various body organs can be estimated. The sum of these initial organ burdens among various body organs can be estimated. The sum of these initial organ burdens should equal the initial body burden. The radiation dose received by each organ would then be proportional to the area under that organ's respective retention curve shown in Fig. 6.6b.
272
Chapter 6
6.2 SOME DEFINITIONS (GLOSSARY) Absorbed dose: The fundamental dosimetric quantity, defined as: de D =~ dm
(6.13)
where D is the absorbed dose, de is the mean energy imparted by ionising radiation to matter in a volume element, and dm is the mass of the matter in this volume element. The energy can be averaged over any defined volume, the average dose being equal to the total energy imparted in the volume divided by the mass of the volume. The SI unit of absorbed dose is joule per kilogram (J kg -1) and its special name is gray (Gy).
Ambient dose equivalent: H*(d), at a point in a radiation field, is the dose equivalent that would be produced by the corresponding aligned and expanded field in the ICRU sphere at a depth d on the radius opposing the direction of the aligned field. A depth d = 10 mm is recommended for strongly penetrating radiation. Annual dose limits: Values of effective or equivalent dose to individuals from controlled practices, which shall not be exceeded in a year. Annual limit of intake (ALl): The intake by inhalation, ingestion or through the skin of a given radionuclide in a year by the reference man which would result in a committed dose equal to the relevant dose limit. Average glandular dose: Average glandular dose, Dg, in mammography can be computed from: Dg= DuNXa
(6.14)
where DgN is the average glandular absorbed dose resulting from an incident exposure in air of 2.58x 10-4 C kg -~ and X~ is the incident exposure in air.
Avertable dose: The dose to be saved by a protective action; that is, the difference between the doses expected with and without the protective action. Chronic exposure: Exposure persisting in time.
Radiation Safety
273
Collective effective dose: The total effective dose to a population, defined as" S -'-
EEiNi
(6.15)
i
where E i is the average effective dose in the population subgroup i and N i is the number of individuals in the subgroup. If can also be defined by the integral"
S=
-dE dE
(6.16)
dN where ~ dE is the number of individuals receiving an effective dose between E and E + d E . dE The collective effective dose committed by an event, a decision or a finite portion of a practice k, S k, is given by" S k = ~S~ (t)dt
(6.17)
0
where S k (t) is the collective effective dose rate at time t, caused by k.
Committed effective dose: After an elapsed time T following an intake of radioactive substances, the committed effective dose is defined as: to +T
E(T)=
~/~(t)dt
(6.18)
to
where to is the time of intake and/~ ( t)dt is the effective dose rate at the time t. When T is not specified it will be taken to be 50 years for adults and to age 70 years for intakes by children.
Committed equivalent dose: After an elapsed time T following an intake of radioactive substances, the committed equivalent dose is defined as" to +T
H r (T) =
~Hr(t)dt
(6.19)
t0
where to is the time of intake and H r (t) is the equivalent dose rate at time t in an organ or tissue T. When T is not specified it will be taken to be 50 years for adults and to age 70 years for intakes by children.
Chapter6
274
Critical group: A group of members of the public whose exposure for a given radiation source and given exposure pathway is reasonably homogeneous and is typical of individuals receiving the highest effective dose or equivalent dose (as relevant) by the given exposure pathway from the given source. Directional dose equivalent: H'(d,~), at a point in a radiation field, is the dose equivalent that would be produced by the corresponding expanded field in the ICRU sphere at depth d, on a radius in a specified direction, s A depth d = 0.07 mm is recommended for weakly penetrating radiation. Dose: Absorbed dose, organ dose, equivalent dose, effective dose, committed equivalent dose, or committed effective dose, depending on the context. The modifying terms are often omitted when they are not necessary for defining the quantity of interest. Dose constraint: A prospective upper bound on the individual dose which is used in the optimisation of protection and safety for sources. For occupational exposures, dose constraint is a source-related value of individual dose used to limit the range of options considered in the process of optimisation. For public exposure, the dose constraint is an upper bound on the annual doses that members of the public should receive from the planned operation of any controlled source. The exposure, to which the dose constraint applied is the annual dose to any critical group, summed over all exposure pathways, arising from the predicted operation of the controlled source. The constraint for each source should ensure that the sum of doses to the critical group from all controlled sources remains within the dose limit. For medical exposure the dose constraint levels should be interpreted as guidance levels, except when used in optimising the protection of persons exposed for medical research purposes or of persons, other than workers, who assist in the care, support or comfort of exposed patients. Effective dose: A summation of the tissue equivalent doses, each multiplied by the appropriate tissue-weighting factor: E = ~Wr H r
(6.20)
T
where H T is the equivalent dose in tissue T and WT is the tissue-weighting factor for tissue T. From the definition of equivalent dose, it follows that: (6.21) T
R
R
T
Radiation Safet3,
275
where WR is the radiation weighting factor for radiation R, and DT,e the average absorbed dose in the organ or tissue T. The unit of effective dose is Jkg -~, termed sievert
(Sv). Equivalent dose: The absorbed dose in an organ or tissue multiplied by the relevant radiation-weighting factor WR: HT.R = W~flgr.R
(6.22)
where DT,R is the average absorbed dose in the organ or tissue T and WRis the radiation weighting factor for radiation R. When the radiation field is composed of radiations with different values of WR, the equivalent dose is: H r = ~WRDT, g
(6.23)
R
The unit of equivalent dose is Jkg -~, termed sievert (Sv).
Exposure: Exposure of people to radiation or radioactive substances, which can either be external exposure from sources outside the body or internal exposure from sources inside the body. The exposure can be classified as either normal or potential exposure; either occupational, medical or public exposure; and, in intervention situations, either emergency or chronic exposure. Ke rma : The quotient of dEtr by dm, where dE,r is the sum of the initial kinetic energies of all the charged ionising particles liberated by uncharged ionising particles in a material of mass dm: d E tr
K =~ dm
(6.24)
The unit of kerma is Jkg -1 with the special name of gray (Gy); 1 Gy = 1 Jkg -1.
Multiple scan average dose: Multiple scan average dose (MSAD) is a term used in computed tomography MSAD = 1 r+,,1/2 -I J-n,/2 D( z )dz
(6.25)
where n is the total number of scans in a clinical series, I is the distance increment that separates scans and D(z) is the dose at position z, parallel to the z (rotational) axis.
276
Chapter 6
Organ dose: The mean dose D Tin a specified tissue or organ T of the human body given by"
D T = (1 / m r ) ~ Ddm
(6.26)
mr
where m r is the mass of the tissue or organ and D is the absorbed dose in the mass element din.
Personal dose equivalent: Hp(d) is defined for both strongly and weakly penetrating radiations. Hp(d) is the dose equivalent in soft tissue below a specified point on the body at an appropriate depth d. Depths of d = 10 mm for strongly penetrating radiation and d = 0.07 mm for weakly penetrating radiation are recommended.
Project dose: The dose to be expected if no protective or remedial action is taken.
Public exposure: Exposure incurred by members of the public from radiation sources, excluding any occupational or medical exposure and the normal local natural background radiation but including exposure from authorised sources and practices and from intervention situations.
Radiation weighting factor: A factor by which the absorbed dose is multiplied in order to account for the relative health hazard of different types of radiation. The values of radiation weighting factor used for radiation protection purposes are as follows. If calculation of the radiation-
Table 6.1 Radiation weighting factor Type and energy range of radiation
Radiation weighting factor W R
Photons; all energies
1
Electrons and muons; all energies
1
Neutrons; energy < 10 keV
5
10 keV to 100 keV
10
> 100 keV to 2 MeV
20
>2 MeV to 20 MeV
10
>20 MeV Protons other than recoil protons; energy > 2 MeV Alpha particles, fission fragments, heavy nuclei
5 5 20
Radiation Safe~
277
weighting factor for neutrons requires a continuous function, the following approximation can be used:
WR = 5 + 17-e -(ln(2E))2/6
(6.27)
where E is the neutron energy in MeV. For radiation types and energies not included in the table, WR can be taken to be equal to Q at 10 mm depth in the ICRU sphere and can be obtained as follows:
_
1!
Q = -~ Q(L)DLdL
(6.28)
where D is the absorbed dose, Q(L) is the quality factor in terms of the unrestricted linear energy transfer L in water, specified in ICRP Publication No. 60, and D L is the distribution of D in L. 1
for
L__ 10
Q(L) = 0.32L-2.2
for 10 < L < 100
300/~
for
(6.29)
L > 100
where L is expressed in keV ~tm-~.
Reference air kerma rate: The reference air kerma rate of a source is the kerma rate to air, in air, at a reference distance of one metre, corrected for air attenuation and scattering. This quantity is expressed in ~tGyh-~ at 1 m.
Reference level: Genetic term for action, intervention, investigation and recording levels. Such levels may be established for any of the quantities determined in the practice of radiation protection.
Reference man: An idealised caucasian person defined by the ICRP for the purpose of radiation protection assessments (ICRP- 1975).
Tissue weighting factors: A factor by which the equivalent dose to an organ or tissue is multiplied in order to account for the different sensitivities of different organs and tissues to the induction of stochastic effects of radiation. The tissue weighting factors used for radiation protection purposes are shown in Table 6.2.
278
Chapter 6
Table 6.2 The tissue weighting factor for radiation protection purposes Tissue or organ
Tissue weighting factor WT
Gonads Bone marrow (red) Colon a Lung Stomach Bladder Breast Liver Oesophagus Thyroid Skin Bone surface Remainderb
0.20 0.12 0.12 0.12 0.12 0.05 0.05 0.05 0.05 0.05 0.01 0.01 0.05
aLower large intestine. bFor the purposes of calculation, the remainder is composed of adrenal glands, brain, upper large intestine, small intestine, kidney, muscle, pancreas, spleen, thymus and uterus. In those exceptional cases in which a single one of the remaining tissues or organs receives an equivalent dose in excess of the highest dose in any of the twelve tissues or organs for which a weighting factor is specified, a weighting factor of 0.025 shall be applied to that tissue or organ and a weighting factor of 0.025 to the average dose in the rest of the remainder as defined here.
6.3 B A S I C S A F E T Y S T A N D A R D S To control the radiation exposure of workers, medical patients and the public, many countries have developed laws, which are supported by administrative measures and enforced by inspectors. Equally important is to have internationally agreed standards. The short history of dose limitations is presented in Table 6.3. It was with the increase in radiation injury that methods for the use of radiation without radiation injury started to be considered. In 1925, an American, Mutscheller, using data on X-ray technicians with no radiation injury, concluded that the quantity of exposure to radiation less than 1/100 of the erythema dose was the safe level. Converted to the present unit of radiation, this equals 2 mSv/day. This quantity was termed the "tolerance dose", i.e., the level which can be tolerated by man without radiation injury. At the Second International Congress of Radiology held in Stockholm in 1928, the International X-ray and Radium Protection Commission was established. At the fifth Congress held in London in 1950, the name of the Commission was changed to the International Commission on Radiological Protection, ICRP, and the term "tolerance dose" was changed to " m a x i m u m permissible dose". The m a x i m u m permissible dose for radiation workers was determined as 0.3 R/week (--3 mSv/week) or 0.2 R/day,
Radiation Safet3,
279
Table 6.3 History of the dose limitations Year
Recommending organisation
Definition of the limits
Name of the limits
1902
Rollins
Quantity of radiation which does not produce any image on the photographic film exposed for 7 min
Quantity which would not harm the human bodies
1925
Mutscheller
1/100 of the radiation dose which induces the erythema
Tolerance dose
1931
British X-ray and Radium Protection Committee
0.2 R/day
as above
1934
International X-ray and Radium Protection Commission
0.2 R/day
as above
1936
American X-ray and Radium Protection Committee
0.1 R/day
as above
1950
International Commission 0.3 R/week on Radiological Protection
Maximum permissible dose (MPD)
1954
as above
0.3 rem/week (15 rem/year)
Same as above
1958
as above
Radiation worker: 5(N-I 8) rem; N = age; 3 rem/week ; General public: 0.5 rem/year
Maximum permissible cumulative dose MPD Permissible limit
1965
as above
Radiation worker: 5 rem/year ; General public: 0.5 rem/year
MPD Dose limit
1977
as above
Effective dose equivalent 50 MSv/year; Lens of the eye 150 mSv/year*; Other tissues 500 mSv/year; Effective dose equivalent of general public 1 mSv/year**
Dose equivalent limit
*Changed from 300 mSv/year to 150 mSv/year in 1980. **Changed from 5 mSv/year to 1 mSv/year in 1985.
which was d e t e r m i n e d by the British X-rays and R a d i u m Protection C o m m i t t e e in 1931. T h e w a y in w h i c h the m a x i m u m permissible dose was u n d e r s t o o d to be at the time was that it s h o u l d be "the quantity of ionising radiation which is d e t e r m i n e d to not cause any somatic effects w h i c h m i g h t be detected at any m o m e n t of a p e r s o n ' s lifetime, b a s e d on the present u n d e r s t a n d i n g s on radiation". H o w e v e r , starting with the r e c o m m e n d a t i o n m a d e in 1958, the hereditary effects began to be included.
280
Chapter 6
Since then, in 1965, there was a new recommendation published, stating that the purpose of radiation protection is "to prevent the early effects of radiation, as well as to limit the late effects to within an acceptable level". Based on these basic principles, the maximum permissible dose was limited to 5 rem/year (5 mSv/year) for the general public. In the recommendation published in 1977, basic principles are "to prevent detrimental non-stochastic effects and to limit the probability of stochastic effects to levels deemed to be acceptable." Table 6.3 lists the history of the dose limitations. The International Atomic Energy Agency has played a lead role in developing and refining these standards. The IAEA~together with the World Health Organisation, International Labour Organisation, OECD Nuclear Energy Agency, Food and Agriculture Organisation and Pan American Health Organisation~recently revised and updated its international Basic Safety Standards (BSS or Standards in further text) for protection against ionising radiation and the safety of radiation sources. The Standards draw upon information derived from extensive research and development work by scientific and engineering organisations, at national and international levels, on the health effects of radiation and techniques for the safe design and operation of radiation sources; and upon experience in many countries in the use of radiation and nuclear techniques. The United Nations Scientific Committee of the Effects of Atomic Radiation (UNSCEAR), a body set up by the United Nations in 1955, compiles, assesses and disseminates information on the health effects of radiation and on levels of radiation exposure due to different sources; this information was taken into account in developing the Standards. Purely scientific considerations, however, are only part of the basis for decisions on protection and safety, and the Standards implicitly encourage decision-makers to make value judgements about the relative importance of risks of different kinds and about the balancing of risks and benefits. The new Standards are intended to ensure the safety of all types of radiation sources and to complement engineering safety standards developed for large and complex radiation sources, such as nuclear reactors and radioactive waste management facilities. The Standards are not mandatory, but can serve as a practical guide to all those involved in radiation protection, taking into account local situations, resources, etc. A wealth of new information about radiation exposure over the past decade prompted the revision of the BSS. First and foremost, a study of the biological effects of radiation doses received by the survivors of the atomic bombing of Hiroshima and Nagasaki suggested that exposure to low-level radiation was more likely to cause harm than previously estimated. Other developments--notably the nuclear accident at Three Mile Island in 1979 and that at Chernobyl in 1986, with its unprecedented transboundary contamination~had a profound effect on the public perception of the potential danger from radiation exposure. There were serious accidents with radiation sources used in medicine and industry in Mexico, Brazil, El Salvador and other countries. In addition, more has been discovered about natural radiation--such as household r a d o n ~ a s a cause of concern for health. Finally, natural radiation exposures of workers such as miners, who were not thought of as radiation workers, were discovered to be much higher than had been realised.
Radiation Safety
281
The BSS apply to both "practices" and "interventions": practices are activities that add radiation exposure to that which people normally receive due to background radiation, or that increase the likelihood of incurring exposure. These include the use of radiation or radioactive substances for medical, industrial, agricultural, educational, training and research purposes and, of course, the generation of energy by nuclear power. Also included are facilities containing radioactive substances or devices such as irradiation installations, mines and mills processing radioactive ores and radioactive waste management facilities. Interventions are any activities that seek to reduce the existing radiation exposure situations such as radon in buildings, and emergency situations such as those created by contamination in the aftermath of an accident. Protection under the BSS is based on the principles of the International Commission on Radiological Protection, which can be summed up as follows: 9 Justification of the practice. No practice involving exposure to radiation should be adopted unless it produces a benefit that outweighs the harm it causes or could cause. 9 Optimisation of protection. Radiation doses and risks should be kept as low as reasonably achievable with economic and social factors being taken into account; constraints should be applied to dose or risk to prevent an unfair distribution of exposure or risk. 9 Limitation of individual risk. Exposure of individuals should not exceed specified dose limits above which the dose or risk would be deemed unacceptable. All three principles apply to the protection of workers and the public. However, to protect patients during the medical use of ionising radiation only justification and optimisation apply. Dose limits are not applicable to medical exposure, but guidance levels which show what is achievable by good practice may be established for use by medical practitioners. Dose limits are also inapplicable to interventions, which are concerned with reducing exposure. The objective of the BSS is to prevent the occurrence of short-term effects of high doses of radiation and to restrict the likelihood of occurrence of long term effects, assuming protection of the exposed individuals and by ensuring the safety of the source of exposure. For any justified interventions, the objective is achieved by keeping the individual doses lower than the threshold levels for deterministic effects and keeping all doses as low as reasonably achievable in the circumstances. Justification of practices and interventions involves many factors, including social and political aspects, as well as radiological considerations. Some practical guidance on justification for practices and interventions is provided by the BSS, and some examples are provided here: an intervention is justified if it is expected to achieve more good than harm, having regard to health, social and economic factors. Protective actions are nearly always justified if, in the absence of intervention, doses are expected to approach certain specified values related to deterministic effects. Unjustified practices are: 9 Addition of radioactive materials to food, beverages or cosmetics.
282
Chapter 6
9 Use of radioactive materials in toys and jewellery. ~ Certain medical exposures, e.g., the exposure of population groups for mass screening, unless the benefit outweighs the risk. A short summary in Fig. 6.7 shows BSS at glance: implicit quantitative requirements and guidance for practice. The BSS list detailed requirements for practices and interventions to protect workers, patients and the general public from radiation exposure. They also recommend procedures for ensuring the safety of sources, for accident prevention, for emergency planning and preparedness and for mitigating the consequences of accidents. Although the majority are of a qualitative nature, the BSS also establish many requirements expressed in terms of restrictions or guidance on the dose that may be incurred by people. The range of doses spreads over four orders of magnitude, from 10 3 - -
Annual dose
-
(mSv)
i Iniervention always justified] 10 2
_
_
Limits for workers undertaking intervention Limits for workers (normal practice, yearly) Limits for workers (normal practice, average)
~,q
~,d
.,d
I
10'
~ - q Range for optimized interventions] - - I --
I
--
I
I
: [Range for optimized remedial action (radon)[
,.41--' Range for optimized protection ,
10 ~
_ _ __
and constrains (occupational) World average background exposure
Public dose limit (individual members of the public)
~,d ~,q
Range of constrains (public, individual sources) I I
10 -I
I
!
,',4-- Range of optimized protection (public) I I ! ! I I 0-2
__
~1
Exemption level
Fig. 6.7. BSS at a glance: implicit quantitative requirements and guidance for practices.
Radiation Safe O,
283
ones that are so minute that they should be exempt from the requirements, to doses that are large enough to make intervention almost mandatory. National governments usually have the responsibility for enforcing radiation safety standards, generally through a system that includes a regulatory authority. In addition, governments usually provide for certain essential services for radiation protection and safety and for interventions that exceed or that complement the capabilities of regulators. The BSS can only be effectively applied when such a national infrastructure is firmly in place. In addition to legislation and regulations, the essential elements are: 9 A regulatory authority. This should be empowered to authorise and inspect, and to enforce the legislation and regulations. It must have sufficient resources, including adequate numbers of trained personnel. There must be arrangements for detecting any build up or radioactive substances in the general environment, for disposing of radioactive waste and preparing for interventions, particularly during emergencies, that could result in exposure of the public. ~ Education, training and public information. There must be adequate arrangements and resources for these, as well as for the exchange of information among specialists. There must also be appropriate means of informing the public, its representatives and the information media about health and safety concerns. Facilities and services for radiation protection and safety must be well established at the national level. These include laboratories for personal dosimetry and environmental monitoring, and calibration and intercomparison of radiation measuring equipment; they could also include central registries for radiation dose records and information on equipment reliability. For the detailed recommendations, we refer the reader to the full text of the Basic Safety Standards published by IAEA. Here we list some of the highlights:
6.3.1 Occupational exposure Occupational exposures are limited, as seen in Section 6.3.3. Here we list paragraphs as listed in the conditions of service which are of special importance:
(i) Pregnant workers: A female worker should, on becoming aware that she is pregnant, notify the employer in order that her working conditions may be modified if necessary. The notification of pregnancy shall not be considered a reason to exclude a female worker from work; however, the employer shall adapt the working conditions of a female worker who has notified pregnancy with respect to occupational exposure, in order to ensure that the embryo or foetus be afforded the same broad level of protection as required for members of the public. (ii) Conditions for young persons: No person under the age of 16 years shall be subjected to occupational exposure.
284
Chapter 6
No person under the age of 18 years shall be allowed to work in a controlled area unless supervised and then only for the purpose of training. 6.3.2 Medical exposure
BSS states that registrants and licensees shall ensure that: a. no patient be administered a diagnostic or therapeutic medical exposure unless the exposure is prescribed by a medical practitioner; b. medical practitioners be assigned the primary task and obligation of ensuring overall patient protection and safety when prescribing, and during the delivery of, medical exposure; c. medical and paramedical personnel be available as needed, and either be health professionals or have appropriate training adequately to discharge assigned tasks in the conduct of the diagnostic or therapeutic procedure that the medical practitioner prescribes; d. for therapeutic uses of radiation (including teletherapy and branchytherapy), the calibration, dosimetry and quality assurance requirements of the Standards be conducted by or under the supervision of a qualified expert in radiotherapy physics. Medical exposures should be justified by weighing the diagnostic or therapeutic benefits they produce against the radiation detriment they might cause, taking into account the benefits and risks of available alternative techniques that do not involve medical exposure. In justifying each type of diagnostic examination by radiography, fluoroscopy or nuclear medicine, relevant guidelines will be taken into account, such as those established by the WHO (1983, 1987, 1990). Mass screening of population groups involving medical exposure is deemed to be not justified unless the expected advantages for the individuals examined or for the population as a whole are sufficient to compensate for the economic and social costs, including the radiation detriment. Account should be taken in justification of the potential of the screening procedure for detecting disease, the likelihood of effective treatment of cases detected and, for certain diseases, the advantages to the community from the control of the disease. 6.3.3 Dose limits
The dose limits specified here apply to exposures attributable to practices, with the exception of medical exposures from natural sources that cannot reasonably be regarded as being under the responsibility of any principal party of the Standards.
A. Occupational Exposure Dose Limits: The occupational exposure of any worker shall be so controlled that the following limits are not exceeded:
Radiation Safe~'
285
a. b. c. d.
an effective dose of 20 mSv per year averaged over five consecutive years; an effective dose of 50 mSv in any single year; an equivalent dose to the lens of the eye of 150 mSv in a year; and an equivalent dose to the extremities (hands and feet) or the skin of 500 mSv in a year. For apprentices of 16-18 years of age who are training for employment involving exposure to radiation and for students of age 16-18 who are required to use sources in the course of their studies, the occupational exposure shall be so controlled that the following limits are not exceeded: a. an effective dose of 6 mSv in a year; b. an equivalent dose to the lens of the eye of 50 mSv in a year; and c. an equivalent dose to the extremities or the skin of 150 mSv in a year. When, in special circumstances, a temporary change in the dose limitation requirements is approved: a. the dose averaging period may exceptionally be up to 10 consecutive years as specified by the Regulatory Authority, and the effective dose for any worker shall not exceed 20 mSv per year averaged over this period and shall not exceed 50 mSv in any single year, and the circumstances shall be reviewed when the dose accumulated by any worker since the start of the extended averaging period reaches 100 mSv; or b. the temporary change in the dose limitation shall be as specified by the Regulatory Authority but shall not exceed 50 mSv in any year and the period of the temporary change shall not exceed 5 years.
B. Public Exposure Dose Limits: Exposure of members of the public attributable to practices shall not exceed the following limits which shall apply to the estimated average doses to the relevant critical groups: a. an effective dose of 1 mSv in a year; b. in special circumstances, an effective dose to up to 5 mSv in a single year provided that the average dose over five consecutive years does not exceed 1 mSv per year; c. an equivalent dose to the lens of the eye of 15 mSv in a year; and d. an equivalent dose to the skin of 50 mSv in a year. C. Dose Limitation for Comforters of Patients and Visitors to Patients: The dose limits set out in this part shall not apply to comforters of patients, i.e., to individuals knowingly exposed while voluntarily helping (other than in their employment or occupation) in the care, support and comfort, including visiting, of patients undergoing medical diagnosis or treatment. However, the exposure of any such comforter of patients shall be constrained so that it is unlikely that his or her exposure will exceed 5 mSv during the period of a patient' s diagnostic examination or treatment. The
286
Chapter 6
dose to children visiting patients who have ingested radioactive materials should be constrained to less than 1 mSv per treatment. The dose limits apply to the sum of the relevant doses from external exposure in the specified period and the relevant committed doses from intakes in the same period; the period for calculating the committed dose shall normally be 50 years for intakes by adults and to age 70 years for intakes by children. For the purpose of demonstrating compliance with dose limits, the sum of the personal dose equivalent from external exposure to penetrating radiation in the specified period and the committed equivalent dose or committed effective dose, as appropriate, from intakes of radioactive substances in the same period shall be taken into account.
6.4 RADIATION DOSE ASSESSMENT Radioactive materials released to the environment reach man through various pathways. Direct exposure from radioactive sources, inhalation and ingestion must be taken into account when making an assessment of the dose to the critical group. Effluents will have different compositions and the physical, chemical and biological properties of the environments into which effluents are released will differ. This is potentially a complex study but an outline study of all the problems will show which nuclides and which pathways are important. In most cases only a few nuclides in a few pathways will surface and the detailed observation of these nuclides becomes an essential task. This may lead to defining a critical pathway through which contaminants reach the critical group. The concept of the critical group was formulated by the ICRP in Publication 26. The local population will normally include the critical group and is geographically close to the point of release or will include consumers of locally harvested foods. In selecting a critical group two requirements must be met: 1. The critical group must be representative of those expected to receive the highest doses. 2. The critical group should be homogeneous with respect to those factors that affect the doses received. The factors that affect the doses must be identified. The major factors are physiological and metabolic characteristics, age and diet. For internal exposures, a range of dietary habits corresponding to a ratio of not more than three between the maximum observed and the mean characterising the critical group is considered sufficient to satisfy the homogeneous criterion. For inhalation, the air intake does not vary widely among people of the same age group. Default values taken from ICRP publication 23 for adults and one-year-old children may be used for the critical group. Because behaviour is more variable than metabolism within a population, the homogeneity criterion does not apply to external exposures. The nature of the critical group is likely to change with time and it is advisable to make allowances for these changes.
Radiation SafeO'
287
6.4.1 Stages from release to human exposure Two different scenarios should be considered: (i) for atmospheric release and (ii) for liquid release. Once the source term is known and the release takes place, the following occurs.
(i) For atmospheric release: 1. Transport by winds and dispersion 2. Deposition 3. Uptake by plants and animals 4. Intake via food chain by man Exposure of man occurs at step 1 above via inhalation of diluted/dispersed release. At step 2, there is direct contamination and direction radiation from cloud of activity or from layer on ground; step 2 can also deposit on water or from run-off of soil into water bodies and further food chain steps. Step 3 is an interim step where reconcentration can occur. In Step 4 intake depends on the habits of the critical group, i.e. what food is eaten, how it is eaten, how much is eaten. Steps 2 and 4 also depend on population habits. It is important to determine eating and drinking habits of the population as a whole and identify the critical group. The Gaussian plume model is the most widely used method of estimating downwind concentrations of airborne material released to the atmosphere. It has been verified under widely different meteorological conditions, although the predicted results generally tend to be on the conservative side. It is useful for making first order approximations that are probably accurate to within a factor of 10. The approach used is that outlined by Dodd B. and Humphries L.L., (1988) although significant input was also obtained from IAEA Safety Series No. 57, Genetic Models and Parameters for Assessing the Environmental Transfer of Radionuclides from Routine Releases: Exposures of Critical Groups (1982). The dilution in air due to the atmospheric dispersion of radioactive aerosols resulting from a release of radioactive material was modelled for positions along the centre line of the plume using the diffusion equation with a "top hat" distribution to account for the directional fluctuations in crosswinds: X(x, z) __ Qfd
~/z
1 6fi-o' ~.l~
[e-~ z-h)~/2o~ + e-{ ~-h)~/2o, ]
(6.30)
Y
where Z(x,z) = air concentration (Bq m -3) at a point (x,z) downwind of the source, Q = release rate (Bq s-~), (~'y "" horizontal Gaussian dispersion coefficient (m), ~z = vertical Gaussian dispersion coefficient (m), ~ = average wind speed in downwind direction (m -1 s ), h = height of release (m), z = height of receptor, x = receptor downwind displacement (m), and fd is a factor accounting for dry deposition given by
Chapter6
288
fd
=
exp[- vd x]
where v d = the deposition velocity (m s-l), and z = the effective height of the plume =
~ . x exp
Cyyand cyz are dependent on the release height and Pasquill atmospheric stability classes. These classes are designated A to F in order of increasing atmospheric stability. To get an indication of the maximum and minimum concentrations, Z(x,z) may be calculated for each of these Pasquill categories. (ii) For liquid release The stages are similar to atmospheric with the important difference that aquatic organismsmmussels, crabs etc.--often have a very high preconcentration factor and that, over time, sediments in the water can also reconcentrate released radionuclides. Plants and organisms making use of sediments then become contaminated and pass this on along the food chain. The public can be exposed via different pathways from the release of radionuclides. For example, for liquid effluent released to a river, the following should be considered; drinking water, eating fish, irrigation of crops and subsequent consumption, animals drinking the water and subsequent human consumption of animal products, and swimming. Only some of the pathways can be considered critical, that is that they contribute most to the exposure dose and may approach the dose upper bound. This depends on the site-specific parameters and on the physico/chemical behaviour of individual radionuclides. For example:
Iodine fallout > grass > cows > milk > human consumption This is a very efficient concentration pathway. Caesium in river > uptake by fish > human consumption This can be a critical path. Caesium in river > irrigation water > uptake in crops > human consumption This is low, relative to the fish pathway because Cs' s biological concentration factor for Cs in plants is much lower than in fish. The critical pathway is also dependent on the habits of the exposed population. If no one ever eats any fish from the river then that pathway can be crossed off. Alternatively if a certain foodstuff is consumed in great quantities and the concentrations are reasonable then that foodstuff pathway can be critical.
Radiation Safer>'
289
6.4.2 Dose assessment models
Because there are so many factors involved in establishing dose pathways and which may be the critical pathway, various dose assessment computer based models have been developed. These models are also used to calculate the dose per unit release of a specified radionuclide for various pathways. The Release Upper Bound, RUB can be evaluated using a dose assessment model. This is done by varying the source term used such that the resulting dose equals the dose constraint or limit. The various exposure scenarios and pathways are also chosen for importance. The names of some dose models are: LADTAP; GENII; MICRO AIRDOS. Following the new recommendations by the ICRP in Publication 60 and the new model of the human respiratory tract in Publication 66 (Fig. 6.8), there is a need to revise the estimates of dose from the inhalation or ingestion of radionuclides. Following the publication of its 1990 recommendations, the ICRP issued Publication 61 which gave annual limits on intake (ALIs) for ingestion and inhalation of radionuclides by workers. It uses the new tissue weighing factors, w r, but with the biokinetic models given in Publication 30, most of which were published between 1979 and 1981. A revision of Publication 61 has now been completed, giving inhalation dose coefficients for workers according to the new respiratory tract model. Ingestion dose coefficients are also included, and revised biokinetic parameters for adults from Publications 56, 57 and 69 have been applied in the calculations. I I
I I
Anterior nasal
~ Extrathoracic
Environment ~
JI I I '
! 0.001
I
Nasooropharynx/~
larynx
~
~" I
_ Surface _ ~ trans_po_rt_
Sequestered in tissue 0.01
'
10
I ! I
0.01
LNTH[
s
! I I
I
I !
I !
Bronchioles ~I
J
I I
.....
i I !
I00
Gastrointestinal] tract ll~
ET2
L
Bronchi
1
I It
0.0001
1
Alveolar i j interstitium ~ I i Thoracic
0.00002
bb,
d' t ' A'd", l l I
0.001
t
I
I
I I
Fig. 6.8. Compartment model representing time-dependent particle transport from each respiratory tract region in the new ICRP model. The rates are in units of reciprocal days.
290
Chapter 6
The new ICRP respiratory tract model has three principal purposes: 9 to provide a qualitative and quantitative description of the respiratory tract as a route for radionuclides to enter the body, 9 to provide a scientifically acceptable method to calculate radiation doses to the respiratory tract for any given exposure, 9 to provide information about the transfer of radionuclides from the respiratory tract to other tissues, to calculate total body doses. The new model is based on the premise that, because of the large differences in radiosensitivity of the many respiratory tract tissues, and the wide variation in doses they may receive, specific tissue doses should be calculated. The ICRP Publication 30 model, however, provides only the average dose to the lungs. Average tissue doses to defined respiratory tract regions are calculated, the weighted sum of which provides an equivalent dose to the thoracic or extrathoracic region which is consistent with the ICRP tissue weighted dosimetry system. The model applies to all members of the population. Reference values are given for children aged 3 months, 1, 5 and 10 years, and for male and female 15-year-olds, and adults. Guidance is provided for adjusting the model for the effects of factors such as smoking, diseases and pollutants. The model is applicable to determining intake limits, and for assessing doses from exposures to both radioactive particles and gases. For modelling purposes the respiratory tract is represented by five regions, based principally on radiobiological considerations, but also taking account of differences in respiratory function, deposition and clearance (see Fig. 6.9). The Extrathoracic (ET) airways are divided into ETa, the anterior nasal passage, and E T 2, which consists of the posterior nasal and oral passages, the pharyns and laryns. The thoracic regions are Bronchial (BB, airway generations 1-8), Bronchiolar (bb), and Alveolar-Interstitial (AI, the gas exchange region). Lymphatics are associated with the extrathoracic and thoracic airways (LNET and LNT,, respectively). LUDEP (LUng Dose Evaluation Program) was developed concurrently with the new ICRP respiratory tract model. LUDEP was designed principally for two applications: 9 to help the ICRP Task Group sample the proposed model in detail by testing the predictions of deposition, clearance and retention against experimental data, and by determining the model's implications for doses to the respiratory tract 9 to test the practicality of implementing the model. LUDEP was originally designed for calculating doses only to the respiratory tract, but has subsequently been extended to calculate doses to all body organs in order to determine the wider dosimetric implications of the new model. As part of the development of the new respiratory tract model, there was an extensive period of consultation with interested parties about different aspects of the model, and several papers describing the model were published. An earlier version of LUDEP was also distributed informally to interested parties at that time. However, many of the methods and default parameter values have since been modified. The current version (Version 1.0) is the only version of LUDEP which implements the
Radiation SafeO'
291
Fig. 6.9. Respiratory tract regions defined in the new ICRP model.
model structure and parameter values approved by ICRP and described in ICRP Publication 66. Further versions of LUDEP are under development, since there are many additional capabilities which would be useful. For example, it is not in general possible to include the contribution to dose from the radioactive progeny of a selected radionuclide, although it is possible for certain radionuclides. Age-dependency is not addressed, as the biokinetic models and organ dosimetry implemented in LUDEP 1.0 are applicable only to adults. It would also be useful to predict urinary and faecal excretion rates and carry out other calculations needed for bioassay interpretation. These areas are currently under development, and it is intended that improved methods of dealing with all radioactive decay products, bioassay calculations and age-dependence will be included in a future version of LUDEP. Furthermore, as the models and methods employed in radiological protection are refined or changed, these may be incorporated in successive versions of LUDEP (Bircholl et al., 1994; Jarvis and Bircholl, 1994).
292
Chapter6
6.5 DOSIMETRY As the use of radioactivity in industry, medicine and power generation increases, there is a general need for dosimetric systems that can be applied to continuous monitoring and accident situations. This is a very broad field and we shall describe only some applications.
6.5.1 Dosimetry in radiotherapy For many common types of tumours it has been clinically observed that a small change in the dose level could change the outcome considerably, e.g. one per cent increase of the dose could increase the probability of tumour local control by several per cent (Brahme et al., 1988). The radiotherapist does not, however, dare to increase the dose over a certain level because of the risk that severe complications may appear in normal tissue. In modern radiotherapy much effort is devoted to finding the dose level for maximum benefit to the patient, that is, the maximum number of patients with uncomplicated local tumour control. This also implies that great accuracy is needed in order to deliver the prescribed dose. For this reason a working group of the EORTC (European Organisation for Research and Treatment of Cancer) states that acceptable practice is +_3% on the calibration of therapy units and +_5% on the delivery of prescribed dose (Johansson et al., 1988). In order to meet these requirements very accurate dose measurements are needed for calibration of the accelerator dose monitor, for measurements of dose distributions, and for regular checks of the radiation beams. In addition, in vivo measurements are needed to verify the dose plan for each individual patient and for checking the precision of the patient set-up. There are a large number of basic requirements for a dosemeter for radiotherapy dosimetry, e.g. regarding precision, recombination losses, temperature dependence, simplicity of handling, and energy dependence. These requirements restrict the number of dosimetry systems in common use today to TL dosemeters, silicon detectors, and film dosemeters. Some of these performance parameters might be improved for a special type of dosemeter. Other properties, inherent for a special detector, are difficult or impossible to change. The energy dependence belongs to this latter class. Every y or X-radiation dosimeter presents a response which depends on the photon energy. Using such dosimeters for the assessment of exposure in a y or X-radiation field, it is necessary to know the energies present in the radiation field in order to consider the response characteristics of the instruments as a function of their energy dependence. The methods used to determine the energy spectrum of a radiation beam are not of practical application routinely. However, some parameters may be used to characterise the energy spectral distribution. This is the case in the use of the "Half Value Layer" HVL concept. Using this parameter, the energy spectral distribution of a radiation beam is characterised by a unique energy value, namely, the "Effective Energy". Using two dosimeters with different energy dependencies, it is possible to determine the effective energy. This method, called the "Tandem Method", consists of using
Radiation Safe~.
293
the energy dependence of the ratio between the values of the energy calibration curves of the dosimeter responses, to determine the effective energy of a radiation beam in order to make the correction of one Tandem dosimeter reading due to its energy dependence, possible. The accuracy of this method depends not only on the inherent uncertainties of the Tandem dosimeters, but also on how pronounced are the energy dependencies of their responses. The more different these dependencies, the higher the ratio between them and, consequently, the higher the accuracy in the determination of the effective energy (Da Rosa and Nette, 1988). In addition to TL dosimeters the silicon diode detectors are extensively used in radiotherapy dosimetry. The man uses are for in vivo dosimetry and for relative dose distribution measurements. The advantage of in vivo dosimetry is that the detector is small and mechanically stable. Furthermore, it is often of great value to have direct measurement during irradiation. The silicon detectors are of both n- and p-type. The n-Si is obtained when silicon is doped with a material from group V in the periodic system, usually phosphorus, and p-Si when the doping material is a group III, usually boron. The n-type has a surplus of free valence electrons available for electrical conductivity. The p-type has instead a surplus of positive charge. To create a p-type diode a thin film of n-type silicon is fixed on top of a p-si disc and vice versa. A new n-Si detector has about twice the sensitivity of a p-Si detector. However, the sensitivity rapidly decreases with the pre-irradiated dose of the n-Si detector. Furthermore, the sensitivity varies for the n-Si detector with dose rate for pulsed radiation of dose rates common in therapy (Rikner and Grusell, 1987a). The magnitude of this variation is a function of the pre-irradiated dose. The n-type of detector is, for these reasons, very difficult to use in accurate dose measurements. The sensitivity of a p-Si detector also varies with the pre-irradiated dose. However, the drop in sensitivity is fairly small after a few kGy, about 2% per kGy has been reported for 20 MeV electron irradiation. The sensitivity drop per unit dose is much smaller, about one magnitude, for irradiation with high energy X-rays than with electrons (Rikner and Grusell, 1987b). A large number of departments rely mainly on film dosimetry for dose distribution measurements. The average of the method is that very short "accelerator time" is needed for the irradiation of the film which is of great importance in a busy radiotherapy centre. Furthermore, the film method might be the only one possible in practice for some purpose, e.g. measurements of distributions from scanned electron beams, and determination of the distribution in regions where the dose gradient is very steep. The film method is generally not very accurate. The reproducibility in density for different areas of the same film can be expected to be about 2% (ICRP, 1984). Also, to reach this precision, the processing conditions need to be optimised. 6.5.2 The I A E A / W H O network of SSDLS Personnel dosimetry is a proper and most effective means of assuring compliance with regulations governing the use of radioactive materials and ionising radiation. The
Chapter 6
294
requirement to keep radiation exposure "as low as reasonably achievable", linked with the growing number of workers whose exposure to radiation must be strictly controlled, requires intensification of efforts directed towards improvement of monitoring programmes. In 1968 the IAEA with the participation of WHO conveyed a panel of experts to discuss the dosimetric requirements of radiotherapy centres. The panel recommended the setting up of dosimeter calibration centres (later called Secondary Standard Dosimetry Laboratories--SSDLs) in developing countries. In 1974 experts, mainly from the large national standard laboratories, discussed the concept of SSDLs and their role in metrology. An SSDL was defined as a laboratory designated by the competent national authority to undertake dosimetry calibrations. For the proper function of the SSDLs, the need for dose intercomparison and for coordination of the work of individual laboratories was recognised. This ultimately led to the establishment of the international IAEA/WHO Network of SSDLs. The network's secretariat is shared by IAEA and WHO, with the Agency's Dosimetry Laboratory (DOL) functioning as the network's central laboratory. Support to the Network is given by most of the major national laboratories, by international bodies, e.g. B IPM, ICRU, IEC, OIML and by an SSDL Scientific Committee. Today the Network comprises 50 laboratories, 36 of them located in developing countries. Most of them are equipped with modern instrumentation including radiation generators and secondary standard dosemeters (with air kerma and/or exposure calibration factors) suitable for therapy and radiation protection dose rate levels. For performance evaluation, all SSDLs participate biannually in a postal dose intercomparison using TLD powder, a system developed by the Agency's DOL (Eisenlohr et al., 1977). The intercomparison quantity is absorbed dose to water (dose of 2 Gy) from 6~ irradiation determined in the IAEA water phantom (30x30x30 cm 3) ..! Irradiation| "l 2Gy I. . . .
I I
9
IAEA ( Virgin ~ _ ~ I Dosimetry TL Annealing Laboratory powder ]
:'i [Irradiation] 9 ~P' / [ ~', calibration I IEVALUATIO~ [irradiation"'"'[ I curve I k 1
/2oy 1.............. .......... : l lab. ref.
't..;.
I ....................................................
Radiotherapy centres
Time scale
.:
J Irradiation ]. . . . . "1 2 Gy ]
I
Sev"rale days
I About 2 months
I u!
1 week I I Between ~: 2 and 11 months
Fig. 6.10. The IAEA WHO dosimetry service to hospitals. The procedure is administratively complicated as the service is world-wide. It can be seen that parallel measurements are always performed at a PSDL and hospitals to secure traceability of dose calibration (After Svensson et al., 1996).
Radiation Safet3,
295
with secondary standard dosemeters calibrated in air kerma or exposure, applying conversion factors published in the open literature (see Fig. 6.10 for the illustration). The 1987/88 results showed that out of 35 participating SSDLs only 6 had a deviation of more than 3% between their stated dose and the dose determined by IAEA and the highest deviation was 5.4%. BIPM had participated in this intercomparison exercise as a reference laboratory; IAEA deviated +0.6% (Nette et al., 1989). We shall discuss results of such intercomparisons for Australian Radiation Laboratory, ARL (Huntley and Nette, 1993) for the years 1971-1992. Every Australian radiotherapy centre participated in the intercomparison. There were sufficient TLD sets provided by the IAEA to enable at least one beam quality to be included from every centre. There were also two special TLD sets provided (together with special jigs) to provide a check on D20/D~0 measurements at two centres. The overall result was quite satisfactory. The IAEA have stated in correspondence with ARL relating to the TLD intercomparison program that: "...the precision of the TLD measurements can be stated to be about 2.4% on the 2c level. Taking into account the uncertainty of the secondary standard dosimeter and the uncertainty of the determination of water absorbed dose within the intercomparison procedure is unlikely to exceed 3.5% on the 2~ level." In view of this, and in view of ARL's experience over the years, there were only two results of significant concern, and these have been adequately addressed (a single capsule receiving only half the required dose, and a 15.4% deviation due to a misunderstanding. Some Australian radiotherapy centres have not yet adopted the IAEA dosimetry protocol TRS 277. This protocol has been recommended by the Australian College of Physical Scientists and Engineers in Medicine, and calibration factors supplied by ARL are appropriate for the use of this protocol. ARL strongly recommends the use of TRS 277 in a water phantom by all radiotherapy centres in Australia (Huntley and Nette, 1993). Many quality assurance procedures in diagnostic radiology require the use of a phantom to simulate the X-ray attenuation of the patient. The phantom should transmit the same quantity and quality (i.e. spectrum) of radiation as that transmitted by the patient. The knowledge of photon spectra at the position of measurement on the surface, inside and behind standard dosimetric and imaging phantoms is helpful for performing dosimetric or calibration measurements and hence helpful in the course of quality control. The experimental approach is limited; however, simulation with Monte Carlo methods has been proved to be a very powerful technique (Petoussi et al., 1992).
6.5.3 High-dose dosimetry Radiation processing plays an important role in industry, health care, agriculture and environmental technology. It involves the use of large radionuclide (gamma-radiation) sources and electron accelerators in industrial and institutional facilities. Quality assurance is vital for the success of this technology. In fact, it is indispensable for
296
Chapter 6
health-related processes, such as the sterilisation of medical products and pharmaceuticals, the treatment of food, the control of insect pests, and the recycling and purification of municipal and industrial wastes and water supplies. This is evident from the strong emphasis that is placed on quality assurance in international and regional technological and standards documents such as those issued by the International Organisation for Standardisation (ISO), World Health Organisation (WHO), Food and Agriculture Organisation (FAO), Codex Alimentarius Commission (CAC), European Committee for Standardisation (CEN), International Consultative Group on Food Irradiation (ICGFI), European Confederation of Medical Suppliers Association (EUCOMED), and Association for Advancement of Medical Instrumentation (AAMI). Standardised dosimetry over the dose range (101-105 Gy) encompassing industrial processes is a key component for achieving and documenting quality assurance. The International Atomic Energy Agency (IAEA) in Vienna, Austria is implementing its high-dose dosimetry programme. The primary goals of the IAEA HighDose Dosimetry Programme have been: 9 to encourage the development and improvement of dosimetry techniques and procedures to ensure consistent dose measurements on a world-wide basis by use of industrial radiation processing facilities, by operators of pilot-scale facilities and by researchers performing studies on radiation effects; 9 to establish and operate the International Dose Assurance Service for radiation facilities, especially those in developing countries without easy access to national dosimetry calibration laboratories, at the request of the Member States; 9 to simulate international technological activities associated with standardisation and traceability of high-dose measurements; 9 to encourage research and development related to high-dose dosimetry techniques; 9 to foster international communication and regional co-operation in the technological practice associated with the use of radionuclides and industrial accelerators, where dosimetry is required. The overall accomplishments of the High-Dose Dosimetry Programme have been: 9 establishment and operation of the international dose Assurance Service (IDAS) for gamma radiation facilities; 9 development and improvement of alanine/ESR as a transfer and reference dosimetry system; 9 improvement of the performance of the high-dose dosimetry systems most widely used through studies of the influences of environmental factors. At the FAO, IAEA, WHO, ITC-UNCTAD/GATT International Conference on the Acceptance, Control of and Trace in Irradiated Food, convened in Geneva, Switzerland in 1988, it was agreed that, because of the nature of the process of food irradiation, which makes it difficult at present to determine the circumstances of irradiation by examination of the food, control of irradiated food has to be established through legally based administrative procedures. This involves accurate dosimetry in accordance with technical guidelines such as developed by the ASTM, proper record-keeping and control by the facility, followed by the issue of certificates. These procedures, whether
Radiation Safe~
297
the product is intended for domestic use or export, should include on the one hand a system of documentation allowing each batch of irradiated food to be identified with the irradiation facility and with the treatment given, and on the other hand a system of labelling. Other methods of control and compliance should be considered as technology progresses; therefore, research on analytical methods for identification of radiation-processed food in trade should be encouraged. Traditionally, radiation dose-setting criteria in radiation-sterilisation processing was based on inactivation efficacy to the defined number of D~0 cycles for a population of highly-radio-resistant indicator bacterium (such as Bacillus pumilus in the case of radiation sterilisation). That approach led to the derivation of a minimum of 25 kGy dose. In contrast, the North American entrepreneurs under the auspices of the Association for the Advancement of Medical Instruments (AAMI) challenged the rationale for the use of an atypical radioresistant indicator Bacillus pumilus as the reference standard for setting the minimum sterilising dose. Instead, the AAMI guidelines recommended the radiation-response criteria of natural bioburden microorganisms (as sampled from pre-sterile medical). Despite the AAMI guidelines and attendant stipulations in North America, until the present time the traditional dose requirement of 25 kGy for radiation sterilisation is most widely followed worldwide. Respective national health authorities demand concurrence with acceptable sterilisation processing for certification of end-products in safe clinical practice based upon a confirmation that the 25 kGy dose has been assured as the minimum. European Economic Community (EEC) countries have attempted to harmonise standards and criteria prevailing in the different countries and have formulated the EUCOMED guidelines, including the mandatory requirement of 25 kGy dose minimum. Currently, further internationalisation of the quality and safety regulatory criteria guidelines for requirements for validation and routine control of radiation-sterilised health-care products have come under the preview of the International Standards Organisation (ISO). Draft international standards criteria are put under scrutiny and periodic updating reviews for ultimate finalisation. The various elements of the ISO standards for a validation programme of a radiation sterilisation process are shown in Fig. 6.11. In a current developing standards document being considered for the international sterilisation community, under a section on "Dosimeters", it is specified that each batch of dosimeters to be used must be properly calibrated. This entails either (1) irradiation of a user' s dosimeter in a standards or accredited reference (secondary) laboratory, and subsequent appropriate evaluation by the user, (2) irradiation in a suitably designed irradiation geometry in the user's laboratory along with dosimeters issued by a standards or reference laboratory, or (3) use of a radiation field where the calibration is traceable to a standards laboratory, according to an acceptable accreditation procedure. Table 6.4 lists some categories of improved and future dosimetry systems, their analytical methods, and several examples of each. This list includes both reference and routine dosimeters, as well as some of those suitable for transfer dosimeters, all three of these being defined in Chapter 5 of McLaughlin et al. (1989).
298
Chapter 6
Product qualification Sterilization dose determination
Product and packaging materials evaluation
I I
Installation qualification Equipment documentation
,
Equipment testing
Equipment calibration
.
~
I 1
Irradiator dose mapping
....
Process qualification Determination of product loading pattern
Product dose mapping [
,,
~JP
,
Certification Documentation accumulation .
.
.
.
i~:
.
Review and approval 9 ~
-
~
'
:
9
"
::
:
::::
.......
::.
84184184184
Maintenance of validation Calibration program
Irradiator requalification
Sterilization dose auditing
Fig. 6.11. The elements of a typical validation program (from Draft ISO/DIS 11137.2; ISO/TC 198; dated March 1993).
Calorimetric devices have been designed especially for radiation processing applications as a means of calibrating electron beams and gamma-ray fields as well as high-dose dosimeters, and for routine on-line radiation processing by electron accelerator beams. These calorimeters include graphite discs (Humphreys and McLaughlin, 1990; Domen and Lamperti, 1974; Burns and Morris, 1988), Petri dishes containing water (Miller and Kovacs, 1985), or polystyrene in cylindrical or spherical geometry (Domen and Wei-Zhen, 1987). The calorimeters are generally relatively simple in geometry, and contain a radiation-resistant thermistor or a thermocouple that is well calibrated. In addition, the specific heat and its temperature dependence for the calorimetric absorber materials are well established.
Radiation Safer3,
299
Table 6.4 Important dosimetry systems for the future (after McLaughlin et al., 1989) Dosimeter type
Method of readout
Examples
Calorimeters
temperature measurement
graphite, water, polystyrene
Amino acids, cellulosics, sugars
EPR spectrometry
alanine, sucrose, cellulose
Diamonds
electrical measurement EPR spectrometry
diamond crystals and films
Inorganic crystals
spectrophotometry EPR spectrometry
LiF, SiO 2, Suprasil TM glass
Semiconductors
electrical measurement
diodes, MOSFETs
Chemical solutions
spectrophotometry spectrofluorimetry
ceric-cerous, organic acids, ethanolchlorobenzen e
Radiochromic films
spectrophotometry microdensitometry
dyed plastics, polydiacetylenes
Fluorescent systems
spectrofluorimetry spectrophotometry
inorganic and organic fluors
Dosimeters that are sufficiently small, such as thin radiochromic films and alanine pellets, can readily be calibrated against the calorimeter, by irradiating in tandem (with a suitable radiation monitor) while encased in a phantom material that is identical in size, shape and substance to the calorimetric absorber. The main appreciable correction that is required is the ratio of mass energy-absorption coefficients of the two materials (in the case of photon irradiations) or the ratio of mass collision stopping powers of the two materials (in the case of electron beam irradiations) (McLaughlin et al., 1989). One of the most promising dosimeters, which may under careful preparation and calibration, qualify as a reference dosimetry system, is L-t~ alanine as measured by EPR spectrometry (Regulla et al., 1993). It is also proving to be useful as a transfer dosimeters, as shown by its application in the IAEA International Dose Assurance Service (IDAS) (Nam and Regulla, 1989) and in the reference dosimetry service to industry by the National Physical Laboratory (NPL). Accident dosimetry using biological systems in which the quantification of chromosome aberrations or the ratios between different blood proteins can give an indication of exposure, is hampered by the individual characteristics of the victim (i.e. general health, diet etc.), and by the complexity of the techniques. These problems can be avoided by adopting a more physical approach, and both chemiluminescence and thermoluminescence of possible dosimeters, for example, have been found to be useful. The drawbacks here concern the solubility with chemiluminescence, the amount of sample required for thermoluminescence, and the impossibility of taking repeated measurements with either system. In contrast, electron spin resonance (ESR) spectroscopy is not subject to these constraints. Measurement is made directly on the sample, very small amounts of material can be used, and repeated measurements are possible
300
Chapter 6
because the method is non-destructive. ESR is a form of spectroscopy used to detect unpaired electrons by the absorption of microwave energy in a strong magnetic field. For a detailed description of the technique and its applications see Wertz and Bolton (1986). The interaction of ionising radiation with organic matter generates free radicals, which, in a solid matrix, may become trapped and hence are detectable by ESR. If the signals are sufficiently stable, and a direct relationship exists between radical concentration and radiation dose, the material may be useful as a dosimeter. The amino acid alanine has received considerable attention in this respect, due to the high stability of the CH3CHCOOH radical that is formed on irradiation, and an alanine dosimetry system based on the use of ESR spectroscopy is currently being developed in Japan. However, with the exception of those personnel who may be issued with such dosimeters in the future, the low probability of alanine being present on accident victims precludes this from being considered as accident dosimetry. One substance that can be regarded as an accident dosimeter is tooth enamel (hydroxyapatite), and this has recently been used in a re-evaluation of the exposure results from Hiroshima and Nagasaki (Okajima, 1985). The CO 3 centre is very stable and accurate measurements have been made 40 years after the event (Tatsumi-Miyajima, 1987). However, this approach is only applicable if human teeth can be readily obtained from accident victims now deceased or from patients in the normal course of dental treatment. ESR dosimetry of some construction materials at Hiroshima and Nagasaki was carried out (Ikeya and Ishii, 1989) to determine the A-bomb radiation dose. Some minerals exposed to low-level natural radiation over a given geological time period can also be used to determine the intense A-bomb radiation dose. Chandra and Symons (1987) have considered the use of human fingernail for measurements of free radical production resulting from radiation damage. Nakajima (1982) has investigated the dosimetric properties of Lucite, polyethylene, paper, wool, human hair and nail. However, following irradiation, the wool, hair and nail showed a rapid decay of the ESR signal intensity, with the hair giving a 50% reduction between the first and second observations, separated by a lapse of 5 min. Consequently it was concluded that these substances were unsuitable for dosimetry. In the work by Dalgamo and McClymont, 1989, materials were sought which could be used as accident dosimeters, i.e. solid, probably organic substances found on or about an accident victim and which could be used with ESR spectroscopy to give a rapid and reliable dose assessment. A number of materials have been shown to be potentially useful for ESR dosimetry, although in some cases only one component of a mixture is suitable. The pseudo-G value gives a useful indication of the sensitivity of a potential dosimeter. Such values should be based on a radiation dose that is known to be within the linear dose-response region. Samples containing sugars (e.g. sucrose and lactose) give a high sensitivity to radiation and the induced radicals are relatively stable. Fingernails, hair and leather look particularly promising for accidental irradiation, as information on the dose distribution can be obtained thereby facilitating medical treatment.
Radiation SafeO'
301
Fig. 6.12. A schematic illustration of the ESR dosimeter reader for the in-vivo human tooth measurement. The front molar of the lower jaw is attached to the end hole of TEl02 microwave cavity (after Ikeya and Ishii, 1989).
Ikeya and Ishii (1989) have separated design and manufacturing of an ESR cavity and a special NdBFe (Neomax) magnet system for in vivo measurement of the radiation dose of a human tooth without extraction (see Fig. 6.12). REFERENCES Bailey, M.R., New ICRP human respiratory tract model. Radiological Protection Bulletin, No. 144 (July 1993). Birchall, A., James, A.C., Jarvis, N.S., Bailey, M.R. and Dorrian, M-D., Implementation of the new ICRP human respiratory tract model: LUDEP 1.0. Chilton, NRPB-R264. Boecker, B. et al., Current status of bioassay procedures to detect and quantify previous exposures to radioactive materials. Health Phys., 60, Suppl. 1 (1991) 45-100. Brahme, A., Chavaudra, J., Landberg, T., McCullough, E.C., Nuesslin, F., Rawlinson, J.A., Svensson, G. and Svensson, H., Accuracy requirements and quality assurance of external beam therapy with photons and electrons. Acta Oncol. Suppl. 1 (1988) 7-76. Brodsky, A., Accuracy and detection limits for bioassay measurements in radiation protection--Statistical considerations, NUREG-1156, USNRC, Washington, D.C. (1986). Burns, D.T. and Morris, W.T., Recent developments in graphite and water calorimeters for electron beam dosimetry at NPL, Proc. Int. workshop on Water Calorimetry, Report NRC-29637, National Research Council of Canada, Ottawa, Canada (1988) 25-30. Chandra, H. and Symons, M.C.C., Sulphur radicals formed by cutting t~-keratin. Nature, 328 (1987) 833. Cross, W.G., AECL 2793 (1967). Da Rosa, L.A.R. and Nette, P., Thermoluminiscent dosimeters for exposure assessment in gamma or x-ray radiation fields with unknown spectral distribution. Appl. Radiat. Isot., 39 (1988) 191. Dalgarno, B.G. and McClymont, J.D., Evaluation of ESR as a radiation accident dosimetry technique, Appl. Radiat. Isot., 40 (1989) 1013. Domen, S.R. and Lamperti, P.J., A heat-loss compensated calorimeter: theory, design and performance. NBS J. Research 78A, National Institute of Standard and Technology, Gaithersburg, MD 20899, USA (1974) 595-612. Domen, S.R., Wei-Zhen, B.A., A polystyrene absorbed dose rate calorimeter. Nucl. Instr. Meth. Phys. Res., B24/25 (1987) 1054-1059. Eisenlohr, H.H., Haider, J.G. and Rud6n, B.I., International Postal Dose Intercomparison using TLD. In: A. Scharmann (ed.), Proc. 5th Int. Conf. on Luminescence Dosimetry, Sao Paulo, 1977. Justus Liebig University, Griessen, 1977, pp. 350-358.
302
Chapter 6
Goldstein, H. and Wilkins, J.E., Calculation of the penetration of gamma-rays, NYO-3075, Nuclear Development Associates Inc., 1954. Humphreys, J.C. and McLaughlin, W.L., Calorimetry of electron beams and the calibration of dosimeters at high doses. Radiat. Phys. Chem., 35 (1990) 744-749. Huntley, R.B. and Nette, H.P., International Atomic energy Agency/World Health Organisation, TLD radiotherapy dosimetry intercomparison. Austr. Phys. Eng. Sci. Med., 16 (1993) 44. IAEA, Radiation protection procedures, IAEA Safety Series No. 38, IAEA, Vienna, 1973. IAEA, Safe handling of radionuclides, 1973 Edition, IAEA Safety Series No. 1, IAEA, Vienna, 1973. ICRP, Limits for intakes of radionuclides by workers, Publication 30, Part 1; Ann. ICRP 2/3/4, 1979. ICRP, Individual monitoring for intakes of radionuclides by workers: Design and interpretation, Publication 54, Ann. ICRP 19(1-3), 1988. ICRP, Data for protection against ionising radiation from external sources, ICRP Publication 21, Pergamon Press, Oxford, 1976. ICRP, Handling and disposal of radioactive materials in hospitals and medical research establishments, ICRP Publication 5, Pergamon Press, Oxford, 1964. ICRP, Protection against ionising radiation from external sources used in medicine, ICRP Publication 33 (Annals of ICRP, 9 (1)), Pergamon Press, Oxford, 1982. ICRU, Determination of Dose Equivalents Resulting from External Radiation sources. Report 39, ICRU Publications, Bethesda, 1985. ICRU, Radiation dosimetry: Electron beams with energies between 1 and 50 MeV. Report 35, International Commission on Radiological Units and Measurements, Washington, DC, 1984. Ikeya, M. and Ishii, H., Atomic bomb and accident dosimetry with ESR: Natural rocks and human tooth in vivo spectrometer. Appl. Radiat. Isot. ARISE, 40 (1989) 1021-1027; INIS ATOMINDEX (1990) 21: 04O585. International Commission on Radiological Protection. Human respiratory tract model for radiological protection, ICRP Publication 66. Ann. ICRP. 24, Nos. 1-4 (1993). International Commission on Radiological Protection. Limits for intakes or radionuclides by workers. ICRP Publication 30. Part 1: Ann. ICRP, 2, Nos. 3-4 (1979); Part 2: Ann. ICRP. 4, Nos. 3--4 (1980); Part 3: Ann. ICRP, 6, Nos. 2-3 (1981); Part 4: Ann. ICRP, 19, No. 4 (1988). International Commission on Radiological Protection; Reference Man: Anatomical, Physiological and Metabolic characteristics; ICRP Publication 23; Pergamon Press, ISSN 0 08 017024 2, 1975. ISO. X and Gamma Reference Radiations for Calibrating Dosemeters and Dose Ratemeters. ISO 4037 (1979). Jarvis, N.S. and Birchall, A., LUDEP, 1.0, A Personal computer program to implement the new ICRP respiratory tract model. Radiation Prot. Dosim., 53 (1-4) (1994) 191-193. Johnsson, K.-A., Hanson, V.W. and Horiot, J.C., Workshop of the EORTC Radiotherapy Group on Quality Assurance in Cooperative Trials of Radiotherapy: A Recommendation for EORTC Cooperative Groups. Radiother. Concol., 11 (1988) 201-203. McLaughlin, W.L., Boyd, A.W., Chadwick, K.H., McDonald, J.C. and Miller, A., Dosimetry for Radiation Processing. Taylor and Francis, London, 1989. Miller, A. and Kov~ics, A., Calorimetry at industrial electron accelerators. Nucl. Instr. Meth. Phys. Res., B10/11 (1985) 994-997. Nakajima, T., External dose to a Japanese tourist from the Chernobyl reactor accident. Health Phys., 53 (1987) 405. Nakajima, T., The use of organic substances as emergency dosimeters. Int. J. Appl. Radiat. Isot., 33 (1982) 1077. Nam, J.W. and Regulla, D.F., The significance of the International Dose Assurance Service for radiation processing. Appl. Radiat. Isot., 40 (1989) 953-955. NCRP, Use of bioassay procedures for assessment of internal radionuclide deposition. NCRP, Bethesda, 1987. Nette, P., Alenikov, V., Griffith, R. and da Silva, T., The possible role of the IAEA/WHO SSDL network in implementing the dose equivalent operational quantities into radiation protection practices for individual monitoring. Radiation Prot. Dosim., 28 (1989) 161-165.
Radiation SafeO'
303
Okajima, S., Reassessment of A-bomb dosimetry, in: M. Ikeya and T. Miki, T. (eds.), 1st Int. Symp on ESR Dating, Tokyo, p. 381. IONICS, Tokyo, 1985. Petoussi, N., Zaukl, M., Pauzer, W., Drexler, G. and Nette, P., Photon spectra in standard dosimetric or imaging photons calculated with Monte Carlo methods. Radiation Prot. Dosim., 43 (1992) 147. Regulla, D.F., Bartolotta, A., Definer, U., Onori, S., Pantaloni and M., Wieser, A., A calibration network based on alanine/ESR dosimetry. Appl. Radiat. Isot., 44 (1993) 23-31. Rikner, G. and Grusell, E., General specifications for silicon semiconductors for use in radiation dosimetry. Phys. Med. Biol., 32(9) (1987a) 1109-1117. Rikner, G. and Grusell, E., Patient dose measurements in photon fields by means of silicon semiconductor detectors. Med. Phys., 14(5) (1987b) 870-873. Ryufuku, H. et al., Evaluation of beta-ray skin dose based on point kernel method (in Japanese), JAERI-M-7354, 1977. Svenson, H., Nette, P. and Zsdanszky, K., Clinical applications of solid state dosimetry. Radiation Prot. Dosim., 34 (1990) 241. Tatsumi-Miyajima J., ESR dosimetry for atomic bomb survivors and radiological technologists. Nucl. Instrum. Meth. Phys. Res., A257 (1987) 417. U.S. Department of Health, Education, and Welfare: Radiological health hand box, 7th Edition, Consumer Protection and Environmental Health Service, Rockville, 1970. Wertz, J.E. and Bolton, J.R., Electron Spin Resonance. Chapman and Hall, New York, 1986. World Health Organisation, A Rational Approach to Radiodiagnostic Investigations, Technical Report Series No. 689. WHO, Geneva, 1983. World Health Organisation, Effective Choices for Diagnostic Imaging in Clinical Practices, Technical Report Series No. 795. WHO, Geneva, 1990. World Health Organisation, Rational Use of Diagnostic Imaging in Pediatrics, Technical Report Series No. 757. WHO, Geneva, 1987.
305
CHAPTER 7
The Nuclear Fuel Cycle
7.1 I N T R O D U C T I O N This subject is a part of a broader subject which could be called "a human need for energy". According to the World Energy Council (WEC), the demand for energy is expected to increase further. The reasons for this are clear: 50% of the countries of the world are in the process of industrialization with economies growing rapidly, some at 10% or more. This calls for an ever-increasing supply of coal, oil and now natural gas. China, for instance, which sustained a 10% growth rate in its market economy for over a decade, burns 1.2 Gt of coal a year (1996 figures) and expects to burn 1.4 Gt of coal a year after the year 2000. In addition, it imports oil, has an embryonic nuclear industry, and is seeking to diversify its renewable energy supplies beyond hydroelectric power. The global situation is made more challenging by the expected rapid increase in the world population. UN figures suggest that the current population of nearly 6 billion will increase to over 10 billion by 2050; 80% of that population will be in developing countries where, to put matters into perspective, 50% of the population do not have an electricity supply connection (neither do they have safe drinking water) so their demand for the services provided by energy and an improved life-style is undeniable (Fells, 1998). There are three energy sources, all of which have enormous potential, which might give us confidence that the world' s energy needs will be met in the second half of the 21st century. These are nuclear energy, renewable energy and nuclear fusion. They have the advantage that they are also environmentally clean and do not emit large quantities of greenhouse gases. Nuclear fusion has enormous potential, but will be very difficult to achieve from an engineering point of view. It is prudent to rely, for the time being at any rate, on the growth in renewable and nuclear energy. They are not, incidentally, mutually exclusive but complementary, and it is very difficult to imagine any future scenario post-2050 without a large slice of energy from both sources. Nuclear power, in particular, plays an important role in controlling carbon-dioxide emissions.
306
Chapter 7
The nuclear future may well lie with the fast breeder reactor which uses uranium some 60 times more efficiently than current fission reactors. Prototype breeder reactors are in operation in Russia and Japan and were so until recently in the UK and France. They will be required post-2030 or so if a major new nuclear programme is embarked on, as uranium resources are predicted to start running into short supply at about that time. The problem with the continuing growth in nuclear power is the public' s perception of its safety, particularly the safety of radioactive waste disposal. If the spent nuclear fuel is reprocessed, as happens in the UK and France, and the high-level radioactive waste glassified and stored in metal cylinders in a dry rock store, there is general international consensus that it will be very safety contained. And the method of reprocessing recovers plutonium as well as unused uranium, both of which can be recycled as fuel for fission reactors. The other method of dealing with spent nuclear fuel is to merely store it either in cooling ponds or directly in a dry store. This presents a less tidy legacy for the future but is the preferred method in some countries, fearful that reprocessing will lead to proliferation (Fells, 1998).
7.2 THE STATUS OF N U C L E A R POWER IN THE WORLD The concept of a nuclear fuel cycle is an old one, almost dating back to the concept of controlled nuclear fission to generate electricity. At the time of the development of the first nuclear power plants, it was generally taken for granted that fuel from power reactors would be reprocessed and that the recovered uranium and plutonium would be recycled. In those days, uranium ore was a scarce and expensive commodity and it was naturally assumed that economically available supplies would not meet the demands required by a widespread use of nuclear power. Consequently, the extraction of all the potential energy content of uranium-235, seemed to be essential. Such a complete exploitation of uranium resources requires reprocessing of the spent fuel and the extraction of plutonium for burning in specially designed "fast" reactors. The approach became more attractive with the concept of fast breeder reactors, which could produce more fuel than they consumed. For such reasons, many countries during the 1960s attached high priority to the development of fast reactors, and it was anticipated that they would be widely deployed in the 1980s (Semenov and Oi, 1993). Until the early 1970s then, the nuclear fuel cycle was pictured as an orderly sequence of processes. It began with uranium mining, milling, and conversion, was followed by fuel enrichment, fuel fabrication, and power generation, and was finally completed by reprocessing, recycling of plutonium and uranium to fast reactors, and final disposal of waste streams from reprocessing plants. In essence, closure of the fuel cycle meant the effective use of plutonium. Three different types of fuel cycle are commonly identified for nuclear power generation, depending on whether fuel is recycled and on the type of reactor used for electricity production.
The Nuclear Fuel Cycle
307
9 The " o n c e - t h r o u g h " f u e l cycle: in this cycle, the spent fuel is not reprocessed but
kept in storage until it is eventually disposed of as waste. 9 The t h e r m a l r e a c t o r cycle: in this cycle, the spent fuel is reprocessed and the
uranium and plutonium can be recycled in new fuel elements. It is also possible to recycle only the uranium and to store the plutonium, and vice versa. 9 The f a s t b r e e d e r reactor cycle: in this cycle, the spent fuel is similarly reprocessed and the uranium and plutonium fabricated into new fuel elements. However, they are recycled to fast breeder reactors, in which there is a central core of uranium/plutonium fuel surrounded by a blanket of depleted uranium (uranium from which most of the uranium-235 atoms have been removed during the process of enrichment) or to burner reactors. This depleted uranium consists mostly of uranium-238 atoms, some of which are converted to plutonium during irradiation. By suitable operation, fast breeder reactors thus can produce slightly more fuel than they consume, hence the name "breeder" (see Fig. 7.1). The situation has changed dramatically during the last 20 years. No closed fuel cycle of the type originally envisaged to be operational in the 1980s exists today. Although the closure of the nuclear fuel cycle has been experimentally demonstrated in France, Japan, Russia and the United Kingdom, it has not been demonstrated yet on a commercial scale. Current thinking is divided into two schools. One believes that plutonium as an energy source has no economic value and spent fuel should be disposed of in a safe way
Fig. 7.1. Schematic presentations of (a) the once-throughcycle (b) the thermal reactor cycle and (c) the fast reactor cycle. U30s = yellowcake, U F 6 = uranium hexafluoride, MOX = mixed oxide fuel (uranium/ plutonium) (after Semenov and Oi, 1993).
Chapter 7
308
Table 7.1 Nuclear power status in the world Under construction
In operation No. of units
Total net MWE
Argentina
2
935
Armenia
1
376
Belgium
7
5712
Brazil
1
626
Bulgaria
6
3 538
Canada
16
11 994
China
3
2 167
China, Taiwan
6
4 884
Czech Republic
4
1 648
Finland
4
2 455
France
59
62 853
Germany
20
22 282
4
1 729
10
1 695
Hungary India
No. of units
692
1 245
2 155
1 824
1 450
808 2111
Iran Japan Kazakstan Korea, Republic of
Total net MWE
54 1
796
43 850 70
5 210
12
9 770
Lithuania
2
2 370
Mexico
2
1 308
Netherlands
1
449
Pakistan
1
125
300
650
650
19 843
3 375
Romania Russian Federation
1 29
South Africa
2
Slovak Republic
4
1 632
Slovenia
1
632
9
7 320
Spain Sweden Switzerland
12 5
1 842 1 552
10 040 3 079
United Kingdom
35
12 928
Ukraine
16
13 765
United States
107
99 188
World total
437
351 795
3 800
35
25 878
The Nuclear Fuel Cycle
309
Fig. 7.2. A schematic of the nuclear fuel cycle.
(the "once-through" option). The other essentially adheres to the traditional nuclear fuel cycle (closed cycle option as illustrated in Fig. 7.2). The difference of opinions stems mainly from the predictions of nuclear electricity growth and the predicted availability of economical supplies of uranium, although it is influenced by political and environmental issues as well. We shall return to this matter again. The present situation on the use of nuclear energy for electricity production is illustrated in Table 7.1 which lists power stations in individual countries. This is the status as of March 1998 as reported by IAEA. All these stations can be grouped by reactor type as shown in Table 7.2. Once a year Nuclear News publishes the world list of nuclear power plants which are operable, under construction, or an order (for power 30 MWe and over). From the March 1997 issue of Nuclear News we list the world power stations, as shown in Table 7.3.
Table 7.2 Nuclear power units by reactor type, worldwide Reactor type
In operation
Total
No. units
Net MWe
No. units
Net MWe
250
220 422
286
253 518
Boiling light-water reactors (BWR)
94
78 285
100
85 605
Gas-cooled reactors; all types
35
11 699
35
11 699
Heavy-water reactors; all types
36
19 377
52
27 647
Graphite-moderated light-water reactors (LGR)
15
14 785
16
15 710
3
928
7
3 308
Pressurized light-water reactors (PWR)
Liquid-metal-cooled fast-breeder reactors (LMFBR)
310
Chapter 7
Table 7.3 World list of nuclear power plants Net MWe
Type
Commision Nacional de Atucha 1 (Lima, Buenos Aires) Energia Atomica Atucha 2 (Lima, Buenos Aires) (CNEA) Embalse (Rio Tercero, Cordoba)
335 692 600
PHWR PHWR PHWR
Ministry of Nuclear Power
Armenia 2 (Metsa, pr. Armenia)
400
PWR
Belgium
Electrabel
Doel 1 (Doel, East Flanders) Doel 2 (Doel, East Flanders) Doel 3 (Doel, East Flanders) Doel 4 (Doel, East Flanders) Tihange 1 (Huy, Liege) Tihange 2 (Huy, Liege) Tihange 3 (Huy, Liege)
392 392 970 1001 863 894 1015
PWR PWR PWR PWR PWR PWR PWR
Brazil
Furnas Centrais Electricas SA
Angra 1 (Itaorna, Rio de Janeiro) Angra 2 (Itaorna, Rio de Janeiro) Angra 3 (Itaorna, Rio de Janeiro)
626 1229 1229
PWR PWR PWR
Bulgaria
National Electric Co.
Kozloduy 1 (Kozloduy, Vratsa) Kozloduy 2 (Kozloduy, Vratsa) Kozloduy 3 (Kozloduy, Vratsa) Kozloduy 4 (Kozloduy, Vratsa) Kozloduy 5 (Kozloduy, Vratsa) Kozloduy 5 (Kozloduy, Vratsa)
400 400 400 400 910 910
PWR PWR PWR PWR PWR PWR
Canada
New Brunswick Power Corp. Ontario Hydro
Point Lepreau (Buy of Fundy, N.B.)
640
PHWR
Pickering 1 (Pickering, Ont.) Pickering 2 (Pickering, Ont.) Pickering 3 (Pickering, Ont.) Pickering 4 (Pickering, Ont.) Pickering 5 (Pickering, Ont.) Pickering 6 (Pickering, Ont.) Pickering 7 (Pickering, Ont.) Pickering 8 (Pickering, Ont.) Bruce 1 (Tiverton, Ont.) Bruce 1 (Tiverton, Ont.) Bruce 1 (Tiverton, Ont.) Bruce 1 (Tiverton, Ont.) Bruce 1 (Tiverton, Ont.) Bruce 1 (Tiverton, Ont.) Bruce 1 (Tiverton, Ont.) Bruce 1 (Tiverton, Ont.) Darlington 1 (Newcastle Twp., Ont.) Darlington 2 (Newcastle Twp., Ont.)
515 515 515 515 516 516 516 516 769 769 769 769 860 860 860 860 881 881
PHWR PHWR PHWR PHWR PHWR PHWR PHWR PHWR PHWR PHWR PHWR PHWR PHWR PHWR PHWR PHWR PHWR PHWR
Country
Authority
Argentina
Armenia
Power Station
The Nuclear Fuel Cycle
Country
China
311
Net MWe
Type
Authority
Power Station
Hydro Quebec
Darlington 3 (Newcastle Twp., Ont.) Darlington 4 (Newcastle Twp., Ont.) Gentilly 2 (Becancour, Que.)
881 881 635
PHWR PHWR PHWR
Qinshan 1 (Haiyan, Zhejiang) Qinshan 2 (Haiyan, Zhejiang) Qinshan 3 (Haiyan, Zhejiang) Qinshan 4 (Haiyan, Zhejiang) Qinshan 5 (Haiyan, Zhejiang) Guangdong 1 (Shenzhen, Guangdong) Guangdong 2 (Shenzhen, Guangdong)
300 600 600 700 700 900 900
PWR PWR PWR PHWR PHWR PWR PWR
Lingao 1 (Lingao, Guangdong) Lingao 2 (Lingao, Guangdong)
985 985
PWR PWR
Ministry of Nuclear Industry
Guangdong Nuclear Power Joint Venture Co. Ltd. Lingao Nuclear Co.
Cuba
Ministry of Basic Industries
Juragua 1 (Cienfuegos, Cienfuegos) Juragua 2 (Cienfuegos, Cienfuegos)
417 417
PWR PWR
Czech Republic
Czech Power Board
Dukovany 1 (Trebic, Jihomoravsky)
408
PWR
Dukovany 2 (Trebic, Jihomoravsky) Dukovany 3 (Trebic, Jihomoravsky) Dukovany 4 (Trebic, Jihomoravsky) Temelin 1 (Temelin, Jihocesky) Temelin 2 (Temelin, Jihocesky)
408 408 408 890 890
PWR PWR PWR PWR PWR
Finland
Imatran Voima Oy (IVO) Teollisuuden Voima Oy (TVO)
Loviisa 1 (Loviisa, UUsimaa) Loviisa 2 (Loviisa, UUsimaa) TVO 1 (Olkiluoto, Turku-Pori) TVO 2 (Olkiluoto, Turku-Pori)
445 445 710 710
PWR PWR PWR PWR
France
Commissariat a l'Energie Atomique Centrale Nucleaire Europeene a Neutrons Rapides S.A. (NERSA) Electricite de France (EdF)
Phenix (Marcoule, Gard)
233
LMFBR
Creys-Malville (Bouvesse, Isere)
600
LMFBR
Chinon B 1 (Chinon, Indre-et-Loire) Chinon B2 (Chinon, Indre-et-Loire) Chinon B3 (Chinon, Indre-et-Loire) Chinon B4 (Chinon, Indre-et-Loire) Saint-Laurent B 1 (Saint-Laurent-des-Eaux, Loir-et-Cher) Saint-Laurent B2 (Saint-Laurent-des-Eaux, Loir-et-Cher) Bugey 2 (Loyettes, Ain) Bugey 3 (Loyettes, Ain) Bugey 4 (Loyettes, Ain) Bugey 5 (Loyettes, Ain)
905 905 905 905 915
PWR PWR PWR PWR PWR
915
PWR
910 910 880 880
PWR PWR PWR PWR continued
312
Chapter 7
Table 7.3 (continuation) Country
Authority
Power Station
Net MWe
Type
Fessenheim 1 (Fessenheim, Haut-Rhin) Fessenheim 2 (Fessenheim, Haut-Rhin) Dampierre 1 (Ouzouer, Loiret) Dampierre 2 (Ouzouer, Loiret) Dampierre 3 (Ouzouer, Loiret) Dampierre 4 (Ouzouer, Loiret) Gravelines B 1 (Gravelines, Nord) Gravelines B2 (Gravelines, Nord) Gravelines B3 (Gravelines, Nord) Gravelines B4 (Gravelines, Nord) Gravelines B5 (Gravelines, Nord) Gravelines B6 (Gravelines, Nord) Tricastin 1 (Pierrelatte, Drome) Tricastin 2 (Pierrelatte, Drome) Tricastin 3 (Pierrelatte, Drome) Tricastin 4 (Pierrelatte, Drome) Blayais 1 (Blaye, Gironde) Blayais 2 (Blaye, Gironde) Blayais 3 (Blaye, Gironde) Blayais 4 (Blaye, Gironde) Paluel 1 (Veulettes, Seine-Maritime) Paluel 2 (Veulettes, Seine-Maritime) Paluel 3 (Veulettes, Seine-Maritime) Paluel 4 (Veulettes, Seine-Maritime) Cruas 1 (Cruas, Ardeche) Cruas 2 (Cruas, Ardeche) Cruas 3 (Cruas, Ardeche) Cruas 4 (Cruas, Ardeche) Saint-Alban 1 (Auberives, lsere) Saint-Alban 2 (Auberives, Isere) Flamanville 1 (Flamanville, Manche) Flamanville 2 (Flamanville, Manche) Cattenom 1 (Cattenom, Moselle) Cattenom 2 (Cattenom, Moselle) Cattenom 3 (Cattenom, Moselle) Cattenom 4 (Cattenom, Moselle) Belleville 1 (Belleville s/Loire, Cher) Belleville 2 (Belleville s/Loire, Cher) Nogent s/Seine 1 (Nogent s/Seine, Aube) Nogent s/Seine 2 (Nogent s/Seine, Aube) Penly 1 (Saint-Martin-en-Campagne, Seine-Maritime) Penly 2 (Saint-Martin-en-Campagne, Seine-Maritime) Golfech 1 (Valence, Tarn et Garonne) Golfech 1 (Valence, Tarn et Garonne) Chooz B 1 (Chooz, Ardennes)
880 880 890 890 890 890 910 910 910 910 910 910 915 915 915 915 910 910 910 910 1330 1330 1330 1330 915 915 915 915 1335 1335 1330 1330 1300 1300 1300 1300 1310 1310 1310 1310 1330
PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR
1330
PWR
1310 1310 1455
PWR PWR PWR
The Nuclear Fuel Cycle
Country
Germany
Hungary
Authority
313
Power Station
Net MWe
Type
Chooz B2 (Chooz, Ardennes) Civaux 1 (Civaux, Vienne) Civaux 1 (Civaux, Vienne)
1455 1450 1450
PWR PWR PWR
1275
PWR
1360
PWR
785 1269
PWR PWR
1326
PWR
771
BWR
870 1330 1260
BWR PWR BWR
1290
PWR
340
PWR
890 1358 1284 1288
BWR PWR BWR BWR
630
PWR
1285
PWR
1167 1240 1219
PWR PWR PWR
430 430 430 430
PWR PWR PWR PWR
Bayernwerk AG
Grafenrheinfeld KKG (Gragenrheinfeld, Ba.) Gemeinschaftskernkraft Grohnde (Emmerthal, Nied.) werk Grohnde GmbH (KWG) GemeinschaftskernNeckar 1 (Neckarwestheim, B.-W.) kraftwerk Neckar Neckar 1 (Neckarwestheim, B.-W.) (GKN) Kernkraftwerk Brokdorf (Brokdorf, S.-H.) Brokdorf GmbH (KBR) Kernkraftwerk Brunsbuettel (Brunsbuettel, S.-H.) Brunsbuettel GmbH (KKB) Kernkraftwerk Isar Isar 1 (Essenbach, Ba.) (KKI) Isar 2 (Essenbach, Ba.) Kernkraftwerk Kruemmel (Geesthacht, S.-H.) Kruemmel Gmbh Hochtief/Hammers/Heitkamp/Holzmann (KKK) Kernkraftwerke Emsland (Lingen, Nied.) Lippe-Ems GmbH (KKE) Kernkraftwerk Obrigheim (Obrigheim, B.-W.) Obrigheim GmbH (KWO) Kernkraftwerk Philippsburg 1 (Philippsburg, B.-W.) Philippsburg (KKP) Philippsburg 2 (Philippsburg, B.-W.) Kernkraftwerk Gundremmingen B (Gundremmingen, Ba.) RWE-Bayernwerk Gundremmingen C (Gundremmingen, Ba.) GmbH (KRB) Kernkraftwerk Stade Stade (Stade, Nied.) GmbH (KKS) Kernkraftwerk Unterweser (Rodenkirchen, Nied.) Unterweser GmbH (KKU) RWE Energie Biblis A (Biblis, Hessen) Aktiengesellshaft Biblis A (Biblis, Hessen) Muelheim-Kaerlich (Muelheim-Kaerlich, R.-P.) Hungarian Power Companies, Ltd.
Paks Paks Paks Paks
1 (Paks, 2 (Paks, 3 (Paks, 4 (Paks,
Tolna) Tolna) Tolna) Tolna)
continued
314
Chapter 7
Table 7.3 (continuation) Country
Authority
Power Station
Net MWe
Type
India
Atomic Energy Commission, Department of Atomic Energy
Tarapur 1 (Tarapur, Maharashtra) Tarapur 2 (Tarapur, Maharashtra) Tarapur 3 (Tarapur, Maharashtra) Tarapur 4 (Tarapur, Maharashtra) Rajasthan 1 (Kota, Rajasthan) Rajasthan 2 (Kota, Rajasthan) Rajasthan 3 (Kota, Rajasthan) Rajasthan 4 (Kota, Rajasthan) Madras 1 (Kalpakkam, Tamil Nadu) Madras 2 (Kalpakkam, Tamil Nadu) Narora 1 (Narora, Uttar Pradesh) Narora 2 (Narora, Uttar Pradesh) Kakrapar 1 (Kakrapar, Gujarat) Kakrapar 2 (Kakrapar, Gujarat) Kaiga 1 (Kaiga, Karnataka) Kaiga 2 (Kaiga, Karnataka)
150 150 470 470 90 187 202 202 155 155 202 202 202 202 202 202
BWR BWR PWHR PWHR PWHR PWHR PWHR PWHR PHWR PHWR PHWR PHWR PHWR PHWR PHWR PHWR
Japan
Chubu Electric Power Co.
Hamaoka 1 (Hamaoka-cho, Shizuoka) Hamaoka 2 (Hamaoka-cho, Shizuoka) Hamaoka 3 (Hamaoka-cho, Shizuoka) Hamaoka 4 (Hamaoka-cho, Shizuoka) Hamaoka 5 (Hamaoka-cho, Shizuoka) Shimane 1 (Kashima-cho, Shimane) Shimane 2 (Kashima-cho, Shimane) Tomari 1 (Tomari-mura, Hokkaido) Tomari 2 (Tomari-mura, Hokkaido) Shika 1 (Shika-machi, Ishikawa)
515 806 1056 1092 1380 439 790 550 550 513
BWR BWR BWR BWR BWR BWR BWR PWR PWR BWR
Tokai 1 (Tokai Mura, Ibaraki) Tokai 2 (Tokai Mura, Ibaraki) Tsuruga 1 (Tsuruga, Fukui) Tsuruga 2 (Tsuruga, Fukui) Mihama 1 (Mihama-cho, Fukui) Mihama 2 (Mihama-cho, Fukui) Mihama 3 (Mihama-cho, Fukui) Takahama 1 (Takahama-cho, Fukui) Takahama 2 (Takahama-cho, Fukui) Takahama 3 (Takahama-cho, Fukui) Takahama 4 (Takahama-cho, Fukui) Ohi 1 (Ohi-cho, Fukui) Ohi 2 (Ohi-cho, Fukui) Ohi 3 (Ohi-cho, Fukui) Ohi 4 (Ohi-cho, Fukui) Genkai 1 (Genkai, Saga) Genkai 2 (Genkai, Saga) Genkai 3 (Genkai, Saga) Genkai 4 (Genkai, Saga)
159 1056 341 1115 320 470 780 780 780 830 830 1120
GCR BWR BWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR
Chugoku Electric Power Co., Inc. Hokkaido Electric Power Co. Hokuriku Electric Power Co. Japan Atomic Power Co. Ltd.
Kansai Electric Power Co., Inc.
Kyushu Electric Power Co., Inc.
1120 1127
1127 529 529 1127 1127
315
The Nuclear Fuel Cycle
Country
Authority
Power Station Sendai 1 (Sendai, Kagoshima) Sendai 2 (Sendai, Kagoshima) Fugen ATR (Tsuruga, Fukui)
Power Reactor & Nuclear Fuel Development Corp. Monju FBR (Tsuruga, Fukui) (PNC) Shikoku Electric Power Ikata 1 (Ikata-cho, Ehime) Co. Ikata 2 (Ikata-cho, Ehime) Ikata 3 (Ikata-cho, Ehime) Tohoku Electric Power Higashidori 1 (Higashidori, Aomori) Co., Inc. Onagawa 1 (Onagawa, Miyagi) Onagawa 2 (Onagawa, Miyagi) Onagawa 3 (Onagawa, Miyagi) Tokyo Electric Power Fukushima Daiichi 1 (Ohkuma, Co. Fukushima) Fukushima Daiichi 2 (Ohkuma, Fukushima) Fukushima Daiichi 3 (Ohkuma, Fukushima) Fukushima Daiichi 4 (Ohkuma, Fukushima) Fukushima Daiichi 5 (Ohkuma, Fukushima) Fukushima Daiichi 6 (Ohkuma, Fukushima) Fukushima Daini 1 (Naraha, Fukushima) Fukushima Daini 2 (Naraha, Fukushima) Fukushima Daini 3 (Naraha, Fukushima) Fukushima Daini 4 (Naraha, Fukushima) Kashiwazaki Kariwa 1 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 2 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 3 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 4 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 5 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 6 (Kashiwazaki, Niigata) Kashiwazaki Kariwa 7 (Kashiwazaki, Niigata) Kazakhstan
Kazakh State Atomic BN-350 (Aktau, Mangyshlak) Power Engineering and Industry Corp. (KATEP)
Net MWe 846 846 148
Type PWR PWR HWLW R
280
LMFBR
538 538 846 1100 497 796 825 39
PWR PWR PWR BWR BWR BWR BWR BWR
760
BWR
760
BWR
760
BWR
760
BWR
1067
BWR
1067 1067 1067 1067 1067
BWR BWR BWR BWR BWR
1067
BWR
1067
BWR
1067
BWR
1067
BWR
1315
BWR
1315
BWR
135
LMFBR
continued
316
Chapter 7
Table 7.3 (continuation) Country
Authority
Power Station
Net MWe
Type
Korea
Korea Electric Power Corp.
Kori 1 (Kori, Kyongnam) Kori 2 (Kori, Kyongnam) Kori 3 (Kori, Kyongnam) Kori 4 (Kori, Kyongnam) Wolsong 1 (Kyongju, Kyongbuk) Wolsong 2 (Kyongju, Kyongbuk) Wolsong 3 (Kyongju, Kyongbuk) Wolsong 4 (Kyongju, Kyongbuk) Yonggwang 1 (Yonggwang, Chonnam) Yonggwang 2 (Yonggwang, Chonnam) Yonggwang 3 (Yonggwang, Chonnam) Yonggwang 4 (Yonggwang, Chonnam) Yonggwang 5 (Yonggwang, Chonnam) Yonggwang 6 (Yonggwang, Chonnam) Ulchin 1 (Ulchin, Kyongbuk) Ulchin 2 (Ulchin, Kyongbuk) Ulchin 3 (Ulchin, Kyongbuk) Ulchin 4 (Ulchin, Kyongbuk) Ulchin 5 (Ulchin, Kyongbuk) Ulchin 6 (Ulchin, Kyongbuk)
556 605 895 895 629 650 650 650 900 900 950 950 950 950 920 920 950 950 950 950
PWR PWR PWR PWR PHWR PHWR PHWR PHWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR PWR
Lithuania
Ministry of Energy
Ignalina 1 (Visaginas) Ignalina 2 (Visaginas)
1380 1380
LGR LGR
Mexico
Comision Federal de Electricidad
Laguna Verde 1 (Laguna Verde, Veracruz) Laguna Verde 2 (Laguna Verde, Veracruz)
654 654
BWR BWR
Netherlands
NV Gemeenschappelijke Kernenergiecentrale Nederland NV Elektriciteits Produktiemaatschappij Zuid-Nederland (NV EPZ)
Dodewaard (Dodewaard, Gelderland)
55
BWR
452
PWR
Borssele (Borssele, Zeeland)
Pakistan
Pakistan Atomic Energy Kanupp (Karachi, Sind) Commission Chasnupp (Mianwali, Punjab)
Romania
Romanian Electricity Authority (RENEL)
Cernavoda Cernavoda Cernavoda Cernavoda Cernavoda
Russia
Ministry of Atomic Power
Balakovo 1 (Balakovo, Saratov) Balakovo 2 (Balakovo, Saratov)
1 (Cernavoda, 2 (Cernavoda, 3 (Cernavoda, 4 (Cernavoda, 5 (Cernavoda,
Constanta) Constanta) Constanta) Constanta) Constanta)
125 300
PHWR PWR
706 620 620 620 620
PHWR PHWR PHWR PHWR PHWR
950 950
PWR PWR
The Nuclear Fuel Cycle
Country
Slovakia
Authority
Slovak Power Board
317
Power Station
Net MWe
Type
Balakovo 3 (Balakovo, Saratov) Balakovo 4 (Balakovo, Saratov) Beloyarskiy 3 (BN-600) (Zarechnyy, Sverdlovsk) Kalinin 1 (Udomlya, Tver) Kalinin 2 (Udomlya, Tver) Kalinin 3 (Udomlya, Tver) Kola 1 (Polyarnyye Zori, Murmansk) Kola 2 (Polyarnyye Zori, Murmansk) Kola 3 (Polyarnyye Zori, Murmansk) Kola 4 (Polyarnyye Zori, Murmansk) Kursk 1 (Kurchatov, Kursk) Kursk 2 (Kurchatov, Kursk) Kursk 3 (Kurchatov, Kursk) Kursk 4 (Kurchatov, Kursk) Kursk 5 (Kurchatov, Kursk) Leningrad 1 (Sosnovyy Bor, St. Petersburg) Leningrad 2 (Sosnovyy Bor, St. Petersburg) Leningrad 3 (Sosnovyy Bor, St. Petersburg) Leningrad 4 (Sosnovyy Bor, St. Petersburg) Novovoronezhskiy 3 (Novovoronezhskiy, Voronezh) Novovoronezhskiy 4 (Novovoronezhskiy, Voronezh) Novovoronezhskiy 5 (Novovoronezhskiy, Voronezh) Smolensk 1 (Desnogorsk, Smolensk) Smolensk 2 (Desnogorsk, Smolensk) Smolensk 3 (Desnogorsk, Smolensk) South Urals 1 (Chelyabinsk, Chelyabinsk) South Urals 2 (Chelyabinsk, Chelyabinsk) Vk-50 (Dimitrovgrad, Ulyanovsk, RSFSR)
950 950 560
PWR PWR LMFBR
950 950 950 411 411 411 411 925 925 925 925 925 925
PWR PWR PWR PWR PWR PWR PWR LGR LGR LGR LGR LGR LGR
925
LGR
925
LGR
925
LGR
385
PWR
385
PWR
950
PWR
925 925 925 750 750 50
LGR LGR LGR LMFBR LMFBR PWR
Bohunice 1 (Trnava, Zapadoslovensky) Bohunice 1 (Trnava, Zapadoslovensky) Bohunice 1 (Trnava, Zapadoslovensky) Bohunice 1 (Trnava, Zapadoslovensky) Mochovce 1 (Mochovce, Zapadoslovensky) Mochovce 2 (Mochovce, Zapadoslovensky) Mochovce 3 (Mochovce, Zapadoslovensky)
408 408 408 408 412
PWR PWR PWR PWR PWR
412
PWR
420
PWR continued
318
Chapter 7
Table 7.3 (continuation) Country
Authority
Power Station
Net MWe
Type
Mochovce 4 (Mochovce, Zapadoslovensky) Krsko (Krsko, Vrbina)
420
PWR
620
PWR
South Africa ESKOM
Koeberg 1 (Melkbosstrand, Cape) Koeberg 2 (Melkbosstrand, Cape)
920 920
PWR PWR
Spain
Asco 1 (Asco, Tarragona) Asco 2 (Asco, Tarragona) Vandellos 2 (Vandellos, Tarragona)
917 936 961
PWR PWR PWR
Trillo 1 (Trillo, Guadalajara) Almaraz 1 (Almaraz, Caceres) Almaraz 2 (Almaraz, Caceres) Santa Maria de Garona (Santa Maria de Garona, Burgos) Cofrentes (Cofrentes, Valencia) Jose Cabrera (Zorita, Guadalajara)
999 939 894 438
PWR PWR PWR BWR
951 153
BWR PWR
Oskarshamn 1 (Oskarshamn, Kalmar) Oskarshamn 2 (Oskarshamn, Kalmar) Oskarshamn 3 (Oskarshamn, Kalmar) Ringhals 1 (Varberg, Halland) Ringhals 2 (Varberg, Halland) Ringhals 3 (Varberg, Halland) Ringhals 4 (Varberg, Halland) Forsmark 1 (Forsmark, Uppsala) Forsmark 2 (Forsmark, Uppsala) Forsmark 3 (Forsmark, Uppsala) Barsebaeck 1 (Barsebaeck, Malmohus) Barsebaeck 2 (Barsebaeck, Malmohus)
445 605 1160 835 875 915 915 970 970 1155 615 615
BWR BWR BWR BWR PWR PWR PWR PWR BWR BWR BWR BWR
Muehleberg (Muehleberg, Bern)
355
BWR
Goesgen (Daeniken, Solothurn)
965
PWR
1030
BWR
Beznau 1 (Doettingen, Aargau)
365
PWR
Beznau 1 (Doettingen, Aargau)
357
PWR
Slovenia
Sweden
Nuklearna Elektrana Krsko
Asociacion Nuclear Asco Endesa- Iberdrola S.A. Central Nuclear Vandellos II, A.I.E. Central de Trillo Central Nuclear de Almaraz Centrales Nucleares del Norte, SA Iberdrola SA Union Electrica, SA, and Fuerzas Electricas del Noroeste, SA OKG Aktiebolag
Vattenfall
Sydkraft AB
Switzerland
Bernische Kraftwerke AG (BKW) Kernkraftwerk Goesgen-Daeniken AG Kernkraftwerk Leibstadt AG Nordostschweizerische Kraftwerk AG (NOK)
Leibstadt (Leibstadt, Aargau)
The Nuclear Fuel Qvcle
319
Country
Authority
Power Station
Net MWe
Type
Taiwan, China
Taiwan Power Co.
Chinshan 1 (Chinshan, Taipei) Chinshan 2 (Chinshan, Taipei) Kuosheng 1 (Kuosheng, Wang-Li, Taipei) Kuosheng 2 (Kuosheng, Wang-Li, Taipei) Maanshan 1 (Herng Chuen) Maanshan 2 (Herng Chuen) Lungmen 1 (Kungliao, Taipei) Lungmen 2 (Kungliao, Taipei)
604 604 948 948 890 890 1350 1350
BWR BWR BWR BWR PWR PWR BWR BWR
Ukraine
Energoatom
Chernobyl 1 (Pripyat, Kiev) Chernobyl 3 (Pripyat, Kiev) Khmel'nitskiy 1 (Neteshin, Khmel'nitskiy) Khmel'nitskiy 1 (Neteshin, Khmel'nitskiy) Khmel'nitskiy 3 (Neteshin, Khmel'nitskiy) Khmel'nitskiy 4 (Neteshin, Khmel'nitskiy) Rovno 1 (Kuznetsovsk, Rovno) Rovno 2 (Kuznetsovsk, Rovno) Rovno 3 (Kuznetsovsk, Rovno) Rovno 4 (Kuznetsovsk, Rovno) South Ukraine 1 (Konstantinovka, Nikolaev) South Ukraine 2 (Konstantinovka, Nikolaev) South Ukraine 3 (Konstantinovka, Nikolaev) South Ukraine 4 (Konstantinovka, Nikolaev) Zaporozhye 1 (Energodar, Zaporozhye) Zaporozhye 2 (Energodar, Zaporozhye) Zaporozhye 3 (Energodar, Zaporozhye) Zaporozhye 4 (Energodar, Zaporozhye) Zaporozhye 5 (Energodar, Zaporozhye) Zaporozhye 6 (Energodar, Zaporozhye)
925 925 950 950 950 950 361 384 950 950 950
LGR LGR PWR PWR PWR PWR PWR PWR PWR PWR PWR
950
PWR
950
PWR
590
PWR
950 950 950 950 950 950
PWR PWR PWR PWR PWR PWR
555 555 1188 585
AGR AGR PWR AGR
585
AGR
575 575 550 550 625 625
AGR AGR AGR AGR AGR AGR
United Kingdom
British Energy plcNuclear Electric plc
Dungeness B 1 (Lydd, Kent) Dungeness B2 (Lydd, Kent) Sizewell B (Sizewell, Suffolk) Hinkley Point B 1 (Hinkley Point, Somerset) Hinkley Point B2 (Hinkley Point, Somerset) Hartlepool 1 (Hartlepool, Cleveland) Hartlepool 2 (Hartlepool, Cleveland) Heysham A1 (Heysham, Lancashire) Heysham A2 (Heysham, Lancashire) Heysham B 1 (Heysham, Lancashire) Heysham B2 (Heysham, Lancashire)
continued
Chapter 7
320
Table 7.3 (continuation) Country
Authority
Power Station
British Energy plc Scottish Nuclear Ltd.
Hunterston B 1 (Ayrshire, Strathclyde) Hunterston B2 (Ayrshire, Strathclyde) Torness 1 (Dunbar, East Lothian) Torness 2 (Dunbar, East Lothian) Calder Hall 1 (Seascale, Cumbria) Calder Hall 2 (Seascale, Cumbria) Calder Hall 3 (Seascale, Cumbria) Calder Hall 4 (Seascale, Cumbria) Chapelcross 1 (Annan, Dumfriesshire) Chapelcross 2 (Annan, Dumfriesshire) Chapelcross 3 (Annan, Dumfriesshire) Chapelcross 4 (Annan, Dumfriesshire)
British Nuclear Fuels plc
United States
Arizona Public Service Co.
Palo Verde 1 (Wintersburg, AZ) Palo Verde 2 (Wintersburg, AZ) Palo Verde 3 (Wintersburg, AZ) Baltimore Gas & Calbert Cliffs 1 (Lusby, MD) Electric Co. Calbert Cliffs 1 (Lusby, MD) Boston Edison Co. Pilgrim (Plymouth, MA) Carolina Power & Light Brunswick 1 (Southport, NC) Co. Brunswick\2 (Southport, NC) Robinson 2 (Hartsville, SC) Shearon Harris (New Hill, NC) The Cleveland Electric Perry 1 (North Perry, OH) Illuminating Co. Commonwealth Edison Braidwood 1 (Braidwood, IL) Co. Braidwood 2 (Braidwood, IL) Byron 1 (Byron, IL) Byron 1 (Byron, IL) Dresden 2 (Morris, IL) Dresden 3 (Morris, IL) LaSalle County 1 (Seneca, IL) LaSalle County 2 (Seneca, IL) Quad Cities 1 (Cordova, IL) Quad Cities 2 (Cordova, IL) Zion 1 (Zion, IL) Zion 2 (Zion, IL) Consolidated Edison Indian Point 2 (Buchana, NY) Co. Consumers Energy Co. Big Rock Point (Charlevoix, MI) Palisades (South Haven, MI) Detroit Edison Co. Fermi 2 (Newport, MI) Duke Power Co. Catawba 1 (Clover, SC) Catawba 2 (Clover, SC) McGuire 1 (Cornelius, NC) McGuire 2 (Cornelius, NC) Oconee 1 (Seneca, SC)
Net MWe
Type
575 575 625 625 50 50 50 50 50 50 50 50
AGR AGR AGR AGR GCR GCR GCR GCR GCR GCR GCR GCR
1270 1270 1270 825 825 670 767 754 683 860 1205
PWR PWR PWR PWR PWR BWR BWR BWR PWR PWR BWR
1120 1120 1105 1105 794 794 1078 1078 789 789 1040 1040 975
PWR PWR PWR PWR BWR BWR BWR BWR BWR BWR PWR PWR PWR
67 781 810 1129 1129 1129 1129 846
BWR PWR BWR PWR PWR PWR PWR PWR
321
The Nuclear Fuel Cycle
Country
Authority
Power Station
Oconee 2 (Seneca, SC) Oconee 3 (Seneca, SC) Duquesne Light Co. Beaver Valley 1 (Shippingport, PA) Beaver Valley 2 (Shippingport, PA) Arkansas Nuclear One 1 (Russellville, AR) Entergy Operations, Arkansas Nuclear One 2 (Russellville, AR) Inc. Grand Gulf (Port Gibson, MS) Waterford 3 (Taft, LA) River Bend (St. Francisville, LA) Florida Power and Light St. Lucie 1 (Hutchinson Island, FL) Co. St. Lucie 2 (Hutchinson Island, FL) Turkey Point 3 (Florida City, FL) Turkey Point 4 (Florida City, FL) Florida Nuclear Corp. Crystal River 3 (Red Level, FL) Georgia Power Edwin I. Hatch 1 (Baxley, GA) Company Edwin I. Hatch 2 (Baxley, GA) Alvin W. Vogtle 1 (Waynesboro, GA) Alvin W. Vogtle 2 (Waynesboro, GA) GPU Nuclear Corp. Oyster Creek (Forked River, NJ) Three Mile Island 1 (Londonderry Twp., PA) Houston Lighting & South Texas Project 1 (Palacois, TX) Power Co. South Texas Project 2 (Palacois, TX) Illinois Power Co. Clinton (Clinton, IL) Indiana/Michigan Donald C. Cook 1 (Bridgman, MI) Power Co. Donald C. Cook 1 (Bridgman, MI) IES Utilities, Inc. Duane Arnold (Palo, IA) Maine Yankee Atomic Maine Yankee (Wiscasset, ME) Power Co. Nebraska Public Power Cooper (Brownville, NE) District New York Power James A. FitzPatrick (Scriba, NY) Authority Indian Point 3 (Buchanan, NY) Niagara Mohawk Power Nine Mile Point 1 (Scriba, NY) Corp. Nine Mile Point 2 (Scriba, NY) North Atlantic Energy Seabrook (Seabrook, NH) Service Corp. Northeast Utilities Millstone 1 (Waterford, CT) Millstone 2 (Waterford, CT) Millstone 3 (Waterford, CT) Monticello (Monticello, MN) Northern States Power Prairie Island 1 (Red Wing, MN) Prairie Island 1 (Red Wing, MN) Co. Omaha Public Power Fort Calhoun (NE) District Pacific Gas & Electric Diablo Canyon 1 (Avila Beach, CA) Co. Diablo Canyon 2 (Avila Beach, CA)
Net MWe
Type
846 846 810 833 836 858 1173 1075 936 839 839 666 666 825 810 820 1162 1162 619 786
PWR PWR PWR PWR PWR PWR BWR PWR BWR PWR PWR PWR PWR PWR BWR BWR PWR PWR BWR PWR
1250 1250 930 1020 1090 538 860
PWR PWR BWR PWR PWR BWR PWR
764
BWR
780 965 61.0
BWR PWR BWR
1080 1150
BWR PWR
660 875 1149 536 503 500 478
BWR PWR PWR BWR PWR PWR PWR
1073 1087
PWR PWR continued
Chapter 7
322
Table 7.3 (continuation) Country
Authority
Power Station
Net MWe
Type
Pennsylvania Power & Light Co. PECO Energy Co.
Susquehanna I (Berwick, PA) Susquehanna 2 (Berwick, PA) Limerick 1 (Pottstown, PA) Limerick 2 (Pottstown, PA) Peach Bottom 2 (Delta, PA) Peach Bottom 3 (Delta, PA) Hope Creek (Salem, NJ) Salem 1 (Salem, NJ) Salem 2 (Salem, NJ) R.E. Ginna (Ontario, NY)
860 860 1055 1055 1159 1035 1031 1106 1106 470
PWR PWR BWR BWR BWR BWR BWR PWR PWR PWR
Virgil C. Summer (Parr, SC)
885
PWR
San Onofre 2 (San Clemente, CA) San Onofre 3 (San Clemente, CA)
1070 1080
PWR PWR
Joseph M. Farley 1 (Dothan, AL) Joseph M. Farley 2 (Dothan, AL) Bellefonte 1 (Scottsboro, AL) Bellefonte 2 (Scottsboro, AL) Browns Ferry 1 (Decatur, AL) Browns Ferry 2 (Decatur, AL) Browns Ferry 3 (Decatur, AL) Sequoyah 1 (Soddy-Daisy, TN) Sequoyah 2 (Soddy-Daisy, TN) Watts Bar 1 (Spring City, TN) Watts Bar 2 (Spring City, TN) Comanche Peak 1 (Glen Rose, TX) Comanche Peak 2 (Glen Rose, TX) Davis-Besse (Oak Harbor, OH) Callaway (Fulton, MO) Vermont Yankee (Vermont, VT)
860 860 1213 1213 1065 1065 1065 1148 1148 1177 1177 1150 1150 877 1171 504
PWR PWR PWR PWR BWR BWR BWR PWR PWR PWR PWR PWR PWR PWR PWR BWR
North Anna 1 (Mineral, VA) North Anna 2 (Mineral, VA) Surry 1 (Gravel Neck, VA) Surry 2 (Gravel Neck, VA) WNP-2 (Richland, WA)
893 897 801 801 1157
PWR PWR PWR PWR BWR
Point Beach 1 (Two Rivers, WI) Point Beach 2 (Two Rivers, WI) Kewaunee (Carlton, WI)
485 485 503
PWR PWR PWR
Wolf Creek (Burlington, KS)
1160
PWR
Public Service Electric & Gas Co. Rochester Gas & Electric Corp. South Carolina Electric & Gas Co. Southern California Edison Co. and San Diego Gas & Electric Co. Southern Nuclear Operating Co. Tennessee Valley Authority
Texas Utilities Electric Co.
Toledo Edison Co. Union Electric Co. Vermont Yankee Nuclear Power Corp. Virginia Power
Washington Public Power Supply System Wisconsin Electric Power Co. Wisconsin Public Service Corp. Wolf Creek Nuclear Operating Corp.
The Nuclear Fuel Cycle
323
It is of interest to refer to the article by H. Blix (1997) who was director general of IAEA for many years until 1997. Presently, oil, gas, and coal--the fossil fuels-provide nearly 85% of the commercial energy that the world uses: close to 37% for oil, 25% for coal, and more than 21% for gas, with nuclear power and hydro power providing around 7% each, and commercial renewables such as solar, wind and biomass nearly 2.5%. (Noncommercial uses of renewable energy are estimated to provide another 10% of world energy consumption). In China, coal presently supplies 75% of energy consumption, oil about 17%, nuclear and hydro 5%, and gas 2%. According to Blix (1997), so-called "renewable sources" total a little more than 2% of world commercial energy. The bulk of that total comes from geothermal installations, new wind and solar technologies, and biomass plantations. This share could increase, but only to a limited extent. The estimate made by the World Energy Council for new renewable supplies in the medium term is that with adequate support, the share of new renewable energy supplies, currently only 2%, could reach 5% to 8% of increased world energy supply by 2020. The argument is related to the energy density which is so variable. For example: 9 1 kg firewood produces about 1 kWh of electricity, 9 1 kg of coal produces about 3 kWh of electricity, 9 1 kg of oil produces about 4 kWh of electricity, 1 kg of natural uranium produces about 50 000 kWh of electricity, and 9 1 kg of plutonium produces about 6 000 000 kWh. The low energy density of the renewable sources means that if you want significant amounts of energy (electricity) from them, you must "harvest" them over large areas-and this is expensive. It has been calculated that to achieve the electricity generating capacity of a 1000-MWe power plant, an area of 50 to 60 km 2would be needed to install solar cells or windmills, or an area of 3000 to 5000 km 2 to grow the necessary biomass. It will not be easy--or cheap--to acquire such large areas, particularly in densely populated areas where the energy will be most needed (Blix, 1997).
7.3 N U C L E A R S A F E T Y Energy production as well as other human activities are always connected with risk taking. Radiation, and everything related to it, generates a fear not easily understood. This probably comes from the "invisibility" of the danger and relation to the bomb. Therefore the safety of nuclear power must be compared with the safety of alternative ways of generating electricity. The largest accidents in terms of casualties in the energy field are connected with the collapse of hydro dams. Some 2500 people perished, for example, in a single dam failure in Macchu, India. There are also, as we know, severe accidents connected with the transport and storage of gas, the mining of coal, and the shipping of oil. A gas pipeline explosion in Guadalajara in Mexico killed 200 people in 1992.
324
Chapter 7
Although one knows that the risk of incidents and accidents is not zero for any form of energy generation, including nuclear, one needs to be aware that most events are not very damaging. To help the nuclear power industry clarify the magnitude of events, the IAEA introduced the International Nuclear Event Scale (INES), which grades accidents from 1 to 7--much as seismologists grade earthquakes. It is hoped that this scale will help the media and public to realize that most incidents are of very minor significance and result in no threat to public health. It should also be remembered that most evolving technologies, whether boilers during the 19th century, airplanes in this century, or nuclear plants, entail some accidents from which lessons are learned. Both the Three Mile Island accident, from which only limited radioactivity escaped to the environment, and the Chernobyl disaster, have led to the introduction of new safety features in nuclear reactors, in plant operating procedures, and in regulations. The development of nuclear and radiation safety standards is a statutory responsibility of the International Atomic Energy Agency, IAEA, in Vienna, Austria. The IAEA Statute authorizes the Agency to establish standards of safety and to provide for the applications of these standards. Until now IAEA has developed and issued more than 200 standards of safety in the Agency' s Safety Series publications. They cover the fields of nuclear safety and radiation safety, including radioactive waste safety and radioactive material transport safety. The IAEA publications on this matter can be grouped into five categories: general safety, nuclear safety, radiation safety, waste safety and transport safety. Some of the most important publications are listed in the references under "safety series". In recent years, legally binding international conventions have come to play a crucial role in improving nuclear, radiation and waste safety. The major international conventions related to safety that have been negotiated and adopted under the auspices of the IAEA are listed in Table 7.4 (from Flakus and Johnson, 1998). (1) The Convention of Early Notification of a Nuclear Accident and the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency deal with aspects of emergency response and preparedness. Both of these Conventions---briefly referred to as the "Notification Convention" and the "Assistance Convention"--were adopted within a very short time span of only five months after the Chernobyl accident in 1986. The Notification Convention applies in the event of any accident involving facilities or activities of a State Party, or those under its jurisdiction or control, from which a release of radioactive material occurs or is likely to occur, and which has resulted or may result in an international transboundary release that could be of radiological safety significance for another State. A State Party involved in an accident covered by the Convention is obliged to immediately notify, directly or through the IAEA, those States which are or may be physically affected. To perform its functions under this Convention, the IAEA set up, at its headquarters in Vienna, an Emergency Response Center (ERC) for receiving, collating, and rapidly transmitting relevant information. Close co-operation with the World Meteorological
325
The Nuclear Fuel Cycle
Table 7.4 The global legal framework for nuclear, radiation, and waste safety (after Flakus and Johnson, 1998)
Convention on the Physical Protection of Nuclear Material
Entry into force
Development & status
8 Feb. 1987
In 1997, two States (Cuba and Lebanon) acceded to the Convention. As of May 1998, the Convention had 60 Parties.
Convention on Early Notification 27 Oct. 1986 of a Nuclear Accident
In 1997, four States (Lebanon, Philippines, Myanmar, and Singapore) agreed to be bound by the Convention. As of May 1998, the Convention had 80 Parties.
Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency
26 Feb. 1987
In 1997, three States (Lebanon, Philippines, and Singapore) agreed to be bound by the Convention. As of May 1998, the Convention had 75 Parties.
Convention on Nuclear Safety
24 Oct. 1996
In 1997, ten States (Argentina, Austria, Belgium, Brazil, Germany, Greece, Luxembourg, Pakistan, Peru, and Singapore) and in 1998 four States (Italy, Republic of Moldova, Portugal, and Ukraine) agreed to be bound by the Convention. As of May 1998, the Convention had 46 Parties.
Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management
Vienna Convention on Civil Liability for Nuclear Damage
Protocol to Amend the Vienna Convention and Convention on Supplementary Compensation for Nuclear Damage
A Diplomatic Conference, held in Vienna in September 1997, adopted the Joint Convention which was opened for signature on 29 September 1997. As of 4 June 1998, the Convention had been signed by 33 States and ratified by three States (Canada, Hungary, Norway). 12 Nov. 1977
In 1997, one State (Lebanon) ratified the Convention, and two States (Belarus, Israel) signed the Convention. The Convention had 29 Parties. Both of these legal instruments were adopted on 12 September 1997 and opened for signature on 29 September 1997. As of 18 June 1998, the Protocol had been signed by 13 States (Argentina, Czech Republic, Hungary, Indonesia, Italy, Lebanon, Lithuania, Morocco, Peru, Philippines, Poland, Romania, and Ukraine); and the Convention on Supplementary Compensation for Nuclear Damage had been signed by 13 States (Argentina, Australia, Czech Republic, Indonesia, Italy, Lebanon, Lithuania, Morocco, Peru, Philippines, Romania, Ukraine, and United States).
326
Chapter 7
Organization (WMO) resulted in the use of WMO's Global Telecommunication System (GTS) for rapid simultaneous transmission of voluminous meteorological and radiological data to national contact points. (2) The Convention on Nuclear Safety was developed during the period 1992-94. It applies to land-based civil nuclear power plants and is the first international legal instrument that directly addresses the issue of safety of such plants. The Convention contains obligations for State Parties to take national measures with respect to safety matters~such as the legislative and regulatory framework, assessment and verification of safety, emergency preparedness and operation of nuclear power plants and to report on the measures taken to implement each of the obligations under the Convention. (3) The Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management was adopted at a Diplomatic Conference in Vienna in September 1997. The Convention is focused predominantly on specific activities rather than on substances. It applies with certain restrictions to: (i) the safety of spent fuel management, (ii) the safety of radioactive waste management, (iii) the safety of management of spent fuel or radioactive waste resulting from military or defense programmes if and when such materials are transferred perm~inently to and managed within exclusively civilian programmes. (4) The Convention on the Physical Protection of Nuclear Material came into force in 1987. This Convention prescribes the levels at which nuclear material used for peaceful purposes is to be protected while in international nuclear transport, and requires each party to the Convention not to permit the export or import of such material unless it is satisfied that the nuclear material will be protected at those levels. (5) At a Diplomatic Conference in September 1997, delegates from 80 States adopted the Protocol to Amend the 1963 Vienna Convention on Civil Liability for Nuclear Damage and also the Convention on Supplementary Compensation for Nuclear Damage. The Protocol sets the possible limit of the operator's liability at an amount roughly equivalent to US $400 million and also contains an enhanced definition of nuclear damage which covers costs of reinstatement of any damaged environment and costs of preventive measures, extends the geographical scope of the Vienna Convention and extends the period during which claims may be made for loss of life and personal injury.
7.4 R E L E A S E S OF E F F L U E N T S
Radioactive materials released to the environment are sources of exposure and potentially harmful. Such releases may be from different activities in the nuclear fuel cycle, mining operations or industrial users. Strict control measures must be employed to keep the resulting doses "as low as reasonably achievable". This implies the implementation of protective and control measures and includes the setting of limits for radiation exposure.
The Nuclear Fuel Cycle
327
A limit is a value that must not be exceeded and the primary dose limits for individuals are set by the ICRP. These limits are related to individuals irrespective of the source. If an individual is likely to be exposed to other sources of radiation, source related limits must be set by a regulatory authority. These limits must be lower than the dose limit and are called the source upper bound. Authorized limits are limits specified by the regulating authority for a specific practice or source. In setting limits the authority must consider the requirements of radiation protection and individual dose limitation. The authorized limits will not exceed the upper bound. For practical reasons limits for releases of radioactive effluents to the environment are expressed as limits of releases over a specified period. It is important to set reference levels for all activities. A reference level is not a limit but indicates a course of action like recording data, investigation or intervention. These levels are determined by radiation protection factors and the extent of the measures taken must be described in the operating procedures. In the case of new practices where reassessment may result in lower or higher release rates being acceptable, setting authorized limits can be difficult. There is often justification for a specific source or practice to be exempted from normal regulations. The regulating authority may exempt such sources or practices on the basis that the individual and collective doses are so low that they may be ignored. The individual dose limit is the starting point for calculating the upper bound. The dose upper bound will be less by the dose contributed by global and regional sources of exposure. The regulating authority may reserve a margin for future development of the activity or practice. This margin is set by specifying that a fraction, F, of the primary dose limit must not be exceeded. The maximum annual dose limit to the critical group is limited by: H~oca, + Hregiona, + Hg,oba, < F x Olimit
(7.1)
where H~im~' is the primary dose limit and the suffixes refer to the components of the total dose to the critical group. The source specific dose upper bound (HUB) for all the controlled sources of exposure is given by:
HuB= F
x Olimi t - mregiona 1 - Ogloba 1
(7.2)
The upper bound for annual release can then be derived from the dose upper bound by using the overall transfer factors ~,kl) wherej represents population group, k represents release mode and 1 represents the radionuclide. If the dose commitment to the critical group j' per unit release of a radionuclide is given by fj,kt then the release upper bound, Rk~, is given by R~I =
HUB f j'kt
provided that no other radionuclides are released.
(7.3)
Chapter 7
328
In practice the situation will be more complex because more than one nuclide may be released and different modes of release will be developed. The total dose contribution to each population group due to a release R~ is given by: Jjk, = fj~,flk,.
(7.4)
If different release modes (k) are developed, the release upper bounds (Rkl) for each release mode is given by:
fj,k/R~,,< Hue
(7.5)
k
This is true only for the release of one radionuclide. If a mixture of nuclides is released that contributes to the exposure of group j', the release upper bound, Rkt, must satisfy the condition: ~ k
f;k,R,, < HuB
(7.6)
l
The condition defines a set of values that constitute the release upper bound. For routine releases of radioactivity two main control options are considered: 1. storage of effluents to allow short-lived radionuclides to decay before release; and 2. treatment of effluents to remove radionuclides before release. Within these categories a number of options may be available. The various possibilities must be identified and investigated. Considerations like operating and maintenance cost, the implications for the waste management program as well as the individual and collective dose for the workers and the public must be taken into account. The first step in optimizing is to ensure that the releases anticipated with the control options meet the requirements of the source upperbound. Any control option that does not meet this requirement is not considered. The final element in implementing a system of dose limitation is to optimize radiological protection by selecting the control option for which radiation doses are "as low as reasonably achievable". For the monitoring of effluent releases the samples collected in the vicinity of nuclear installations must be representative of land and water utilization as well as meteorological factors. Samples must be analyzed for those nuclides which contribute most to public exposure. Air sampling is of special interest. Usually fixed monitoring instruments are used for continuous routine monitoring in the vicinity of the installation. If a limit has been exceeded the cause must be traced and corrective measures must be taken immediately. Two types of sampling monitors are in general use: air samples are used to assess the airborne contamination levels at selected points. In the case of particulate materials a volume of air is drawn through a filter paper on which the particulates are deposited. An alarm may be set on increase of activity.
The Nuclear Fuel Cycle
329
For gaseous materials carbon cartridges are used to trap the contaminating materials. Special devices are used for trapping iodine. "Stack" monitors for gaseous effluents give a rough estimate of the radioactivity of the effluent from a stack. The radioactive content of the samples can be assessed by using standard counting equipment. Sample measuring instruments are operated in contamination-free laboratories. For the monitoring of the released liquid effluents the following methods are used. Samples of effluents are collected by simple dipping devices and analyzed before release. In the case of monitoring streams in the neighbourhood of installations, automatic samplers collecting samples over a 24-hour period are used. Samples are analyzed and records must be kept of results. The water effluent meter monitors water or coolants and may be connected to a rate meter, recorder or alarm system. On site and off-site environmental monitoring at and near nuclear power plants, nuclear reactors and other fuel cycle activities are shown in Tables 7.5 and 7.6. By way of illustration, we shall mention the case of the Sellafield reprocessing plant in Cumbria, UK, as discussed by Jones et al. (1995). For the last decade, the existence of a higher than average rate of childhood leukaemia in young people from the village of Seascale in Cumbria has led to speculation that radioactive discharges from the reprocessing plant at Sellafield may be a causative factor, even though the calculated doses are too small for the leukaemia risk observed in epidemiological studies (Stother et al., 1984; 1986). These estimates of historical doses from discharge from the plant have relied on calculations based on recorded discharges and conventional environmental models. This has left open the question of whether the recorded discharges, particularly in the earlier years of plant operation, might have been seriously underestimated. A major reassessment of historical discharges and doses has been carried out, prompted in large part by civil litigation instigated by a number of local families against British Nuclear Fuels plc, the operators of the Sellafield plant. The reassessment involved the development of the Sellafield Environmental Assessment Model (SEAM), which was used both to calculate doses and to build confidence in the discharge chronology from recorded measurements of environmental concentrations and current assessments of environmental inventories. The SEAM model has put together established models of atmospheric dispersion and deposition, terrestrial foodchains, marine dispersion and concentration in marine biota and the sea-to-land transfer of radionuclides. Environmental measurements from a wide variety of sources have been compared against values calculated from the discharge chronology and the SEAM model in order both to validate the model and to build confidence in the discharge chronology. The established chronology for liquid discharges to the Irish Sea was confirmed by validation against historic environmental monitoring data and dated sediment cores (Kershaw et al., 1990). It was possible to establish good agreement, thus building confidence in the discharge chronology. The review of atmospheric discharges has indicated that earlier figures for emissions, particularly for particulates, in the earlier years of plant operation were
330
Chapter 7
Table 7.5 On-site monitoring at nuclear power plants, nuclear reactors and uranium mill and/or fuel cycle facilities Sample type
Collection frequency
Analysis frequency
Airborne particulates
Continuous
Continuous readout
Liquid effluents
Continuous
Continuous readout
Drinking water
Quarterly on composites
Surface water
Semi-continuous (samples taken 3-6 hours) Monthly
Noble gases
Continuous
Continuous readout
Groundwater
Quarterly
y-spectrometry on each batch sample; annual composite on other nuclides detected
Quarterly on composites
Table 7.6 Off-site environmental monitoring near nuclear power plants, nuclear reactors and/or uranium mill and/or fuel cycle facilities Sample type
Collection frequency
Analysis frequency
Drinking water
Semi-continuous composite
T-spectrometry on each batch sample; annual composite on other nuclides detected
Milk
Weekly or bi-weekly at farms; monthly at dairy. T-spectrometry weekly
Food crops
At harvest
T-spectrometry
Fish
During fishing season or semi-annually
T-spectrometry
Shellfish
Semi-annually
y-spectrometry
Sediments
Semi-annually
T-spectrometry
significantly underestimated. Further, i m p r o v e d estimates were m a d e of u n m o n i t o r e d e m i s s i o n s of p l u t o n i u m f r o m the site, which w e r e k n o w n to have occurred in the 1950s and 1960s by utilizing the results of cumulative deposition m e a s u r e m e n t s in soil cores. Despite the higher assessed discharges to a t m o s p h e r e , calculated doses to m e m b e r s of the public in Seascale remain low and are insufficient to account for any excess of l e u k a e m i a . F u r t h e r m o r e , where specific m e a s u r e m e n t s of radionuclide body contents of local residents are available (Statner et al., 1988) the m o d e l significantly overestimates body content.
The Nuclear Fuel @cle
331
7.5 M A N A G E M E N T OF RADIOACTIVE WASTES Concern about nuclear power is usually focused on the highly toxic and radioactive spent fuel and nuclear waste. What is characteristic of these, however, in addition to their toxicity and radioactivity, is that they are limited in volume, which facilitates waste disposal. This contrasts sharply with the waste disposal problem for fossilfuelled plants. More specifically, a 1000-MWe coal plant with optimal pollution abatement equipment will annually emit into the atmosphere 900 tonnes of SO 2, 4500 tonnes of NO x, 1300 tonnes of particulates, and 6.5 million tonnes of CO 2. Depending on the quality of the coal, up to 1 million tonnes of ashes containing hundreds of tonnes of toxic heavy metals (arsenic, cadmium, lead, and mercury) will have to be disposed of. By contrast, a nuclear plant of 1000-MWe capacity produces annually some 35 tonnes of highly radioactive spent fuel. If the spent fuel is reprocessed, the volume of highly radioactive waste will be about 3 m 3. The entire nuclear chain supporting this 1000-MWe plant, from mining through operation, will generate, in addition, some 200 m 3 of intermediate-level waste and some 500 m 3 o f l o w level waste of year. Most countries using nuclear electrical generation have programmes for safe disposal of the wastes. Technical alternatives for disposal of spent fuel and high-level wastes have been assessed by several countries and international organizations. Scientific consensus exists that geologic disposal using a system of natural and engineered barriers is the preferred method to be used. Unlike chemically hazardous industrial wastes, the much smaller volumes of spent fuel and high-level waste make containment and isolation a feasible disposal option, and their radiological hazard will decrease with time. Generic studies of geologic disposal conducted by the Swedish KBS, the Commission of European Communities (CEC), and others have concluded that geologic disposal systems can achieve an acceptable level of safety to protect future generations from the radiological hazards associated with these wastes. In 1991 IAEA established the Radioactive Waste Safety Standards (RADWASS) programme to develop a special series of safety documents specifically directed at radioactive waste management. The purpose of the RADWASS programme is to document existing intemational consensus in the approaches and methodologies for safe waste management and disposal; to create a mechanism for establishing consensus where it does not exist; and to provide Member States with a comprehensive series of intemationally agreed documents to complement national standards and criteria. RADWASS has been organized in a hierarchical structure of four levels of safety documents. The top-level publication is a document of safety fundamentals which provides the basic safety objectives and fundamental principles to be followed in national waste management programmes. The lower levels include safety standards, safety guides, and safety practice documents. The series has been structured in a logical and clear manner to reflect the systems approach to waste management.
332
Chapter 7
Radioactive waste is any material that contains, or is contaminated with, radionuclides at concentrations of radionuclides greater than the "exempted quantities" established by the regulatory body and for which no future use is foreseen. This is after a definition by IAEA. Five main activities produce such waste: 9 uranium and thorium mining and milling; 9 nuclear fuel cycle operations such as uranium conversion and 9 enrichment, fuel fabrication, and spent fuel reprocessing; 9 operations of nuclear power stations; 9 decontamination and decommissioning of nuclear facilities; 9 institutional uses of isotopes. The waste resulting from the above activities comes in various forms (i.e., gaseous, liquid, or solid). These wastes have different characteristics. For safety and technical reasons, the various forms of wastes are usually categorized by their levels of radioactivity, heat content, and potential hazard. With regard to disposal the wastes are categorized as follows. Low-level wastes (LLW) contain a negligible amount of long-lived radionuclides. Produced by peaceful nuclear activities in industry, medicine, research, and by nuclear power operations, such wastes may include items such as packaged gloves, rags, glass, small tools, paper, and filters which have been contaminated by radioactive material. Disposal in near-surface structures or shallow burial is practised widely. Intermediate-level wastes (ILW) contain lower levels of radioactivity and heat content than high-level wastes, but they still must be shielded during handling and transport. Such wastes may include resins from reactor operations or solidified chemical sludges, as well as pieces of equipment or metal fragments. Commercial engineering processes are being used to treat and immobilize these wastes. Disposal options are similar to those for low-level wastes. High-level wastes (HLW) arise from the reprocessing of spent fuel from nuclear power reactors to recover uranium and plutonium. These wastes contain transuranic elements, and fission products that are highly radioactive, heat-generating, and longlived. Liquid HLW is usually immobilized as a solid glass matrix and stored in interim storage facilities prior to final disposal and isolation in deep, stable, geologic formations, as currently planned by many national programmes. Spent nuclear fuel that is not reprocessed is also considered high-level waste. Alpha-bearing wastes (also called transuranic, plutonium-contaminated material, or alpha wastes) include wastes that are contaminated with enough long-lived, alphaemitting nuclides to make near-surface disposal unacceptable. They arise principally from spent fuel reprocessing and mixed-oxide fuel fabrication. The wastes may be disposed of in a similar manner to HLW. Management and disposal of nuclear waste depends mainly on its type. For example, LLW and ILW are often treated (volume reduction) and/or conditioned (waste immobilization) prior to disposal. This area of LLW and ILW waste management, having been established and proven over past years, is considered to be quite
The Nuclear Fuel Cycle
333
mature in terms of technology development. As a result, several effective, safe, and feasible treatment and conditioning options exist for these types of wastes. They include: storage and decay, compaction and super compaction, incineration, chemical precipitation, evaporation, filtration, and ion-exchange; these may be followed by immobilization in materials like concrete, bitumen, or polymers (Chan, 1992). The most common disposal methods for ILW and LLW involve disposal in shallow earthen or concrete lined trenches or in structures on the ground (commonly referred to as engineered surface facilities). Safe near-surface disposal of LLW has been practised in a number of countries for almost 30 years. The rationale behind near-surface disposal is that the isolation period for this type of waste is relatively limited (up to 300 years) and, therefore, the institutional or administrative control of the disposal site can be assured. HLW management and disposal is quite different. After its useful life, spent nuclear fuel is removed from the reactor. Once removed, it is usually placed into temporary on-site storage before it is either: 9 placed in interim away-reactor storage (5-100 years), conditioned after a sufficient decay period, and stored before its eventual final disposal in a geologic repository; or 9 reprocessed after additional away-from-reactor storage. The resulting liquid high-level waste, containing mostly fission products and a small proportion of the actinides, is then immobilized in a stable matrix (i.e., borosilicate glass), and would then be disposed of in a geologic repository. Regardless of which option is chosen, there is broad scientific agreement that deep geologic disposal using a system of engineered and natural barriers to isolate these wastes is the preferred method for their disposal (Chan, 1992). According to the report by Oi (1998) at the end of 1997, more than 130,000 tonnes of spent fuel from power reactors were estimated to be stored world-wide containing about 1000 tonnes of plutonium. Another 170 tonnes of separated plutonium were in storage from civilian reprocessing operations, and about 100 tonnes of excess plutonium from dismantled warheads no longer required for defense purposes were scheduled to be released from the military sector of Russia and United States. Plutonium represents a dual challenge because it is a valuable energy source and a matter of global concern because of its potential health hazards and possible use for the production of nuclear weapons. Spent fuel from light water reactors contains about 1% of plutonium. According to Oi (1998) the IAEA estimates that in 1997 about 10,500 tonnes of spent fuel were discharged from nuclear power reactors world-wide; this amount contains about 75 tonnes of plutonium. It is estimated that the annual production figure will remain more or less the same until 2010. The cumulative amount of plutonium in spent fuel from nuclear power reactors worldwide is predicted to increase to about 1700 tonnes by 2010. It is estimated that about 3000 tonnes of spent fuel discharged from power reactors were reprocessed in 1997, which corresponds to about 30% of the total. About 24
Chapter 7
334
~-~ 150 O ,,.a
=
~"
lOO
"~
5o
O
o
i
i
i
I
I
I
I
I
I
I
i
i
i
i
i
I
I
I
I
i
i
1995 1997 1999 2001 2003 2005 2007 2009 2011 2013 2015
Year Fig. 7.3. Projected worldwide civil inventories of separated plutonium. Solid line: minimum Pu inventory, no Pu market; dashed line: end of year Pu inventory, no Pu market; dotted line: end of year Pu inventory with Pu market; dashed-dotted line: minimum Pu inventory with Pu market.
tonnes of plutonium were separated in reprocessing plants and nine tonnes of plutonium were used mainly as mixed uranium-plutonium oxide fuel (MOX) in light-water reactors. The imbalance between the separation and use of plutonium had resulted in an accumulated inventory of separated civil plutonium of about 170 tonnes at the end of 1997. IAEA projections of plutonium inventories show that the rate of separation of civil plutonium and its rate of use will fall into balance in a few years. This is due to an enhanced capacity of MOX fuel production which will amount to 360 tonnes of heavy metal per year in 2000. Beyond this period, the inventory is expected to decrease modestly and level off at around 130 tonnes. Despite the efforts to reduce the current inventories of separated civil plutonium, the worldwide inventories still remain at a substantial level, as shown in Fig. 7.3. In addition to the amounts of civil plutonium, plutonium is being released from dismantled warheads. Under the START-I and -II Treaties, many thousands of US and Russian nuclear war-heads are slated to be retired within the next decade. As a result, at least 50 tonnes of plutonium from each side are expected to be removed from military programmes. Oi (1998) points out the problem, which is what to do with plutonium either in a separated form or contained in spent fuel. A number of issues arise because of plutonium's potential use as an energy source and for the production of nuclear weapons. Presently, plutonium is used in light-water reactors as MOX fuel and also in small amounts for the development of fast-breeder reactors. Currently 22 power reactors in five countries (France, Germany, Switzerland, Belgium, and Japan) are loaded with MOX fuel and this number is expected to rise to between 36 and 48 by 2000. The use of MOX reduces the inventory of separated plutonium and is regarded as an interim measure before plutonium's possible full-scale use in fast reactors later in the next century. It is known that multiple recycling in light-water reactors degrades plutonium, which in turn limits the number of times it can be recycled to two or three. Such
The Nuclear Fuel Cycle
335
degraded plutonium can, however, be used as fuel in fast reactors. Without such reactors, spent MOX fuels will still end up in a final depository or in storage facilities (Oi, 1998). Recently much attention has been given to the accelerator driven systems, burning in inert matrices, and the use of thorium to burn plutonium. The concept of a closed nuclear fuel cycle was traditionally considered as transmutation (burning) of only plutonium and recycled uranium, with minor actinides (neptunium, americium, curium) destined for final geological disposal. But as time goes on, a new understanding is emerging: reduction of the quantity of actinides would ease requirements for final repositories and make them relatively less expensive. Neutron transmutation of long-lived radioactive minor actinides by the fission process~which entails producing energy and simultaneously turning them into shorter-lived nuclides~is being intensely analyzed in the technical community. Also being proposed is the neutron transmutation of selected long-lived fission products. Several possibilities for the transmutation of long-lived nuclides by nuclear reactions have been suggested. In the beginning, the best choice appeared to be the use of nuclear reactors. However, recently there has been renewed interest in what are called accelerator-driven systems (ADS), a technology that seems to show good promise. ADS would produce large amounts of electrical energy while simultaneously destroying the plutonium. This appears to offer a better solution to the plutonium problem than multi-millennium storage. The use of accelerators for nuclear energy applications is not a new idea and was proposed as early as the late 1940s by E. Lawrence, inventor of the cyclotron. In the 1950s he promoted the development of a Materials Test Accelerator at Livermore to produce intense neutron fluxes for plutonium production. The Canadian Chalk-River Laboratory began intensive studies of accelerator-based systems to breed nuclear fuel for heavy-water reactors. Scientists at Brookhaven National Laboratory also actively promoted accelerator-based options in the late 1970s and early 1980s. For the last five years, scientists at Low Alamos National Laboratory have been re-evaluating the accelerator-based technology in the light of new advances in technology and the world energy perspective (Boowman et al., 1992). When 1.6 GeV protons strike a large radius target consisting of heavy nuclei such as lead, approximately 55 neutrons are generated per proton. The energy deposition for this process is about 30 MeV of proton energy per neutron compared with about 200 MeV of fission energy deposited per useful neutron from a sustained chain reaction in fissile material such as 235U. The heat per unit volume which must be handled for a given neutron production rate is therefore considerably smaller for the spallation source than for the reactor. The reactor has the further disadvantage that in nearly all designs the fuel is fixed in position with coolant flowing past. Heat deposition in the fuel is therefore limited by the conductivity of the fuel and by the heat capacity and conductivity of the coolant. For the accelerator-driven neutron source the target is itself a flowing liquid heavy metal. The heat load on the target is therefore limited only by the thermal properties of the liquid metal and by the rate at which it flows. This accelerator
Chapter 7
336
target is therefore capable of generating a much higher density (and therefore flux) of neutrons than a reactor with fixed fuel because of the much greater power density capability of the flowing target and the much lower energy deposition in the target per neutron produced. The neutron-production-transmutation system considered by Bowen et al. (1992) consists of an accelerator for the proton beam, a flowing heavy metal proton target for neutron production, and a surrounding blanket containing primarily heavy water (DzO) for moderating the neutrons into the thermal range. The neutron flux may be further enhanced by neutrons from actinide fission in the blanket. The actinide material is transported through the blanket as a molten salt mixed with the carrier salt LiF-BeF 2. Heat from the fission process is deposited in the molten salt and carried away by the salt at an exit temperature of 720~ which makes possible electric power generation at a high thermal-to-electric efficiency. A continuous flow system is essential because of the high burn-up rates of fissile material. For example the lifetime of 239pu in a thermal flux of 1016 n/cm2-s is only about one day so that use of fuel assemblies along the lines of standard practice for reactors is impractical. In addition there must be chemistry facilities for removing stable or short-lived fission products and returning radioactive waste to the blanket. The accelerator, target/blanket, electric power extraction and chemistry facilities are shown schematically in Fig. 7.4. The expected efficiency for conversion of thermal to electric power is 44% and the bussbar efficiency of the accelerator is 45%. A fraction of the electrical power is fed back to power the accelerator which operates at an energy of 1.6 GeV and produces neutrons in a Pb target. The beam power deposited in the Pb and the thermal power Accelerator
1 h F
power "
Heavy-water I moderator
i
Electric
I
] Energ3 [extracti(
Waste feed --~
I
~ I~l'~
'-~
., I - b - I
---
!
>11' Neutron flux 9
I Advanced ~ _ chemical ~ . _ separationsl -
I
I
Stable and short-lived products
'
9 I ii ~J ~
Molten salt loop
Liquid lead target
Fig. 7.4. The proton beam strikes the lead target generating neutrons which are moderated in the surrounding heavy water blanket. Molten salt carrying fissile material for heat generation and electric power production circulates in the heavy water blanket through double-walled pipes. Some of this power drives the accelerator. Nuclear waste including that produced in the molten salt is also circulated through the blanket in a separate loop and transmuted to stable and short-lived nuclides which are extracted and stored (after Bowen et al., 1992).
The Nuclear Fuel Cycle
337
ssion
\" Protons ~
k"
. Spallationtarget
1 GeV
Spallation residues
$ +
9
9
9
neutrons transmutation
Fig. 7.5. Scheme of a hybrid plant. An intense beam of proton from an accelerator hits a massive target (Pb, Pb-Bi, W, Hg .... ) and produces fast neutrons. In a surrounding blanket, the neutrons are used to drive a subcritical medium.
deposited in the D 2 0 blanket are not converted to electrical power although energy recovery loops on these systems could boost the overall system efficiency. The material to be transmuted can be fed either into the molten salt carrier as molten salt or into the heavy water as dissolved salt depending on the application (Bowen et al., 1992). Recently there has been increased interest in the idea of accelerator-driven reactors, see for example Tokizuka (1994), Van Tuyle et al. (1993), Carminati et al. (1993), Rubia et al. (1995). In such systems (Fig. 7.5), spallation reactions induced by a high-intensity beam (10 to 250 mA) of GeV protons on a heavy target produce an intense neutron flux. These neutrons, after being more or less moderated, are used to drive a sub-critical blanket. The extra neutrons provided by the accelerator allow the maintenance of the chain reaction while burning the long-lived nuclear waste. The plant generates electricity, part of which is used to supply the accelerator. Besides the more favourable neutron economy, additional advantages of accelerator-driven systems are safety and versatility. Obviously the operation of the blanket in a subcritical state is a major safety advantage. It could for instance allow the introduction of a large amount of Pu or minor actinides which is difficult in classical reactors because of control problems due to the smaller fraction of delayed neutrons. Accelerator-driven systems are also more flexible than reactors since the intensity of the accelerator can be adjusted to counteract the growth of poisonous isotopes or when adding elements to be transmuted. Their main draw-backs are their complexity and the technological progress they imply for the accelerator, the target-blanket and the interface between them (Boudard et al., 1998). Recently, it has been proposed to construct a demonstrator facility of significant power of the order of 100 MW (thermal) on a 10-year time schedule as a regional European facility.
Chapter 7
338
7.6 RESEARCH REACTORS Nuclear reactors have supported research in many different fields and have contributed to discoveries in many scientific disciplines. Altogether there are about 180 research reactors in operation in the world (IAEA, 1996). Research reactors have a very wide variety of uses, including neutron scattering (in which beams of thermal neutrons are scattered by the atoms in a sample, revealing its structure, magnetic state, and atomic binding energies); neutron activation analysis; radiography; irradiation testing of materials; and production of radioisotopes for medical, research, and industrial use. These capabilities are applied by researchers in many fields, ranging from archaeology to materials science and from fusion research to environmental science. Few generalizations can be made about the applications for research reactors or about their users. Research reactors themselves tend to have a very different set of safety-related parameters from power reactors. Some are helpful differences like simplicity, relatively low power, and low-temperature coolant. Other differences, especially the need for a high-power density core, pose challenges not faced in a power reactor. These challenges can be met through thoughtful design solutions. A research reactor' s power is usually in the range 0-100 MW thermal. The fission product inventory and the stored energy in research reactors are smaller than in power reactors. However, some of the research reactors have large power density (>5000 kW thermal per kg of fuel). The typical research reactor is of the swimming pool type, as shown in Fig. 7.6.
~
IILI I IIII i I ~ ~
REACTOR CORE
Fig. 7.6. Basic H20-cooled swimming-pool reactor (after West, 1997).
The Nuclear Fuel Cycle
339
In many of these reactors the core is made up of so-called "materials testing reactor"-type fuel elements which are aluminum-clad, curved plates of fuel arranged in long rectangular boxes arranged between grid plates to form the core. several positions in the grid are not occupied by fuel elements, but by control rods, beryllium reflectors, or experimental capsules. Cooling may be by natural convection of the pool water, although this is augmented, for operation at higher power, by pumping pool water through the core. More powerful research reactors, of which the international Institut Laue-Langevin (ILL) facility at Grenoble, France, and the High Flux Isotope Reactor (HFIR) at Oak Ridge, Tenn., are well-known examples, have tanks that are full pressure vessels~for example, the coolant inlet pressure at HFIR is nominally 470 psi, and at ILL it is 200 psi. Again, aluminum-clad fuel plates are used, the fuel meat being a layer, about 50 mils thick, of U308 particles mixed with powdered aluminum for enhanced thermal conductivity, the layer being clad with aluminum plates about 10 ml thick. In these two reactors, the fuel elements are annular, with curved (involute) plates fitting into axial grooves down two concentric cylinders.
7.7 ADVANCED NUCLEAR P O W E R PLANTS New generations of nuclear power plants have been or are being developed, building on this background of success and applying lessons learned from the experience of operating plants. Advanced designs currently under development comprise three basic types: water-cooled reactors, using water as coolant and moderator, 9 fast reactors, using liquid metal, e.g. sodium, as coolant, and gas-cooled reactors, using gas, e.g. helium, as coolant and graphite as moderator. Global developments in this field have been summarized by Juhn et al. (1997). We present here their findings for some countries. 9
9
United States Important programmes in development of ALWRs were initiated in the mid-1980s in the United States. In 1984, the Electric Power Research Institute (EPRI), in cooperation with the US Department of Energy initiated a programme to develop utility requirements for ALWRs to guide their design and development. Utility requirements were established for large boiling-water reactors (BWRs) and pressurized-water reactors (PWRs) having power ratings of 1200 to 1300 MWe, and for mid-sized BWRs and PWRs having power ratings of about 600 MWe. In 1986, the US Department of Energy, in cooperation with EPRI and reactor design organizations, initiated a design certification programme for evolutionary plants based on a new licensing process, followed in 1990 by a design certification programme for mid-size plants with passive safety systems. The new licensing process allows nuclear plant designers to submit their designs to the US Nuclear Regulatory Commission
340
Chapter 7
(NRC) for design certification. Once a design is certified, the standardized units will be commercially offered, and a utility can order a plant, confident that genetic design and safety issues have been resolved. The licensing process will allow the power company to request a combined licence to build and operate a new plant, and as long as the plant is built to pre-approved specifications, the company can start up the plant when construction is complete, assuming no new safety issues have emerged. Four advanced reactor designs developed in the United States have been submitted to the NRC for certification under the US Department of Energy ALWR programme. Two large evolutionary plantsmthe System 80+ of ABB-Combustion Engineering and the ABWR of General Electricmreceived Final Design Approval in 1994 and Design Certification in May 1997. The 600-MWe AP-600 of Westinghouse is under NRC review and a Final Design Approval is expected by March 1998. Up to mid-1996, the 600-MWe simplified BWR developed by General Electric was also under review, but then the company stopped work on the 600-MWe version and shifted its emphasis to a unit with larger output. The first-of-a-kind engineering programme (FOAKE, the detailed design needed to verify cost and the construction schedule) authorized by the 1992 Energy Policy Act was completed for the ABWR in September 1996, and similar work on the AP-600 has also been done. The power company in Taiwan, China, recently selected General Electric's ABWR design for two new units slated for operation in 2004.
France and Germany In Europe Framatome and Siemens have established a joint company, Nuclear Power International, which is developing a new advanced reactor, the European pressurizedwater reactor (EPR), a 1500-MWe plant with enhanced safety features. The basic design will be completed in mid-1997, and the design will be reviewed jointly by the French and German safety authorities. This procedure will provide strong motivation for the practical harmonization of the safety requirements of two major countries, which could later be enlarged on a broader basis. Siemens is also, together with German utilities, engaged in the development of an advanced BWR design, the SWR-1000, which will incorporate a number of passive safety features, for initiation of safety functions, for residual heat removal and for containment heat removal. Sweden and Finland In Sweden, ABB Atom, with involvement of the utility Teollisuuden Voima Oy (TVO) of Finland, is developing the BWR-90 as an upgraded version of the BWRs operating in both countries. Republic of Korea In the Republic of Korea, an effort started in 1992 to develop an advanced design known as the Korean Next Generation Reactor (KNGR), a 4000-MWth PWR design. The basic design is currently being developed by the Korea Electric Power Corporation (KEPCO) with the support of the Korean nuclear industry. The goal is to complete a detailed standard design by the year 2000.
The Nuclear Fuel Cycle
341
Russian Federation In the Russian Federation, design work is under way on the evolutionary V-392, an upgraded version of the VVER-1000, and another design version is being developed in cooperation with the Finnish company Imatran Voima Oy (IVO). Also being developed is a mid-sized plant, the VVER-640 (V-407), an evolutionary design which incorporates passive safety systems, and the VPBER-600, which is a more innovative, integral design. Construction of the first unit of the VVER-640 is planned to start at Sosnovy Bor in 1997. Construction of two 1000-MWe VVERs is being discussed with the People's Republic of China. Japan In Japan, the Ministry of Trade and Industry is conducting an "LWR Technology Sophistication" programme focusing on development of future LWRs and including requirements and design objectives. A large, evolutionary 1350-MWe advanced PWR is being developed by Japanese utilities together with nuclear vendors, with construction of a twin unit being planned at the Tsuruga site. In addition, an advanced BWR Improvement and Evolution study was started in 1991. It involves development of a reference 1500-MWe BWR that reflects the accumulated experience in operation and maintenance of BWRs. Also in progress are development programmes for a Japanese Simplified BWR (JSBWR) and PWR (JSPWR), projects which involve vendors and utilities. The Japan Atomic Energy Research Institute (JAERI) has been investigating conceptual designs of advanced water-cooled reactors with emphasis on passive safety systems. These are the JAERI Passive Safety Reactor (JPSR) and the System-Integrated PWR (SPWR). China In China, the Nuclear Power Institute (Chengdu) is developing the AC-600 advanced PWR, which incorporates passive safety systems for heat removal. In all of these countries, the advanced LWRs under development incorporate significant design simplifications, increased margins, and various technical and operational procedure improvements. These include better fuel performance and higher burn-up, a better man-machine interface using computers and improved information displays, greater plants standardization, improved constructability and maintainability, and better operator qualification and simulator training. Canada The continuing design and development programme for heavy-water cooled reactors (HWRs) in Canada is primarily aimed at reduction of plant costs and at an evolutionary enhancement of plant performance and safety. Two new 715-MWe CANDU-6 units with improvements over earlier versions of this model are under construction in Quinshan, China. Up-front basic engineering continues on the 935-MWe CANDU-9 reactor, a single unit adaptation of reactor units operating in Darlington, Canada. The two-year licensability review by the Canadian Nuclear Safety Commission was completed in Japan 1997, and found that the CANDU-9 meets the country's licensing
342
Chapter 7
requirements. Further studies are being carried out for advanced versions of these reactor models to incorporate further evolutionary improvements and to increase the output of the larger reactor up to 1300 MWe.
India Also under development is an advanced 500-MWe HWR in India, and construction of such units is planned. This HWR design takes advantage of experience feedback from the 220-MWe HWR plants of indigenous design operating in India. On the other hand, liquid metal-cooled fast reactors (LM-FRs), or breeders, have been under development for many years. With breeding capability, fast reactors can extract up to 60 times as much energy from uranium as can thermal reactors. The successful design, construction, and operation of such plants in several countries, notably France and the Russian Federation, has provided more than 200 reactor-years of experience on which to base further improvements. In the future, fast reactors may also be used to burn plutonium and other long-lived transuranic radioisotopes, allowing isolation time for high-level radioactive waste to be reduced. Significant activities are occurring in the development of high-temperature gascooled reactors (HTGRs), particularly with regard to the utilization of the gas-cooled reactor to achieve high efficiency in the generation of electricity and in process heat applications. Technological advances in component design and processes--coupled with the international capability to fabricate, test, and procure the components-provides an excellent opportunity for achieving HTGR commercialization. United Kingdom, Germany, and United States Gas-cooled reactors have been in operation for many years. In the United Kingdom, nuclear electricity is mostly generated in CO2-cooled Magnox and Advanced GasCooled Reactors (AGRs). Other countries also have pursued development of hightemperature reactors (HTGRs) with helium as coolant, and graphite as moderator. The 13-MWe AVR reactor has been successfully operated for 21 years in Germany demonstrating application of HTGR technology for electric power production. Other helium-cooled, graphite-moderated reactors have included the 300-MWe Thorium High Temperature Reactor in Germany, and the 40-MWe Peach Bottom and 330-MWe Fort St. Brain plants in the United States. South Africa In South Africa, the large national utility, Eskom, which has an installed generation capacity of about 38,000 MWe, is in the process of performing a technical and economic evaluation of a helium-cooled pebble bed module reactor. It would be directly coupled to a gas turbine power conversion system for consideration in increasing the capacity of the utility's electrical system. China and Japan In China and Japan, test reactors are under construction which will have the capability of achieving core outlet temperatures of 950~ for the evaluation of nuclear process
The Nuclear Fuel Cycle
343
heat applications. Construction of China's High Temperature Reactor (HTR-10) at the Institute of Nuclear Energy Technology (INET) continues with initial criticality anticipated for 1999. This pebble bed reactor of 10 MWth will be utilized to test and demonstrate the technology and safety features of the HTGR. Development of the HTGR by INET is being undertaken to evaluate a wide range of applications. They include electricity generation, steam and district heat production, combined steam and gas turbine cycle operation, and the generation of process heat for methane reforming. The HTR-10 is the first HTGR to be licensed and constructed in China (Juhn et al., 1997).
7.8 NUCLEAR FUSION A central issue for economic growth, prosperity and the quality of life in the industrialized world is the availability of secure, sustainable and financially competitive sources of energy. Given the expected growth in energy demand in the future, even with vigorous measures for energy savings, use will need to be made of all potential energy sources. The world Energy Council (WEC) projects growth in energy demand of anywhere between 50% and 300% over the next five decades, depending on environmental and economic factors (see Fig. 7.7). Strategic considerations favour the development of energy sources that offer greater sustainability and have less impact on health and the environment. Nuclear fusion, for which the fuel source is virtually limitless in quantity, could in the long term be an important option in this energy mix. There are several approaches to the problem of nuclear fusion. The most promising is definitely magnetic confinement fusion. In the course of the last 50 years research on magnetically confined plasmas has brought magnetic confinement fusion to the threshold of net power production and has revealed much of the physics underlying the complex behaviour of hot plasmas immersed in a magnetic field.
[Z
50 40
/i
/
30
O' ~
20
o,
lO 0
low !
!
i
1850
1900
1950
I,,
2000
i
i,
2050
2100
YEAR Fig. 7.7. Global energy use in gigatonne oil equivalent (Gtoe).
Chapter 7
344
The focus of contemporary fusion research is the deuterium-tritium reaction: 2H + 3H ----) 4He (3.5 MeV) + n (14.1 MeV)
(7.7)
which is the fusion reaction with the largest cross-section at the temperatures which are likely to be achieved in laboratory experiments (several 108 K). Eighty per cent of the reaction energy appears as the kinetic energy of the neutron, which would be absorbed in the structure of a power plant and provide most of the energy for steam generation. The o~-particle would be trapped in the plasma where its energy would heat the plasma and maintain the conditions required for fusion reactions to occur. Since tritium is a radioactive gas and since a high flux of 14.1 MeV neutrons would induce significant radioactivity in the structure surrounding the plasma, current experiments on magnetically confined plasma are usually carried out in hydrogen, so that no neutrons are produced, or in deuterium, for which the neutron production rate is almost two orders of magnitude lower than in a deuterium-tritium mixture. To achieve the conditions necessary for "ignition", where the m-particle power produced by fusion reactions exactly balances the heat loss due to transport processes, the plasma must be heated to a temperature of approximately 108 K at a particle density in the region of 10 ~~ions per cubic metre, while maintaining an energy replacement time of about 5 seconds. There are three different toroidal confinement configurations, each of them being a potential route to a possible fusion power plant: 1. Tokamak uses a strong toroidal field of several Tesla produced by a set of discrete coils. 2. The reversed field pinch (RFP) is a closely related configuration, since the plasma formation and ohmic heating are essentially identical to the tokamak. However, in the tokamak the average poloidal field is limited by stability requirements to approximately an order of magnitude smaller than the toroidal field, whereas the two are of similar magnitude in the RFP, both being typically less than 1 Tesla. 3. The third class of toroidal confinement devices is the stellarator which differs in an essential way from tokamaks and RFPs in that the helical fields are created entirely by coils external to the plasma, with no net toroidal current following within the plasma. In order for plasma to achieve ignition the product of plasma density, energyconfinement time and ion temperature, must reach a value of--5x102~ m -3 skeV. The increasing scale of magnetic confinement experiments, together with the accompanying improvements in the understanding of the physics of magnetoplasmas, has raised the values attained experimentally by 7 orders of magnitude since 1955 and has brought the field to the present point, where the largest experiments are within a factor of 5 of the required value. On this basis it can be expected that the parameters of the ITER tokamak are adequate to ensure that ignition will be achieved (Wesson, 1997). According to Campbell (1998), three principal conclusions can be drawn about the present status of magnetic confinement fusion. Firstly, there is now substantial, though
The Nuclear Fuel Q~,cle
345
still incomplete, understanding of plasma behaviour in the principal toroidal confinement configurations, and there is a much deeper appreciation of the complexity of the physics of high temperature magnetoplasmas. Secondly, new opportunities for further improvement in plasma performance are opening with the advent of a new generation of large stellarators such as LHD, the development of 'advanced tokamak scenarios", which may offer a viable route to steady-state tokamak operation, and realization of a variety of new tools for enhancing plasma performance in RFPs. Finally, given the production of over 10 MW of DT fusion power in TFTR and 16 MW in JET, plasma performance in tokamaks has advanced to the point where the construction of a DT-burning plasma experiment such as ITER would be a timely next step. An alternative to magnetic confinement is so-called inertial confinement fusion (ICF). The basic idea is to ignite and burn a few milligrams of deuterium-tritium fuel by means of high-power laser or ion beam pulses. Two large laser facilities are presently under construction which should demonstrate within the next 5-10 years the feasibility of single micro-explosions. These are the National Ignition Facility (NIF) in Livermore, US and the Laser MegaJoule (LMJ) in Bordeaux, France. In contrast to magnetic confinement fusion (MCF), inertial confinement involves no magnetic fields to contain the fuel, but relies exclusively on mass inertia. In ICF fusion burn occurs in highly compressed deuterium-tritium fuel, heated to an ignition temperature of 108 K. In the standard scheme compression and heating is achieved by spherical implosion of small capsules containing the fuel. A short pulse of radiation (laser, ion beam, or X-ray radiation) is used to ablate the outer layer of the capsule and to implode the inner part, driven by the ablation pressure like a spherical rocket. The energy yield of the ignited capsule (up to some 100 MJ) can be contained in a reactor vessel. For energy production the scheme implies pulsed operation with a few microexplosions per second. Presently, there are two paths to achieving uniform irradiation. Firstly, the direct drive approach, where a large number of overlapping beams is shone directly on the fusion capsule, and secondly, the indirect drive approach, where one converts the beam energy into X-rays which then drive the capsule implosion. At present, direct drive is thought to be possible only with lasers. The scientific and technological basis has been developed to ignite and burn micro-fusion targets by means of MJ laser pulses. In scaled experiments, implosions with high convergence ratio and neutron yields have been achieved, showing close agreement between experiment and multidimensional simulations. The crucial problem of symmetry and stability is approached along two lines, direct drive using laser smoothing techniques and indirect drive using gas-filled hohlraums (Lindl, 1995). Let us describe in some detail the European Union fusion programme (Bruhns, 1998). The starting point of this programme could be considered the creation of the European Atomic Energy Community (Euroatom) in 1957. Today, all EU member states have institutions actively participating in the fusion programme~all states except Greece participate through "association" contracts. The Community's own Joint Research Centre (JRC), which has institutes in various
346
oo
(-
g
I~
--
r:.
V
c~
~
r~ --
d
~
~ e,i
=
Z
~i
~
9-
E
o
c~
~.
._= --
~.'~ :.E
~
.~.
.~
"~ ~
V
"
V
~
"
~j
~ ~ --(:5
.-
V
oo ~
"-
"
~
~
A
_
9
,,=
o
..~
r.~
r
A
~
I=
~~
m
hr~ ~
I_
.,.
,-,
-'::
~ =
"~.,
=
~
o::1
~
~I~
~'.=
~
~
~6 c~
~
6
I~
~ N
r,--
l::z,
"~
ox
.-
I=
v
~
~
Chapter 7
c5
r.~
The Nuclear Fuel Cycle
347
locations, also undertakes work for the programme. Switzerland is fully associated to the programme (as Sweden was before it had become an EU Member State). Associations were established in Finland (1995) and in Austria (1996) after enlargement of the Union took in these countries. The associations are the backbone of the fusion programme. They operate a number of fusion devices in their laboratories (see Table 7.7). Most of these fusion devices have been built along the tokamak principle, but there are also stellarators and reversed field pinches. And there are a number of facilities for technological development such as large superconducting-magnet-testing facilities. At the end of the seventies it was decided to build, under the name of the JET Joint Undertaking, a fusion device (a tokamak) of much larger size than any fusion experiment existing at the time, JET, the Joint European Torus, located at Abingdon in the UK, began operation in 1983 and has become the flagship of the whole EU fusion programme. Around the same time, the Next European Torus team (NET) was established and given the task of enhancing the programme' s activities on safety and the environment, concentrating on the preparation (in particular the engineering and technological side) of the next-step experiment beyond JET. The NET team has become the pivotal point for initiating and coordinating R&D in fusion technology, as well as for Europe's contribution to the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER EDA) which was established in 1992 by the EU, Japan, Russia and the US.
REFERENCES Arkhipov, V., Future nuclear energy systems: generating electricity burning wastes. IAEA Bulletin, 39 (1997) 30. Blix, H., Nuclear energy in the 21st century. Nuclear News, September 1997, p. 34. Boudard, A., Leroy, S. and Volant, C., Spallation studies for nuclear waste transmutation. Nucl. Phys. News, 8 (1998) 18. Bowman, C.D., et al., Nuclear energy generation and waste transmutation using an accelerator driven intense thermal neutron source. Los Alamos Report LAUR-91-260 and Nuclear Instruments and Methods A320 (1992) 336. Bruhns, H., The EU fusion programme. Europhysics News, November/December 1998, p. 206. Campbell, D., Magnetic confinement fusion. Europhysics News, November/December 1998, p. 196. Carminati F. et al., preprint CERB/AT/93-47/ET (1993); C. Rubbia et al., preprint CERN/AT/95-44 (ET) (1995). Chan, C.Y., Radioactive waste management: an international perspective. IAEA Bulletin, 3 (1992) 7. Fells, I., The need for energy. Europhysics News, November/December 1998, p. 193. Flakus, F.-N. and Johnson, L.D., Binding agreements for nuclear safety: the global legal framework. IAEA Bulletin, 40 (1998) 21. IAEA Bulletin, 39 (1997) 13. Jones, S.R., Williams, S.M., Smith, A.D. and Gray, J., Review of discharge history and population doses from the Sellafield reprocessing plant in Cumbria, UK: The Sellafield environmental assessment model (SEAM), Reported at International Symposium on Environmental Impact of Radioactive Releases, Vienna, 8-12.05.1995, IAEA-SM-339/II.
348
Chapter 7
Juhn, P.-E., Kupitz, J., Cleveland, J.: Advanced nuclear power plants--high lights of global development. IAEA Bulletin, 39 (1997) 13. Kershaw, P.J., Woodhead, D.S., Malcolm, S.J., Allington, D.J. and Lovett, M.B., A sediment history of Sellafield discharges. J. Environ. Radioact., 12 (1990) 201-241. Lindl, J., Plasmas, 2 (1995) 3933. Nuclear Research Reactors in the World, Dec. 1996 Edition, Reference Data Series No. 3, International Atomic Energy Agency, Vienna, 1996. Oi, N., Plutonium challenges---changing dimensions of global cooperation. IAEA Bulletin, 40 (1998) 12. Safety Series No. 101: Operational Radiation Protection: A Guide to Optimization (1990). Safety Series No. 105: The Regulatory Process for the Decommissioning of Nuclear Facilities (1990). Safety Series No. 107: Radiation Safety of Gamma and Electron Irradiation Facilities (1992). Safety Series No. 108: Design and Operation of Radioactive Waste Incineration Facilities (1992). Safety Series No. 109: Intervention Criteria in a Nuclear or Radiation Emergency (1994). Safety Series No. 110: The Safety of Nuclear Installations (1993). Safety Series No. 111: The Principles of Radioactive Waste Management (1995). Safety Series No. 11 l-G- 1.1: Classification of Radioactive Waste (1994). Safety Series No. 11 l-G-3.1 : Siting of Near Surface Disposal Facilities (1994). Safety Series No. 11 l-G-4.1 : Siting of Geological Disposal Facilities (1994). Safety Series No. 11 l-S- 1: Establishing a National System for Radioactive Waste Management (1995). Safety Series No. 112: Compliance Assurance for the Safe Transport of Radioactive Material (1994). Safety Series No. 113: Quality Assurance for the Safe Transport of Radioactive Material (1994). Safety Series No. 120: Radiation Protection and the Safety of Radiation Sources (1996). Safety Series No. 35-G 1: Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report (1994). Safety Series No. 35-G2: Safety in the Utilization and Modification of Research (1994). Safety Series No. 50-C-Q: Quality Assurance for Safety in Nuclear Power Plants and other Nuclear Installations (1996). Safety Series No. 50-C-S (Rev. 1): Code on the Safety of Nuclear Power Plants: Siting (1988). Safety Series No. 50-SG-010: Core Management and Fuel Handling for Nuclear Power Plants (1985). Safety Series No. 50-SG-012: Periodic Safety Review of Operational Nuclear Power Plants (1994). Safety Series No. 50-SG-06: Preparedness of the Operating Organization (Licensee) for Emergencies at Nuclear Power Plants (1982). Safety Series No. 50-SG-DI: Safety Functions and Component Classification for BWR, PWR and PTR (1979). Safety Series No. 50-SG-D12: Design of the Reactor Containment Systems in Nuclear Power Plants (1985). Safety Series No. 50-SG-D2 (Rev. 1): Fire Protection in Nuclear Power Plants (1992). Safety Series No. 50-SG-D4: Protection Against Internally Generated Missiles and their Secondary Effects in Nuclear Power Plants (1980). Safety Series No. 50-SG-D5 (Rev. 1): External Man-induced Events in Relation to Nuclear Power Plant Design (1996). Safety Series No. 50-SG-D7: Emergency Power Systems at Nuclear Power Plants (1991). Safety Series No. 50-SG-D9: Design Aspects of Radiation Protection for Nuclear Power Plants (1985). Safety Series No. 50-SG-G4 (Rev: 1): Inspection and Enforcement by the Regulatory Body for Nuclear Power Plants (1996). Safety Series No. 50-SG-G6: Preparedness of Public Authorities for Emergencies at Nuclear Power Plants (1982). Safety Series No. 50-SG-G8: Licenses for Nuclear Power Plants: Content, Format and Legal Considerations (1982). Safety Series No. 50-SG-G9: Regulations and Guides for Nuclear Power Plants (1984). Safety Series No. 50-SG-O 1 (Rev. 1): Staffing of Nuclear Power Plants and the Recruitment, Training and Authorization of Operating Personnel (1991). Safety Series No. 50-SG-O4: Commissioning Procedures for Nuclear Power Plants (1980).
The Nuclear Fuel Cycle
349
Safety Series No. 50-SG-Q 1: Establishing and Implementing a Quality Assurance Programme (1996). Safety Series No. 50-SG-Q10: Quality Assurance in Design (1996). Safety Series No. 50-SG-Q 11: Quality Assurance in Construction (1996). Safety Series No. 50-SG-Q 12: Quality Assurance in Commissioning (1996). Safety Series No. 50-SG-Q 13: Quality Assurance in Operation (1996). Safety Series No. 50-SG-Q 14: Quality Assurance Decommissioning (1996). Safety Series No. 50-SG-Q2: Non-conformance Control and Corrective Actions (1996). Safety Series No. 50-SG-Q3: Document Control and Records (1996). Safety Series No. 50-SG-Q4: Inspection and Testing for Acceptance (1996). Safety Series No. 50-SG-Q5: Assessment of the Implementation of the Quality Assurance Programme (1996). Safety Series No. 50-SG-Q6: Quality Assurance in the Procurement of Items and Services (! 996). Safety Series No. 50-SG-Q7: Quality Assurance in Manufacturing (1996). Safety Series No. 50-SG-Q8: Quality Assurance in Research and Development (1996). Safety Series No. 50-SG-Q9: Quality Assurance in Siting (1996). Safety Series No. 50-SG-S 1 (Rev. 1): Earthquakes and Associated Topics in Relation to Nuclear Power Plants Siting (1991). Safety Series No. 50-SG-$8: Safety Aspects of the Foundations of Nuclear Power Plants (1986). Safety Series No. 50-SG-$9: Site Survey for Nuclear Power Plants (1984). Safety Series No. 79: Design of Radioactive Waste Management Systems at Nuclear Power Plants (1986). Safety Series No. 90: The Application of the Principles for Limiting Releases of Radioactive Effluents in the Case of the Mining and Milling of Radioactive Ores (1989). Safety Series No. 93: System of Reporting Unusual Events in Nuclear Power Plants. Safety Series No. 96: Guidance for Regulation of Underground Repositories for Disposal of Radioactive Wastes (1989). Safety Series No. 98: On-site Habitability in the Event of an Accident at a Nuclear Facility (1989). Semenov, B.A. and Oi, N., Nuclear fuel cycles: adjusting to new realities. IAEA Bulletin, 3 (1993) 2. Semenov, B.A., Disposal of spent fuel and high-level radioactive waste: building international consensus. IAEA Bulletin, 3 (1992) 2. Stather, J.W., Clarke, R.H. and Duncan, K.P., The risk of childhood leukaemia near nuclear establishments. NRPB-R215, HMSO, 1988. Stather, J.W., Dionian, J., Brown, J., Fell, T.P. and Muirhead, C.R., The risks of leukaemia and other cancers in Seascale from radiation exposure: Addendum to report R171. NRPB-R171 Addendum, HMSO, 1986. Stather, J.W., Wrixon, A.D. and Simmonds, J.R., The risks of leukaemia and other cancers in Seascale from radiation exposure, NRPB-R 171, HMSO, 1984. Takizuka, T., Proceedings of the 8th Journ6es SATURNE, Saclay, May 5-6, 1994. Van Tuyle, G.J. et al., Nucl. Technol., 101 (1993) 1. Wesson, J.A., Tokamaks, second edition. Oxford University Press, 1997. West, C.D., Research reactors: an overview. Nuclear News, October 1997, p. 50.
351
CHAPTER 8
The Bomb
8.1 INTRODUCTION Late in 1939 the possibility of using atomic energy for military purposes was brought to the attention of President Roosevelt, who appointed a committee to survey the problem. In June 1942 sufficient progress had been made to warrant a great expansion of the project and the assumption of its direction by the War Department with Major General Leslie R. Groves in executive charge. By December 1942 a decision had been reached to proceed with plant construction on a large scale, two of which were located at the Clinton Engineer Works in Tennessee and a third at the Hanford Engineer Works in the State of Washington. A special laboratory to deal with the many technical problems involved was located in an isolated area in the vicinity of Santa Fe, New Mexico, under the direction of Dr. J. Robert Oppenheimer. Certain other manufacturing plants much smaller in scale are located in the United States and Canada and the facilities of certain laboratories of the Universities of California, Chicago, Columbia, Iowa State College and at other schools as well as certain industrial laboratories were utilised. Up to 30 June 1945, Congress had appropriated a total of $1,950,000,000.00 for the operation of the huge project. The atomic bomb has been developed with the full knowledge of and cooperation of the United Kingdom and substantial patent control has been accomplished in the United States, the United Kingdom and Canada. Uranium is the essential ore in the production of the weapon and steps have been taken and will continue to be taken to ensure adequate supplies of this mineral. The series of discoveries which led to development of the atomic bomb started at the turn of the century when radioactivity became known to science. Prior to 1939 the scientific work in this field was worldwide, but more particularly in the United States, the United Kingdom, Germany, France, Italy and Denmark. One of Denmark's great scientists, Dr. Niels Bohr, a Nobel Prize winner, was whisked from the grasp of the Nazis in his occupied homeland and later assisted in developing the atomic bomb. It is known that Germany worked desperately to solve the problem of controlling atomic energy.
352
Chapter 8
Britain, suffering repeated air attacks early in the war, agreed to a concentration of the atomic bomb project in the United States and transferred many of her scientists to that country to assist. The attention of President Roosevelt was drawn to the potential of the atomic bomb in 1939. Research which had been conducted on a small scale with Navy funds was put on a greatly expanded basis. At the end of 1941 progress had been sufficient to warrant additional expansion. In the meantime the project had been placed under the direction of the Office of Scientific Research and Development with Dr. Vannevar Bush, Director of OSRD, in charge. At the same time the President appointed a General Policy Group, consisting of former Vice-President Henry A. Wallace, Secretary of War Henry L. Stimson, General George C. Marshall, Dr. James B. Conant, and Dr. Bush. The General Policy Group recommended in June 1942 that the atomic bomb project be greatly expanded and placed under the direction of the War Department. This action was taken and Major General Groves, experienced and resourceful U.S. Army construction engineer, placed in complete control. At the same time, in addition to the General Policy Group, there was appointed a Military Policy Committee consisting of Dr. Bush as chairman with Dr. Conant as his deputy, Lt. General Wilhelm D. Styer, USA, and Rear Admiral William R. Purnell, USN. The need for the weapon and its potential led to the decision in December 1942 to start the construction of an industrial empire that was to eventually consist of entire cities and employ upwards of 200,000. The ramifications of the atomic bomb project reached such proportions that in August 1943 it was decided to establish a Combined Policy Committee, composed at the outset of Secretary of War Stimson, Dr. Bush, Dr. Conant for the United States, Field Marshall Sir John Dill and Colonel J.J. Llewellin, for the United Kingdom; and Mr. C.D. Howe for Canada. Col. Llewellin was later replaced by Sir Ronald I. Campbell who in turn was succeeded by the Earl of Halifax; the late Field Marshal Dill was succeeded by Field Marshal Sir Henry Maitland Wilson. The United States members have had as their scientific adviser, Dr. Richard C. Tolman; the British, Sir James Chadwick; and the Canadian, Dean C.J. Mackenzie. As a curiosity we shall here describe a set of five lectures given by Serber (1943) during the first two weeks of April 1943 as an "introduction course" in connection with the starting of Los Alamos project. The notes were written by E.U. Condon. According to the notes, the object of the project was to produce a practical military weapon in the form of a bomb in which the energy is released by a fast neutron chain reaction in one or more of the materials known to show nuclear fission. The direct energy release in the fission process is of the order of 170 MeV per atom. This is considerably more than 10 times the heat of reaction per atom in ordinary combustion processes. This is 170x106x4.3x10-1~ = 2.7x10 -4 erg/nucleus. Since the weight of 1 nucleus of 235U is 3.88x 10-22 gram/nucleus the energy release is 7x 10 ~7 erg/gram. The energy release in TNT is 4• 10 ~~org/gram or 3.6x 10 ~~erg/ton. Hence 1 kg of 235U ~- 20000 tons of TNT.
The Bomb
35 3
Release of this energy in a large-scale way is a possibility because of the fact that in each fission process, which requires a neutron to produce it, two neutrons are released. Consider a very great mass of active material, so great that no neutrons are lost through the surface and assume the material so pure that no neutrons are lost in other ways than by fission. One neutron released in the mass would become 2 after the first fission, each of these would produce 2 after they each had produced fission so in the nth generation of neutrons there would be 2 n neutrons available. Since in 1 kg of 235U there a r e 5 • 25 nuclei it would require about n = 80 generations (2 8~ 5x10 25) to fission the whole kilogram. While this is going, on the energy release makes the material very hot, developing great pressure and hence tending to cause an explosion. In a natural finite setup, some neutrons are lost by diffusion out through the surface. There will be therefore a certain size of say a sphere for which the surface losses of neutrons are just right to still sustain a chain reaction. This radius depends on the density. As the reaction proceeds the material tends to expand, increasing the required minimum size faster than the actual size increases. The whole question of whether an effective explosion is made depends on whether the reaction is stopped by this tendency before an appreciable fraction of the active material has finished. Note that the energy released per fission is large compared to the total binding energy of the electrons in any atom. In consequence even if but 0.5% of the available energy is released the material is very highly ionised and the temperature is raised to the order of 40• 10 6 degrees. If 1% is released the mean speed of the nuclear particles is of the order of 10 8 cm/s. Expansion of a few centimetres will stop the reaction, so the whole reaction must occur in about 5• 10 -8 s otherwise the material will have blown out enough to stop it. It is just possible for the reaction to occur to an interesting extent before it is stopped by the spreading of the active material. Slow neutrons cannot play an essential role in an explosion process since they require about a microsecond to be slowed down in hydrogenic materials and the explosion is all over before they are slowed down. The materials in question are 235U, 238U, 239ptl and some others of lesser interest. Ordinary uranium as it occurs in nature contains about 1/140 or 235U, the rest being 238U except for a very small amount of 234U. The nuclear cross-section for fission of the two kinds of U and/or 239ptl a r e shown roughly in Fig. 8.1 where ~ is plotted against the log ~ the incident neutron' s energy. We see that 235U has a cross-section of about cyI --- 1.5• 10 -24 c m 2 for neutron energies exceeding 0.5 MeV and rises to much higher values at low neutron energies (oj--640• 10 -24 cm 2 for thermal neutrons). For 238U, however, threshold energy of 1 MeV occurs below which cyI -- 0. Above the threshold cys is fairly constant and equal to 0.7• 10 -24 cm 2. The energy distribution of the neutrons released in the fission process is shown in Fig. 8.2. The mean energy is about 2 MeV but an appreciable fraction of the neutrons released have less than 1 MeV of energy and so are unable to produce fission in 238U.
354
Chapter 8 I
10
I
I
I
I
I
I
I
I
I
-~1
1 0 -22 l 0 -23
~
"~
10.24
239pu
~"~ " ~ 235 - U-
3.0 1.5
] 0 .25
I
I
I
I
I
I
I
I
-2
-1
0
1
2
3
4
5
S
6
0.7
I
7
LOG NEUTRON ENERGY / eV Fig. 8.1. Fission cross section for 235U, 238U and 239pu as a function of neutron energy.
Z 0
Z 0
R
Z
about 2 MeV 0
I 1
,
I 2
1 3
NEUTRON ENERGY / MeV Fig. 8.2. Energy distribution of neutrons produced in fission.
One can give a quite satisfactory interpretation o f the energy distribution of Fig. 8.2 by supposing it to result from evaporation of neutrons from the fission product nuclei with a temperature of about 0.5 MeV. Such a Maxwellian velocity distribution is to be relative to the moving fission product nuclei giving rise to a curve like Fig. 8.2. The average n u m b e r of neutrons produced per fission is denoted by v. At that time it was not k n o w n whether v has the same value for fission processes in different materials, induced by fast or slow neutrons or occurring spontaneously. The best value at present is v = 2.2 +_0.2 although a value v = 3 has been reported for spontaneous fission. (Keep in mind these notes were produced in 1943 !). In these notes,
The Bomb
355
isotopes and elements U and Pu are called by symbols 25, 28 and 49. The notes continue with the following text. When neutrons are in uranium they are also caused to disappear by another process represented by the equation 28 + n ~ 29 + ~/. The resulting element 29 undergoes two successive 13transformations into elements 39 and 49. The occurrence of this process in 28 acts to consume neutrons and works against the possibility of a fast neutron chain reaction in material containing 28. It is this series of reactions, occurring in a slow neutron fission pile, which is the basis of a project for large-scale production of element 49. Based on this simple physics it was possible to perform a simple estimate of the minimum size of the bomb. Assuming a spherical geometry of radius, R c, it was estimated that R c = 13.5 cm yielding a critical volume of 1 0 . 5 • 3 c m 3 having a critical mass of about 200 kg. The value of critical mass, M c, calculated in this way is, however, considerably overestimated by the elementary diffusion theory. The more exact diffusion theory, allowing for the long free path, drops R c by a factor of about 2/3 giving R C-- 9 cm and M e - 60 kg of 235U. If one surrounds the core of active material by a shell of inactive material the shell will reflect some neutron active material will be enough to give rise to an explosion. The surrounding case is called a tamper. The tamper material serves not only to retard the escape of neutrons but also by its inertia to retard the expansion of the active material. (The retardation provided by the tensile strength of the case is negligible.) For the latter purpose it is desirable to use the densest available materials (Au, W, Rc, U). Present evidence indicates that for neutron reflecting properties also, one cannot do better than use these heavy elements. Needless to say, a great deal of work has been done afterwards on the properties of tamper materials. The introduction of a tamper reduces the critical mass by a factor of at least 4 giving R c -- 6 cm and
M c
-- 15 kg of
235U.
In the case of Pu (or mass 49 as it is called in the original document) M~ is, because of its larger cross section, smaller by about a factor of 3 resulting in M c -- 5 kg. The discussion of the damage is as follows: several kinds of damage will be caused by the bomb. A very large number of neutrons will be released in the explosion. One can estimate a radius of about 1000 yards around the site of explosion as the size of the region in which the neutron concentration is great enough to produce severe pathological effects. Enough radioactive material is produced that the total activity will be of the order of 106 curies even after 10 days. Just what effect this will have in rendering the locality uninhabitable depends greatly on very uncertain factors about the way in which this is dispersed by the explosion. However the total amount of radioactivity produced, as well as the total number of neutrons, is evidently proportional just to the number of fission processes, or to the total energy released.
356
Chapter 8
The mechanical explosion damage is caused by the blast or shock wave. The explosion starts acoustic waves in the air which travel with the acoustic velocity, c, superposed on the velocity u of the mass motion with which material is convected out from the centre. Since c -- ~ where Tis the absolute temperature and since both u and c are greater farther back in the wave disturbance it follows that the back of the wave overtakes the front and thus builds up a sharp front. This is essentially discontinuous in both pressure and density. It has been shown that in such a wave front the density just behind the front rises abruptly to six times its value just ahead of the front. In back of the front the density falls down essentially to zero. If E is the total energy released in the explosion it has been shown that the maximum value of the pressure in the wave front varies as p -. E ] r 3
(8.1)
the maximum pressure varying a s I / r 3 instead of the usual I / r 2 because the width of the strongly compressed region increases proportionally to r. This behaviour continues as long as p is greater than about 2 atmospheres. At lower pressures there is a transition to ordinary acoustic behaviour, the width of the pulse no longer increasing. If destructive action may be regarded as measured by the maximum pressure amplitude, it follows that the radius of destructive action produced by an explosion varies as 3V~-.Now in a 0.5 ton bomb, containing 0.25 ton of TNT, the destructive radius is of the order of 150 feet. Hence in a bomb equivalent to 100,000 tons of TNT (or 5 kg of active material totally converted) one would expect a destructive radius of the order of ~100000 feet or about 2 miles. These were roughly the effects expected from the device which was built in Los Alamos. Detonation of the bomb was a much-discussed process during these early days. Before firing, the active material must be disposed in such a way that the effective neutron number v' is less than unity. The act of firing consists in producing a rearrangement such that after the rearrangement v' is greater than unity. This problem is complicated by the fact that, as can be seen, one needs to deal with a total mass of active material considerably greater than the critical in order to get appreciable efficiency. For any proposed type of rearrangement one may introduce a co-ordinate ~ which changes from 0 to 1 as the rearrangement of parts proceeds from its initial to its final value. Schematically, v' will vary with Z along some such curve. Since the rearrangement proceeds at a finite speed there will be a finite time interval during which v', though positive, is much smaller than its final value. There will always be some unavoidable sources of neutrons in the active material. In any scheme of rearrangement some fairly massive amount of material will have to be moved a distance of the order of R c -- 10 cm. Assuming a speed of 3000 ft/s can be imparted with some type of gun, this means that the time it takes to put the pieces of the bomb together is --10-4 s. Since the whole explosion is over in a time --75 T / v " = 10-6/v' s, one can see that, except for very small v'(v' << 0.01), an explosion started by a premature neutron will be all finished
The Bomb
357
before there is time for the pieces to move an applicable distance. Thus if neutron multiplication happens to start before the pieces reach their final configuration an explosion will occur that is of lower efficiency corresponding to the lower value at the instant of explosion. To avoid predetonation it is therefore necessary to keep the neutron background as low as possible and to effect the rearrangement as rapidly as possible. Since it will be clearly impossible to reduce the neutron background rigorously to zero, there will always be some chance of predetonation. This is discussed in the following paragraphs in order to see how this affects the firing problem. The chance of predetonation is dependent on the likelihood of a neutron appearing in the active mass while v' is still small and on the likelihood that such a neutron will really set off a chain reaction. With just a single neutron released when v' > c it is by no means certain that a chain reaction will start, since any particular neutron may escape from the active material without causing a chain reaction. The importance of taking great pains to get the least possible neutron background, and of shooting the firing rearrangement with the maximum possible velocity is stressed. It seems one should strive for a neutron background of 10000 neutron/s or less and firing velocities of 3000 ft/s or more. Both of these are difficult to attain. To avoid predetonation one must make sure that there is only a small probability of a neutron appearing while the pieces of the bomb are being put together. On the other hand, when the pieces reach their best position one wants to be very sure that a neutron starts the reaction before the pieces have a chance to separate or break. It may be possible to make the projectile seat and stay in the desired position. Failing this, or in any event as extra insurance, another possibility is to provide a strong neutron source which becomes active as soon as the pieces come into position. For example one might use a Ra+Be source in which the Ra is on one piece and the Be on the other so neutrons are only produced when the pieces are close to the proper relative position. One can easily estimate the strength of source required. After the source starts working, one wants a high probability of detonation before the pieces have time to move more than, say, 1 cm. This means that N, the neutrons/s from the source must be large enough that
1 Nunu' - ~ > 1 (say = 10) 2 v
(8.2)
N = 107 neutrons/s. A source of this strength that can be activated within 10 -5 seconds and is mechanically rugged enough to stand a shock associated with firing, is easily achievable today with the sealed tube neutron generators. There are three recognised sources of neutrons which provide the background which gives rise to danger of predetonation: (a) cosmic ray neutrons, (b) spontaneous fission, (c) nuclear reactions which produce neutrons. In this context one has to consider (c~, n) reactions and light elements which might be present as impurities. One can base some rough guesses on the standard barrier penetration formulas and find the
358
Chapter 8
Fig. 8.3. Autocatalytic methods: (a) active material is disposed in a hollow shell; (b) boron bubble scheme.
following upper limits on the concentration by weight for several light elements for production of 104 neutron/s. The effect of several impurities simultaneously present is of course additive. Preparation and handling of 239puin such a way as to attain and maintain such high standards of purity is a difficult problem. With 235U the situation is more favourable. It is of interest to present here ideas which existed at that time on the subjects of mechanism of shooting and in particular "autocatalytic method" (Serber 1943): We now consider briefly the problem of the actual mechanics of shooting so that the pieces are brought together with a relative velocity of the order of 105 cm/s or more. This is the part of the job about which least was known at that time. One way is to use a sphere and to shoot into it a cylindrical plug made of some active material and some tamper, as in Fig. 8.3. This avoids fancy shapes and gives the most favourable shape for shooting to the projected piece whose mass would be of the order of 100 lbs. The highest muzzle velocity available in U.S. Army guns is one whose bore is 4.7 inches and whose barrel in 21 ft long. This gives a 50 lb projectile a muzzle velocity of 3150 ft/s. The gun weighs 5 tons. It appears that the ratio of projectile mass to gun mass is about constant for different guns, so a 100 lb projectile would require a gun weighing about 10 tons. The weight of the gun varies very roughly as the cube of the muzzle velocity hence there is a high premium on using lower velocities of fire. Another possibility is to use two guns and to fire two projectiles at each other. For the same relative velocity this arrangement requires about 1/8 as much total gun weight. Here the worst difficulty lies in timing the two guns. This can be partly overcome by using an elongated tamper mass and putting all the active material in the projectiles so it does not matter exactly where they meet. At that time it would have been possible to synchronise so the spread in places of impact on various shots would be 2 or 3 feet. One serious restriction imposed by these shooting methods is that the mass of active material that can be gotten together is limited by the fact that each piece separately must be non-explosive. Since the separate pieces are not of the best shape, nor surrounded by the best tamper material, one is not limited to two critical masses for the completed
The Bomb
359
bomb, but might perhaps get as high as four critical masses. However in the two gun scheme, if the final mass is to be --4 M c, each piece separately would probably be explosive as soon as it entered the tamper, and better synchronisation would be required. It seems worthwhile to investigate whether present performance might not be improved by a factor of 10. Severe restrictions on the mass of the bomb can be circumvented by using pieces of a shape more difficult to shoot. For example, a flat plate of actual material tamped on only one side, has a minimum thickness below which it can no longer support a chain reaction, no matter how large its area, because of neutron leakage across the untamped surface. If two such plates were slid together, untamped surfaces in contact, the resulting arrangement could be well over the critical thickness for a plate tamped on both sides, and the mass would depend only on the area of the plates. Calculations show that the critical mass of a well tamped spheroid, whose major axis is five times its minor axis, is only 35% larger than the critical mass of a sphere. If such a spheroid 10 cm thick and 50 cm in diameter were sliced in half, each piece would be sub-critical though the total mass, 250 kg, is 12 times the critical mass. The efficiency of such an arrangement would be quite good, since the expansion tends to bring the material more and more nearly into a spherical shape. Thus there are many ordnance questions one would like to have answered. One would like to know how well guns can be synchronised. One needs information about the possibilities of firing other than cylindrical shapes at lower velocities. Also, one needs to know the mechanical effects of the blast wave preceding the projectile in the gun barrel. Also whether the projectile can be made to seat itself properly and whether a piston of inactive material may be used to drive the active material into place, this being desirable because thus the active material might be kept out of the gun barrel, which to some extent acts as a tamper. Various other shooting arrangements have been suggested. For example it has been suggested that the pieces might be mounted on a ring, as in Fig. 8.4. If explosive material were distributed around the ring and fired, the pieces would be blown inward to form a sphere. Another more likely possibility is to have the sphere assembled but with a wedge of neutron-absorbing material built in, which on firing would be blown out by an explosive charge causing v' to go from less than unity to more than unity. Here the difficulty lies in the fact that no material is known whose absorption coefficient for fast neutrons is much larger than the emission coefficient of the bomb material. Hence the absorbing plug will need to have a volume comparable to that of the absorb or and when removed will leave the active material in an unfavourable configuration, equivalent to a low mean density. The term "autocatalytic method" is being used to describe any arrangement in which the motions of material produced by the reaction will act, at least for a time, to increase v' rather than to decrease it. Evidently if arrangements having this property can be developed they would be very valuable, especially if the tendency toward increasing v' was possessed to any marked degree.
360
Chapter 8
Fig. 8.4. Three possibilities of autocatalytic schemes: (i) hollow shell, (ii) two pieces, (iii) four pieces.
Suppose we had an arrangement in which, for example, v' would increase of its own accord from a low value like 0.01 up to a value 10-50 times greater. The firing problem would be simplified by the low initial value of v', and the efficiency would be maintained by the tendency to develop a high value of v' as the reaction proceeds. It may be that a method of this kind will be absolutely essential for utilisation of 49 owing to the difficulties of high neutron background from (or, n) reactions with the impurities. The simplest scheme which might be autocatalytic is indicated in the sketch where the active material is disposed in a hollow shell as indicated in Fig. 8.3a. Suppose that when the firing plug is in place one has just the critical mass for this configuration. If as the reaction proceeds the expansion were to proceed only inward it is easy to see from diffusion theory that v' would increase. Of course in actual fact it will proceed outward (tending to decrease v') as well as inward and the outward expansion would in reality give the dominant effect. However, even if the outward expansion were very small compared to the inward expansion, it has been calculated that this method gives very low efficiency: with 12 M~ an efficiency of only about 10-9 was calculated. A better arrangement is the "boron bubble" scheme (see Fig. 8.3b). B ~~ has the largest known absorption cross-section for fast neutrons, 1.52x 10-24 cm 2. Suppose one takes a large mass of active material and puts in enough boron to make the mass just critical. The device is then fired by adding some more active material or tamper. As the reaction proceeds the boron is compressed and is less effective at absorbing neutrons than when not compressed. This can be seen most readily if one considers the case in which the bubbles are large compared to the mean depth to which a neutron goes in boron before being absorbed. Then their effectiveness in removing neutrons will be
The Bomb
361
proportional to their total area and so will drop on compression. Hence v' will increase as the bubbles are compressed. If the bomb is sufficiently large this tendency is bound to overweigh the opposing one due to the general expansion of the bomb material, since the distance the edge of the bomb must move to produce a given decrease in v' increases with the radius of the bomb, whereas for a larger bomb the distance the edge of a bubble must move is unchanged, since it is not necessary to increase the radius of the bubbles but only to use more of them. The density of particles (electrons plus nuclei) in boron is 8 . 3 • 23 particle/cm 3 while in uranium it is more than five times greater. Therefore as soon as the reaction has proceeded to the point where there is a high degree of ionisation and the material behaves as a gas there will be a great action to compress the boron. An opposing tendency to the one desired will be the stirring or turbulence acting to mix the boron uniformly with the uranium, but the time scale is too short for this to be effective. It can be shown that if initially v' = 0, allowing for the boron absorption, and if no expansion of the outer edge occurs then v' will rise to v' -- )/2(v - 1) by compression of the boron. This scheme requires at least five times the critical mass for no boron, and the efficiency is low unless considerably more is used. If one uses just that amount of boron which makes twice the no-boron critical mass be just critical, then the efficiency is lower by a factor of at least 30. All autocatalytic schemes that have been thought of so far require large amounts of active material, are low in efficiency unless very large amounts are used, and are dangerous to handle. Some bright ideas are needed; see Fig. 8.4 for some proposals. The initial combat use of the bomb was the culmination of three years of intensive effort on the part of science and industry, working in cooperation with the Military. In the USA it was heralded as the greatest achievement of the combined efforts of science, industry, labour and the military in history. President Truman and Secretary of War Henry L. Stimson made the first announcements of the new weapon, declaring that the atomic bomb had an explosive force such as to stagger the imagination. Improvements were revealed as forthcoming which would increase several-fold the effectiveness accomplished at that time. The schematic sketches based on the photographs of the bombs dropped on Nagasaki ("Fat Man Bomb") and Hiroshima "Little Boy Bomb") are shown in Figs. 8.5 and 8.6. The detailed information about the mission summaries and strike aircrafts are shown in Table 8.1, while the details of the mission summaries for instrument and photo aircrafts are shown in Tables 8.2 and 8.3. While the use in combat permitted a slight relaxation in the security that cloaked the project, the War Department declined for security reasons to disclose the exact methods by which the bombs are produced or the nature of their action and requested that the American press and radio, as well as all those connected with the project, refrain from disclosing information. Security remained rigorous. The Headquarters of First Technical Service detachment in San Francisco, California issued the following security requirements in August 8, 1945:
Chapter 8
362
Fig. 8.6. Schematic sketch of the bomb dropped on Hiroshima, "Little Boy Bomb".
Table 8.1 Concentrations of elements which can produce 104 n/s when irradiated by Ra+Be source Element
Concentration
Li
2x10 -5
Be
10-6
B
2x10-6
C
2x10 -4
Low yield because only C 13 contributes
O
2x 10-3
Low yield because only 017 contributes
F
2x10 -5
N
(a-n) reaction not energetically possible
"1. Although the President of the United States has announced the existence of a military project concerned with the use of atomic power, and has also announced the first combat use of such weapons, it should be clearly evident to all project personnel that intelligent security concerning certain phases of project activities remains a necessity. General principles pertaining to categories of information still of a classified nature are set out below; protection of project information depends, more than ever, upon the good judgement and discretion of the individual possessing the information.
363
The Bomb
Table 8.2 Mission summaries, strike aircraft Hiroshima
Nagasaki
Bomb Designation
L-11 (Little Boy)
F-31 (Fat Man)
Mission number
13
16
Strike aircraft
V-82 (Enola Gay)
V-77 (Bock's Car)
Aircraft Commander
Col. P.W. Tibbets
Maj. C.W. Sweeney
Pilot
Capt. R.A. Lewis
1st Lt. C.D. Albury
Navigator
Capt. T.T. Van Kirk
Capt. J.F. Van Pelt
Bombardier
Maj. T.W. Ferebee
Capt. K.K. Beahan
Weaponeer
Capt. W.S. Parsons (USN)
Cdr. F.L. Ashworth (USN)
Time of detonation (Japan Time)
0815 August 6, 1945
1158 August 9, 1945
Indicated air speed
200 mph
200 mph
True air speed
328 mph
315 mph
Wind
8 kts at 170 ~
1 kt head wind
True heading
262 ~
True course
265 ~
Indicated altitude
30,200 ft
True altitude
32,700 ft*
Temperature
Ind.-22~
Time of fall
28,000 ft 28,900 fy True-33~
45.5 s
47.7 s
*From Parsons' log, corrected pressure altitude 32,200 ft.
Table 8.3 Mission summaries, instrument and photo aircraft Hiroshima
Nagasaki
Instrument aircraft
V-89 (Great Artiste)
V-89 ( Great Artiste)
Position
300 ft
300 ft or 0.5 mile
Aircraft commander
Maj. C.W. Sweeney
Capt. F. Bock
Bombardier
Capt. K.K. Beahan
Scientists/Observers
L.W. Alvarez; H.M. Agnew; L. Johnston
W. Goodman; J. Kupferberg; L. Johnston; W. Laurence (NY Times)
Photo aircraft
V-91
V-90 (Full House)
Aircraft commander
Capt. Marquardt
Maj. J. Hopkins
Scientists
Observers
B. Waldman
364
Chapter 8
2. Additional information on the publicity status of the project will be issued from this office from time to time, but for the present the information released by the President and officially by the War Department contains all information which is presently authorised for dissemination. 3. Some of the categories which remain of a classified nature are: a. Information pertaining to the unit design and details, and character of and details concerning the critical material. b. Production schedules of critical material, actual or proposed. c. Proposed tactical uses and/or schedules. d. Results of tests employing the project weapons or dummies thereof. e. Data concerning past tactical uses. f. Communication codes peculiar to the project. g. Information pertaining to air and water shipments of project supplies. h. All information pertaining to shipments of critical materials. i. Specific contributions with regard to engineering or design details, or other information contributed by various project installations or by specific project individuals. j. The operational inter-relationship of project sites. k. The association of tactical units or the names of tactical personnel engaged in the combat delivery of the weapon to the enemy. 4. In connection with the publicity which has been released concerning the project, the mere mention of atomic bombs or atomic power is no longer classified, unless used in connection with the categories listed above or in connection with other information which would result in a disclosure greater than that made by the publicity releases indicated in paragraph 1. 5. For the information of all concerned the following definitions, quoted from AR 380-5, are set forth: a. Top Secret. Certain secret documents, information, and material, the security aspect of which is paramount, and whose unauthorised disclosure would cause exceptionally grave damage to the nation shall be classified TOP SECRET. b. Secret. Documents, information, or material, the unauthorised disclosure of which would endanger national security, cause serious injury to the interests or prestige of the nation, or any governmental activity thereof, or would be of great advantage to a foreign nation shall be classified SECRET. c. Confidential. Documents, information, or material, the unauthorised disclosure of which, while not endangering the national security, would be prejudicial to the interests or prestige of the nation, any governmental activity, an individual, or would cause administrative embarrassment, or difficulty, or be of advantage to a foreign nation shall be classified CONFIDENTIAL. d. Restricted. Documents, information or material (other than TOP SECRET, SECRET, or CONFIDENTIAL) which should not be published or communicated to anyone except for official purposes shall be classified RESTRICTED.
The Bomb
365
6. With regard to the reclassification of documents, persons possessing documents which they feel should be declassified will bring them to Col. de Silva for reclassification authority. 7. All project personnel returning to the United States will, before departure, submit to the Detachment Orderly Room, all developed negatives and prints which have been taken since leaving the United States. No exposed but undeveloped rolls of film may be taken or returned to the United States. Adherence to this rule will, in addition to protecting project security, enable the individual to avoid difficulties with Customs Officials in Honolulu. 8. Project personnel are advised that they will not allow themselves to be interviewed by press correspondents, radio commentators, or other individuals engaged in releasing news or background stories to the general public, except when officially authorised and arranged. This covers representatives of military or naval publications as well. Requests for such interviews will be referred at once to Col. de Silva. Project personnel are reminded that, prior to publication of personal notes or manuscripts, such material must be submitted to a project officer for approval. Such notes will be referred to Col. de Silva."
8.2 H E A L T H EFFECTS OF THE ATOMIC BOMB
Nuclear devices are basically of two types, fission (the "atomic" bomb) and fusion (the thermonuclear or "hydrogen" bomb). Fission of plutonium-239 or uranium-235 produces over 100 radioisotopes with half-lives varying from fractions of a second to millions of years. Other radioisotopes are produced by neutron absorption in the fuel or surrounding materials (these are termed activation products). Fusion of light elements (isotopes of hydrogen) produces smaller quantities of radioisotopes, mainly activation products, but requires a fission device to trigger the fusion reaction. Small nuclear explosions are usually produced by fission while larger explosions generally involve both fission and fusion. Table 8.4 lists the half-lives of the more important radioisotopes produced by nuclear testing. This table is organised in increasing order of half-life of the isotopes. One should keep in mind that any increase in radioactivity above natural background levels requires an assessment of possible effects on human health. Severe exposure to radiation causes immediate damaging health effects and, frequently, death. At much lower levels, radiation exposure can initiate cancer, which emerges 10-20 years later. Exposure to radioactivity from contaminants released in a nuclear explosion can occur via several pathways. These include external irradiation, and internal irradiation by inhalation of airborne particulates and ingestion of food or water containing radioactivity. The consequences of the radiation dose and the level of risk will depend on how much of the body and what organs or tissues are irradiated, over what period exposure occurs, and whether one individual or a large number of people are exposed.
Chapter 8
366
Table 8.4 Radioisotopes produced from underground nuclear explosions (only isotopes with half-lives greater than four hours are shown) Radioisotope
Half- life
Ori gin
Krypton-85m
4.5 hours
noble gas (fission)
Xenon- 135
9 hours
noble gas (fission)
Sodium-24
15 hours
activation
Zirconium-97
17 hours
fssion
Iodine- 133
21 hours
fission
Cerium- 143
33 hours
fission
Rhenium- 105
35 hours
fission
Xenon-133m
2 days
noble gas (fission)
Neptunium-239
2.4 days
activation
Molybdenum-99
2.9 days
Xenon-133
5 days
fission noble gas (fission)
Uranium-237
7 days
activation
Iodine- 131
8 days
fission
Neodymium- 147
11 days
fission
Xenon-13 lm
12 days
noble gas (fission)
Barium- 140
13 days
fission
Cerium- 141
33 days
fission
Tellurium- 129m
34 days
fission
Niobium-95
35 days
fission
Ruthenium- 103
39 days
fission
Iron-59
45 days
activation
Strontium-89
54 days
fission
Yttrium-91
59 days
fission
Zirconium-95
64 days
fission
Cerium- 144
285 days
fission
Manganese-54
313 days
activation
Ruthenium- 106
372 days
fission
Antimony- 125
2.7 years
fission
Iron-55
2.7 years
activation
Europium- 155
4.8 years
fission
Cobalt-60
5 years
activation
Krypton-85
10.8 years
noble gas (fission)
Tritium
12 years
fission, fusion, activation
Plutonium-241
15 years
weapon, activation
Strontium-90
29 years
fission
Caesium- 137
30 years
fission
Plutonium-238
86 years
activation
367
The Bomb
Radioisotope
Half-life
Origin
Americium-241
433 years
weapon
Plutonium-240
6540 years
weapon, activation
Plutonium-239
24.000 years
weapon
Technetium-99
210.000 years
fission
Chlorine-36
300.000 years
activation
Neptunium-237
2 million years
activation
Caesium- 135
3 million years
fission
Iodine- 129
16 million years
fission
Uranium-235
700 million years
weapon
Uranium-238
4.5 billion years
weapon
Of the large number of radioisotopes produced in a nuclear explosion, most are either present in small amounts or decay quickly. A small number of long-half-life radioisotopes could have a potential longer-term effect on human health (e.g. those towards the bottom of Table 8.4). The short-lived radioactive components in fallout from an atmospheric nuclear explosion, comprising mainly particulate fission products, can contribute significantly to radiation exposures in humans either by external irradiation due to material deposited on the ground or by uptake of radioactive iodine from food, especially milk. Numerous studies have been done on the effects of the atomic bombings of Hiroshima and Nagasaki; these studies have led to the most recent modifications of Basic Safety Standards.
8.3 WEAPON PRODUCTION The processes involved in weapon production as related to the nuclear fuel cycle are presented schematically in Fig. 8.7. It should be kept in mind that a significant quantity of the material needed for a single, relatively simple nuclear device is: plutonium: 5-8 kg enriched uranium: 25 kg It has been estimated that the time needed for a country with the appropriate industrial infrastructure and pre-assembly activities to convert a significant quantity of fissionable material into a nuclear explosive device is: 1. for significant quantities of plutonium or uranium-235 metal or oxides: days to weeks; 2. for spent fuel or low-enriched uranium: several months to a year. More sophisticated devices use less plutonium; it is assumed that the French bombs tested at Mururoa and Fangataufa Atolls used a plutonium pit of about 3.7 kg, 3.4 kg of which w a s 239pu and 0.3 kg of which w a s 24~
368
Chapter 8
0
0
~=
z
~
The Bomb
369
An amount of 3.7 kg plutonium is also reported as typical for pits in U.S. and USSR devices. Corroborating evidence is given in a Russian report on early tests (Dubasov et al., 1995) for example, which describes many devices with 3-4 kg of plutonium, and in press reports that refer to nearly 3 kg of plutonium in each of 32 USSR nuclear warheads in a sunken submarine east of Bermuda. Similar values have been reported for the two torpedoes that sank with the Russian submarine Komsomolets north of Norway in 1989, and for the U.S. weapons involved in the air crashes at Thule and Palomares in the 1960s. The critical mass of, for example, a sphere of pure plutonium-239 metal in its densest form (alpha-phase, density 19.8 g/cm) is about 10 kg. The radius of the sphere is about 5 cm, about the size of a small grapefruit. If the plutonium sphere were surrounded by a natural uranium neutron reflector, about 4.4 kg, the radius of the sphere would be about 3.6 cm, about the size of an orange. A 32 cm thick beryllium reflector reduces the critical mass to about 2.5 kg, a sphere with a radius of 3.1 cm, about the size of a tennis ball. Using a cunning technique called implosion, in which conventional chemical explosives are used to produce a shock wave which uniformly compresses the plutonium sphere, the volume of the plutonium sphere can be slightly reduced and its density increased. If the original mass of the plutonium is just less than critical it will, after compression, become super-critical and a nuclear explosion will take place. Using implosion, a nuclear explosion could, with a good modem design including an effective, practicable reflector, be achieved with about 2.5 kg of plutonium. The trick is to obtain very uniform compression of the sphere. In such a design, the plutonium would be surrounded by a spherical shell, called a tamper, made from a heavy metal, like natural uranium. The tamper has two functions: first to reflect back into the plutonium some of the neutrons which escaped through the surface of the plutonium core to minimise the mass of plutonium needed; second, because the tamper is made of heavy metal, its inertia helps hold together the plutonium during the explosion to prevent the premature disintegration of the fissioning material and thereby obtain a greater efficiency. Most modem nuclear weapons are thermonuclear devices that utilise a plutonium pit (hollow sphere) as the primary trigger to ignite a secondary stage containing a fusion package. In the secondary, the thermonuclear fuel is wrapped in a heavy metal blanket, a pusher (e.g. uranium, lead or tungsten). There may also be fissile material present in the centre of the fusion package forming a so-called "spark plug" (usually 235U), where fission can be induced by neutrons of all energies. The spark plug has been commonly referred to in modem public literature (e.g. Rhodes, 1995, and CotE, 1995), although this is the only concept used that has not been officially disclosed by any nuclear power. However, the use of 235Uin secondaries was declassified in the United States in 1993. The conceptual outline of a thermonuclear device is given in Fig. 8.8. This is a notional device and should in no way be interpreted as a blueprint of a nuclear weapon. The information provided here is based on the published literature only, for example Robinson (1983), De Geer (1991), Rhodes (1995), Cote (1995) and USDOE (1996).
370
Chapter 8
Fig. 8.8. A conceptual sketch of a thermonuclear bomb with a boosted fission primary to the left and a secondary fusion to the right within the chamber.
The bomb explosion is initiated in many points around the outer surface of the primary. The high explosives detonate and cause the tamper to move inwards. The heavy tamper (in a primary the tamper can in fact be minimised to maximise the radiation output), together with the beryllium reflector inside, gains momentum during the implosion through the gap and hits the plutonium pit which becomes compressed and supercritical. A neutron source supplies initiating neutrons at the right moment and the chain reactions begin in the plutonium pit. The beryllium reflector helps stop neutrons from escaping and also multiplies neutrons by (n,2n)-reactions. The nuclear implosion that follows compresses the tritium/deuterium mixture, which has just before been injected into the centre, to such a degree that fusion reactions occur, and high energy neutrons are produced which initiate new fission chains in the plutonium before it fully blows apart. This boosting process typically multiplies the fission energy by a factor of two, but higher gains are also possible. The thermonuclear energy provided by the boosting process is small compared to the boosted fission energy, and the plutonium fission induced by high-energy neutrons is not a significant part of the total fission process. The DT-reactions act as an extra fission initiator where the majority of the energy is developed in the subsequent fission chain reaction phase. An intense stream of X-rays leaves the primary once it is heated to a temperature of many millions of degrees. This stream travels down the chamber where geometry and materials are designed to guide photons onto the heavy pusher around the fusion fuel. High levels of energy are absorbed in the outer layers of the pusher, resulting in material being boiled off and a strong inward momentum being generated (ablation). An extremely large force builds up, squeezing the fusion fuel to super density. The central string of fissile material (spark plug) is highly compressed and becomes supercritical and fission begins. This process increases the density and temperature further, thus improving the conditions for thermonuclear burning. The fusion neutrons contribute to tritium production through reactions with lithium in the thermonuclear fuel. It is essential that the fusion fuel is highly compressed, to avoid it being trans-
The Bomb
371
parent to the bremsstrahlung produced by the electrons in the plasma, as this would reduce the temperature below that which is required for fusion reactions to continue. The last stage in the process is optional. If the heavy pusher is made of depleted uranium the large amounts of high energy neutrons produced by fusion can be utilised to split 238Uatoms. Most of the very large bombs tested by the US and the Soviet Union in the 1950s used large pushers of depleted uranium. In those tests 238U fission completely dominated the fission processes in the bomb and often supplied as much energy as the fusion stage itself. In weapons where a high yield to size ratio is important, a fissile pusher is beneficial because the lower energy neutrons from the fusion process can contribute to fission together with neutrons degraded in energy. Enriched uranium has generally been cheaper to produce than plutonium, making 235U the most economical candidate for this kind of pusher.
8.4 ILLICIT T R A F F I C K I N G AND NUCLEAR TERRORISM For many years, those concerned with the spread of nuclear weapons worried more about their acquisition by nation-states than by terrorists. This was probably for two main reasons. First, it was believed that terrorists could not acquire the nuclear explosive materials---highly enriched uranium and separated plutonium~needed to make nuclear weapons. The problems of producing these weapons-usable materials were thought to be technically beyond the reach of small groups, and States having the ability to produce them were believed to have adequate physical protection against their acquisition by thieves or smugglers. Second, many experts believed that terrorist groups did not want to kill thousands of people~only enough to force the public to pay attention to the messages the terrorists wished to convey. As a result, the 1968 Treaty on the Non-Proliferation of Nuclear Weapons (NPT), and the IAEA safeguards it requires of non-nuclear-weapon States who sign it, were designed primarily to deal with the fear that States, not terrorists, might turn peaceful nuclear activities into bomb-building efforts. However, as pointed out by Barnaby (1990), there is a real risk that sub-national groups will in the future acquire fissile material~particularly plutonium~and construct a nuclear explosive. Equally disturbing, and perhaps more likely, is the possibility that plutonium may be acquired by a group who will threaten to disperse it, by an explosion, and radioactively contaminate a large urban area. These risks exist because: a large amount of civilian plutonium is being produced and stockpiled; a relatively small amount of such plutonium is needed for a nuclear explosive; the technical information required to fabricate a nuclear device is available in the open literature; and only a small number of competent people are necessary to fabricate a primitive nuclear device. The design of a "first generation" nuclear weapon, such as the bomb that destroyed Nagasaki in 1945, is no longer secret. Lovins (1980), for example, in an article in the scientific journal Nature, summarised the bulk of the necessary physics data showing
372
Chapter 8
Fig. 8.9. Types of simple A-bomb.
that a competent nuclear physicist can find the relevant information in the open literature. The basic nuclear weapon is the fission bomb, or A-bomb (A for atomic) as it was first called. A fission chain reaction is used to produce a very large amount of energy in a very short time--roughly a millionth of a second~and therefore a very powerful explosion. Several types of this device are possible (see Fig. 8.9 for illustration): 1. Gun type: A fuse sets off an explosive, which drives a uranium wedge through a gun barrel into the uranium target. 2. Implosive type: (a) A conventional explosive surrounds a sphere of uranium or plutonium that will be compressed into a high-density, supercritical mass. (b) As the core of fissile material implodes, an initiator releases neutrons which accelerate the fission reaction. (c) Implosion creates a supercritical mass and a chain reaction resulting in a nuclear blast. This seems to be a favourite "first try". For example Iraq's importation of explosives and electronics suggested development of an implosion-type bomb.
The Bomb
373
For this last type, high-explosive charges are required. The Nagasaki bomb used high-explosive charges of Composition B, a mixture of cyclotrimethylene-trinitramine (RDX) and trinitrotoluene (TNT), a fast-burning explosive more effective than TNT on its own. More modern implosion charges are diaminotrinitrobenzene (DATB) or triaminotrinitrobenzene (TATB). The amount of high explosive used in a fission weapon has decreased considerably since 1945--from about 500 kg to about 15 kg or less. Normally, the more explosive charges there are, the more perfect is the spherical symmetry of the shock wave--40 or so detonations would be typical. Getting the timing of the detonation sequence and the chemistry and geometrical shapes of the explosive lenses right are difficult problems in designing an implosion-type atomic bomb. But competent electronic and explosive engineers, given adequate resources and access to the literature, could solve them without too much difficulty. Explosive lenses and detonators adequate for an implosion-type atomic bomb are commercially available (see Carson et al., 1987). The crudest design could be very easily constructed by a team of technicians (or a competent technician working alone) from, say, a sub-critical mass of plutonium. The plutonium need not be in metal form; plutonium oxide, for example, is more convenient and safer to handle. The plutonium oxide could be contained in a spherical vessel placed in the centre of a large mass of conventional high explosive. When detonated remotely by an electronic signal the conventional explosive could compress the plutonium enough to produce some nuclear fission. Such a device could be positioned so that, even if it did not produce any nuclear fission, the conventional explosion would widely disperse the plutonium. Although a sub-national group could choose to use either plutonium or highly enriched uranium as the fissionable material for nuclear explosives, plutonium is increasingly the more likely option. A sub-national group that in the future decides to manufacture a nuclear explosive is, therefore, most likely to try to steal or to buy plutonium. According to Carson et al. (1987) the following would apply to such a crude nuclear device: 1. Such a device could be constructed by a group not previously engaged in designing or building nuclear weapons, providing a number of requirements are adequately met. 2. Successful execution would require the efforts of a team having knowledge and skills additional to those usually associated with a group engaged in hijacking a transport or conducting a raid on a plant. 3. To achieve rapid turnaround (that is, to make the device ready within a day or so of obtaining the material), careful preparations extending over a considerable period would have to be carried out, and the materials acquired would have to be in the form prepared for. 4. The amounts of fissile material necessary would tend to be large--certainly several times the minimum quantity required by expert and experienced nuclear-weapon designers.
374
Chapter 8
5. The weight of the complete device would also be large--not as large as the first atomic weapons (about 4.5 tons), since these required aerodynamic cases to enable them to be handled as bombs, but probably more than a ton. 6. The conceivable option of using oxide powder (whether of uranium or plutonium) directly, with no postacquisition processing or fabrication, would seem to be the simplest and most rapid way to make a bomb. However, the amount of material required would be considerably greater than if metal were used. 7. There are a number of obvious potential hazards in any such operation, among them those arising in the handling of a high explosive; the possibility of inadvertently inducing a critical configuration of the fissile material at some stage in the procedure; and the chemical toxicity or radiological hazards inherent in the materials used. Failure to foresee all the needs on these points could bring the operation to a close; however, all the problems posed can be dealt with successfully provided appropriate provisions have been made. As commercial reprocessing increases, plutonium will be increasingly transported worldwide on virtually all the main transportation systems--road, rail, sea, and air. Plutonium is most vulnerable to theft while it is being transported. When considering the problem of plutonium, one has to keep in mind its three different forms: 1. Weapons-grade plutonium. This plutonium is 90% or more plutonium-239, the most suitable isotope for nuclear explosive. 2. Separated reactor-grade plutonium. This plutonium has been produced as a by-product of commercial nuclear power plants, and normally contains much less than 90% plutonium-239. Reactor-grade plutonium can also be used to make nuclear explosives, but is much less suitable for this purpose than is weapons-grade plutonium. 3. Reactor-grade plutonium in spent fuel. Most of the plutonium that has been produced in the world so far remains in the spent fuel removed from civil nuclear power plants. Plutonium in this form can be used for nuclear explosives only after separation from the spent fuel and purification, through an operation known as reprocessing. Each of these forms of plutonium presents proliferation risks, but of differing degrees and nature. The issue of proliferation cannot be assessed without distinguishing between national threats and sub-national threats. To date, national proliferation has depended on the use of nuclear material--weapons-grade plutonium or highly enriched uraniummproduced specifically for weapons purposes. This will continue to be the preferred route to national proliferation should it be attempted, but diversion of materials from the safeguarded civil nuclear fuel cycle is also a possible, although improbable, means of acquiring material for nuclear explosive purposes. Sub-national threats involve the theft or seizure of nuclear material by individuals or groups acting without governmental authorization. While such threats clearly exist, thus far no sub-national group or individual is known to have succeeded in acquiring nuclear
The Bomb
375
material sufficient for fabrication of a weapon. The distinction between national and sub-national threats is important because the capabilities of national authorities and sub-national groups both to obtain and to make use of nuclear materials differ greatly. According to the report by Bunn (1997), smuggling of sensitive nuclear materials has in fact already occurred. L. Koch of the European Commission's Joint Research Centre that analyses material from nuclear smuggling cases says that some involved "weapon" material or "weapon-usable" material. Indeed, there have been multiple seizures by authorities in Russia and elsewhere of kilogram quantities of weaponsusable material, mostly highly enriched uranium. Given the huge quantities of weapons-usable material produced by both Russia's predecessor and the United States, given the changes taking place in Russia, and given the current dismantlement of 1500-2000 nuclear weapons per year by both countries, theft and smuggling of weapons-usable material should not be surprising. Moreover, many familiar with law enforcement believe that crimes of many kinds go undetected and therefore unknown. Successful smuggling of weapons-usable material could have occurred without detection. One can no longer assume that terrorists, whether domestic or international, cannot acquire weapons-usable material. In addition, the assumption that terrorists do not want to kill thousands of people and therefore would not use weapons of mass destruction has turned out to be wrong. The bombing of the World Trade Center in New York City by international terrorists, had it gone according to plan, might have killed many of the 10,000 people in the twin towers that were supposed to fall. The bombing in Oklahoma City by a domestic terrorist killed 169 and injured 600. The release of the chemical-weapon nerve gas in the Tokyo subway by the Japanese Aum Shinrikyo sect was meant to kill more than twelve; it did injure 5000. Why wouldn't these terrorists have used nuclear explosives, even crude devices, if such explosives had been within their reach? (a question asked by Bunn, 1997). This has also been recognised by the international community and in November 1997 more than 200 experts from 48 countries and organisations attended the IAEA's International Conference on the Physical Protection of Nuclear Materials. The meeting focused on national and global experience in regulation, implementation, and operation of physical protection systems and standards. Reviews of national experience included papers and presentations covering a wide range of topics. They included the implementation of protection programmes at specific types of nuclear facilities; organisational, regulatory, and legal aspects of national infrastructures; methods and approaches for assessing and improving procedures and systems; bilateral co-operative programmes for physical protection; physical protection during the transport of nuclear materials; research, development, and use of instrumentation and computerised security systems; and programmes that have been put into place for combating and preventing illicit trafficking in nuclear materials. The conference concluded that only State authorities can be responsible for detecting and responding to illicit trafficking activities on their territory. However, no clear minimum requirements exist on what measures are necessary to meet this responsibility.
376
Chapter 8
In s o m e countries, the anti-trafficking i n f r a s t r u c t u r e ~ w h i c h e n c o m p a s s e s responsible authorities including customs, police, nuclear, intelligence and defence a g e n c i e s ~ a r e co-operating and co-ordinating their efforts against trafficking. Threat and r e s p o n s e scenarios are identified. Personnel are also trained in nuclear-related matters (e.g. at schools for c u s t o m s and police staff). Detection e q u i p m e n t for nuclear material is available. R e g u l a t i o n s and procedures are established and the public is informed. T h e s e are g o o d m o d e l s f r o m which other States m a y benefit (Thorstensen, 1996; see also Willrich and Taylor, 1974).
REFERENCES Barnaby, F., Weapons of mass destruction: A growing threat in the 1990s? Research Institute for the Study of Conflict and Terrorism, London, 1990. Bunn, G., Physical protection of nuclear materials--strengthening global norms. IAEA Bulletin, 39 (1997) 4. Carson, M.J., Taylor, T., Eyster, E., Maraman, W. and Weshsler, J., Can terrorists build nuclear weapons? In: P. Leventhal and Y. Alexander (eds.), Preventing Nuclear Terrorism. Lexington Books, Massachusetts, 1987. Committee for the compilation of materials on damage caused by the atomic bombs in Hiroshima and Nagasaki: Hiroshima and Nagasaki, The Physical, Medical and Social Effects of the Atomic Bombings, Iwanami Shoten, Pub., Tokyo. CotE, O.R., A primer on fissile materials and nuclear weapon design, Appendix 1. In: G.T. Allison, O.R. CotE, R. Falkenrath and S. Miller, Avoiding Nuclear Anarchy. MIT Press, 1995. De Geer, L.-E., The radioactive signature of the hydrogen bomb. Sci. Global Security, 2 (1991) 351. Dubasov, Y.V., Zelenov, S.A., Krasilov, G.A., Logatjev, V.A., Matusjtjenko, A.M., Smagulov, S.G., Tsaturov, Y.S., Tsyrkov, G.A. and Tjerysjev, A.K., The chronology of atmospheric nuclear tests at the Semipalatinsk test site and their radiation characteristics, Report to the SCOPE-Radtest Meeting in Barnaul, Siberia, September 1995. Lovins, A.B., Nuclear weapons and power-reactor plutonium, Nature, 28 February 1980, pp. 817-823 and typographical corrections, 13 March 1980, p. 190. Rhodes, R., Dark Sun--The Making of the Hydrogen Bomb. Simon & Schuster, New York, 1995. Robinson, C.P., The weapons program, Overview. Los Alamos Sci., 7 (1983) 10. Serber, R., The Los Alamos Primer, Report LA-1, April, 1943. Thorstensen, S., Safeguards and illicit nuclear trafficking: towards more effective control. IAEA Bulletin, 4 (1996) 29. US Department of Energy, Drawing back the curtain of secrecy, information published on the USDOE interest site www.doe.gov under OPENNET, 1996. Willrich, M. and Taylor, T., Nuclear Theft: Risks and Safeguards. Ballinger, Cambridge, Massachusetts, 1974.
377
CHAPTER 9
Monitoring Accidentally Released Radionuclides in the Environment
9.1 I N T R O D U C T I O N In this chapter we will address the problem of accidental release of radionuclides into the environment. Accidental release could be the consequence of the following activities: 9 mining, milling, enrichment and conversion of uranium 9 nuclear fuel fabrication 9 reactor operation or meltdown 9 nuclear fuel reprocessing 9 nuclear materials transport 9 satellite nuclear power source 9 nuclear propulsion 9 weapons production and formerly utilised sites 9 nuclear bomb accidents 9 plutonium fabrication 9 tritium production 9 fusion reactors 9 industrial uses: radioisotope production, irradiation facilities, industrial processing and mining which enrich natural radionuclides, breaks in retaining walls at tailing sites 9 medical radioisotopes. The group of radionuclides to be considered are the ones produced by nuclear explosions and the ones present in the irradiated reactor fuel. This group comprises several hundred radionuclides, but only a limited number of them contributes significantly to human exposure. These would normally include fission products and activation products. Radioactive noble gases, e.g. 85Kr, ~33Xe, are not considered since they are unlikely to contribute significantly to internal exposure via the food chain.
Chapter 9
378
Table 9.1 Fission and activation products which may be of concern in human exposure
Fission products
Activation products
Nuclide
Half-life
Fission yield %
Major decay
89Sr 90Sr' 90y 95Zr, 95Nb 99Mo, 99mTc I~ l~ l~ l~ 129roTe 131I 132Te, 1321 137Cs, 137mBa 14~ 14~ 144Ce, 144pr
50.5 d 28.7 a, 64.1 h 64.09 d, 35.0 d 2.747 d, 6.006 h 39.272 d, 56, 116 min 372.6 d, 29.92 s 33.6 d 8.021 d 76.856 h, 2.3 h 30.0 a, 2.55 min 12.751 d, 1.6779 d 284.45 d, 17.28 d
4.77 5.76 6.51 6.09 3.03 0.4 0.661 2.875 4.282 6.136 6.134 5.443
~
3H 14C 55Fe 59Fe 54Mn 6~ 65Zn 134Cs 239Np 241pu, 241Am 242Cm 238pu 239pu 24~ 242pu
12-35 a 5730 a 2.75 a 44.53 d 312.5 d 5.27 a 243.9 d 754.2 d 2.355 d 14.35 a, 432,0 a 162.94 d 87.7 a 2.411 x 104 a 6.563 x 103 a 3.735 x 105 a
w
~-,~~-, ~, ~-, ~ ~-, ~, ~-, v f~-, ~, ~-,~, ~-,~ ~-, ~-, ~ I~-, ~' ~-,~, [~-,~ ~-,% ~-,~
[3EC
13-, ~' EC, ~/
13-,~, EC, y ~-,
~-,~, 13-,~ '
(x
(x o~ o~
Note: Half-life is given in minute (min), hours (h), days (d) and years (a). One year = 365.25 days.
Radionuclides produced in fission and activation processes which may contribute significantly to human exposure in the event of an accident are listed in Table 9.1. The primary source of radionuclides produced in the fission process and found in the environment is atmospheric testing of nuclear weapons. The public has been exposed to these and other radionuclides for five decades, but there has been a substantial decline in atmospheric testing in the past two decades. Therefore the major source of fission product radionuclides in recent years has been from nuclear accidents. A nuclear reactor meltdown could release a spectrum of radionuclides similar to that of a nuclear bomb explosion, but the ratios of nuclides would greatly differ for the two cases. The reason for the differences in ratios of radionuclides is that during the reactor operation the long-lived radionuclides tend to build up progressively, whereas the
379
Monitoring Accidentally Released Radionuclides in the Environment
short-lived radionuclides tend to reach an equilibrium state at which the rate of decay equals the rate of production. The proportion of various radionuclides produced in the operation of a nuclear reactor changes with operating time and with fuel burn-up. Radionuclides classified as activation products are created in nuclear reactors and other nuclear devices by the reactions of neutrons with fuel and construction materials. Activation products include the isotopes of the transuranic elements and radioisotopes of hydrogen, carbon, caesium, cobalt, iron, manganese, zinc, and a host of other radionuclides, all of which should be recognised and considered in determining the environmental pathways to human exposure. As an illustration, Table 9.2 shows core inventory and estimate of total release of radionuclides following the accident in April 1986 at the Chernobyl nuclear power station.The list of radionuclides and specific matrices which should secure an increased interest depends on the source of release and the localities of the site at which it occurs. In general the scenario is shown in Table 9.3. Table 9.4 lists specific radionuclides and matrices which should be receive special attention in most cases. Table 9.2 Core inventory and estimate of total release of radionuclides Radionuclide
Half-life
Inventorya ( E B q )
Percentage releasedb
85Kr 133Xe 13111 132Te
10.72 a 5.25 d 8.04 d 3.26 d 30.0 a 2.06 a 50.5 d 29.12 d 64.0 d 2.75 39.3 368 d 12.7 d 32.5 d 284 2.36 d 8774 a 24065 a 6537 d 14.4 a 163 d
0.033 1.7 1.3 0.32 0.29 0.19 2.0 0.2 4.4 4.8 4.1 2.1 2.9 4.4 3.2 0.14 0.001 0.0008 0.001 0.17 0.026
--100 ~ 100 20 15 13 10 4 4 3 2 3 3 6 2 3 3 3 3 3 3 3
137Cs 134Cs 89Sr 9~ 95Zr
99Mo l~ l~ 14~
141Ce lance 239Np 238pu 239pu 24~ 241pu Z42Cm
aDecay corrected to 6 May 1986. bStated accuracy: -+50% except for noble gases.
Chapter 9
380
Table 9.3 Activities required after an accident Initial Stage: Day 0-2
Early warning systems data. Field gamma survey and sampling should start as soon as possible. Existing filtration systems in buildings and engines and eventual precipitation collected by chance will constitute important samples for a retrospective qualitative assessment of the source term.
Intermediate Stage: Day 2-14
Air sampling in order to see if fallout continues. Sampling every h - 6 h. Daily sampling of precipitation (dry and wet) over a well defined area (0.5-5 m2). Sampling of grass/mosses/lichens soil over well defined areas 0.25/m 2. Sampling of water. Analysis of actinides and beta-emitters. Control of contamination in vegetables, milk etc. In situ gamma spectrometry for assessment of the extent of contamination. Isoactivity curves drawn.
Later Stage: Day 14 +
Air sampling daily; weekly to assess eventual repercussions. Sampling and analysis of different important foodstuffs for gamma, alpha and beta emitters. Modelling for dose assessment.
Table 9.4 Nuclides of interest in specific matrices Nuclide
Matrix A
W
V
3H
X
X
X
14C
X
41Ar
X
S
F
6~ 85Kr
X
89Sr
X
X
X
9~ 95Zr/95Nb
X X
X
X
99Tc 1311 132Te 133Xe
X X X
X X
X
134Cs 137Cs
X X
X X
X X
X
X
X
X
X X
X
X X
X X
X
X
X
X X X X
X X X X
X
210po 226Ra X
D
X
210pb
Actinides
M
Matrix: A = air, W = water, V = vegetation, S = soil, F = food (general), M = food (marine), D = sediment.
381
Monitoring Accidentally Released Radionuclides in the Environment
9.2 P A T H W A Y S AND S A M P L E S OF I N T E R E S T In this discussion, we follow IAEA-Report 295 (1989) and describe samples and pathways relevant to the analysis of radionuclides in foods, and of environmental materials that are part of the immediate pathways leading to contamination of food. The sources of a release and the conditions at the site where it occurs, determine one or more critical pathways in the environment between the point of discharge and man. The season of the year determines to a great extent the magnitude of contamination of different foods or environmental components. It is very important to consider the pathways of radionuclides in the environment for design of the environmental monitoring program. Radionuclides enter the receiving environment via direct emissions to atmosphere, direct discharges to water bodies or releases from land burials of radioactive wastes. Simplified pathways between releases to atmosphere and man are shown in Fig. 9.1. Exposure may occur by direct irradiation from radionuclides in air or deposited on surfaces, by inhalation of airborne radionuclides or by consumption of contaminated food such as vegetables, milk and water. Direct irradiation from a plume and inhalation of radionuclides in a plume are direct pathways of exposure. The others may involve many transfer processes between sections of the environment. An example of such a pathway is the deposition of ~3~Ion grazing land, its direct retention on grass or its uptake into the grass from the soil, its ingestion by cows and the subsequent ingestion of cow's milk by people, especially children. Figure 9.2 shows similarly simplified pathways between releases of liquid wastes to sea or river and man. Direct irradiation may occur from the accumulation of radionuclides in surface soils and sediments. Water and food may become contaminated, because of various transfer processes through the environment and again the compartments shown are simplified representations of the sectors that may be involved in real processes. ,,
Directradiation b
,,,
Deposition
~ Ingestion 1' Air.... ~ [----~plantsI ] ' [DepositionS[ :[ JRadioactive[ [ ~ 1 radiir~ict~ Manl [materials[ l~inhalation, i ~ Ingestion b.
)11
iii
All
i
Inhalation Fig.9.1.Simplifiedpathwaysbetweenradioactivematerialsreleasedto atmosphereandman.
Chapter 9
382 Radioactive materials
.,
Soil
Direct radiation
Aquatic plants Fishing
] ~ Direct radiation Surface or ground water ,
Aquatic animals
P
t
Radioactive materials
Plants
Animals
Ingestion
Fig. 9.2. Simplified pathways between radioactive materials released to ground or surface waters (including oceans) and man.
The main purpose of analysis should be fast identification of the most critical samples and the most important radionuclides so that the necessary rapid actions can be carded out. Let us first discuss food items. Only those foods should be sampled and those radionuclides analysed whose consumption contributes significantly to population exposure. If, for example, 131I is being released in proximity to cow pastures, its concentration in the milk produced will provide far more meaningful information than its concentration in air, or deposition on forage samples. Nevertheless, measurements of 131Iin pasture grass may be very important in providing an indication of the expected concentration in milk. For other circumstances, the need for food sampling should be based on a thorough understanding of agricultural practice and of food consumption in specific areas of interest. It is recommended that food analyses be based on the determination of radionuclides in individual food items rather than a mixed diet sample. Only the analysis of individual foodstuffs can indicate whether and which countermeasures should be taken to reduce doses. Food sampling for estimation of total consumption should be carried out at the retail level when appropriate; otherwise, it should be carried out at the consumption level. The selection of foods to be sampled can be based on individual diet or food consumption statistics. Analyses of individual foodstuffs should preferably be performed after preparation, taking into account the effect of kitchen activities such as washing, cleaning and cooking.
Monitoring Accidentally Released Radionuclides in the Environment
383
Milk and milk products are important components of the diet in many countries. Milk is one of the few foods produced over large areas and collected on a daily basis. Its composition is almost identical all over the world, and it is easy to collect a representative sample that can be analysed in liquid or dried form. Milk is likely to be contaminated by radioactive iodine and caesium with the first days after a release of volatile radionuclides. Contamination of milk will be greatest when cows are grazing during the fallout period, but even when cows are kept indoors the contamination of milk may occur by inhalation of radionuclides or their ingestion via drinking-water or contaminated feed. Milk from goats and sheep should be checked periodically over a longer period, because of their grazing habits. After harvesting, grain and rice are subjected to contamination only during storage, and only the outer layers would be contaminated if fallout occurs during the growing process. It is relatively easy to select representative samples of grain and rice at harvest time. If the fallout occurs during the winter, the grain will be contaminated only through root uptake in the next growing season. Following an accidental release of radiocaesium, meat becomes one of the main sources of dietary contamination. This mainly results during animal grazing, but contaminated drinking-water might also be an important pathway. Inhalation of radiocaesium is not likely to be a significant pathway to meat. Meat sampling should normally be done in such a way that the composite sample is representative of a large number of animals, although screening measurements of individual animals may be necessary after heavy fallout. Following an accident, contamination of fish in nutrient-deficient lakes may constitute a particularly significant pathway to the uptake of radiocaesium by man. Obtaining a representative sample from an area containing many lakes may require some compromise, since the collection of samples from a large number of the lakes may be impracticable. Ocean-fish will not take up as much radiocaesium as freshwater fish because of the dilution through the depth of the ocean and the effective dilution associated with the high potassium content in the water. Particulate-associated radionuclides can, however, be enriched to high levels. Mussels such as Mytilus edulis, some species of macro algae, and other filter-feeders quickly take up the contaminants from sea-water and can also be used as biological indicators. Green leafy vegetables are very prone to external contamination during their growing season. Other vegetables, including root vegetables, may also become contaminated. It is important to obtain representative samples, and the sampling should be planned carefully. In the early stages of fallout, green vegetables can provide a very significant pathway for short-lived radionuclides. Game, and food such as mushrooms and berries, can be contaminated markedly, although only in very rare cases would they contribute significantly to the ingestion dose. It may still be advisable to analyse these foods in order to decide whether the levels comply with international export regulations. Environmental samples to be analysed for the activity of different radionuclides include air, water, soil, grass and sediments.
384
Chapter 9
Measurement of airborne radioactivity provides the first opportunity of identifying the spectrum of radionuclides making up the contamination. Radionuclides will appear very rapidly in ground-level air, and air samples can give the first indication of the nature of the contamination. Radioactive materials in the air may result in exposure to man by inhalation, by ingestion of particulate matter deposited on vegetation, or by ingestion of products derived from animals which were exposed to radioactive materials through inhalation or ingestion. Rainwater and snow are also early indicators of radioactive contamination. In some places drinking-water and rain-water can be significant pathways of short-lived radionuclides such as radioiodine to man or animals. Drinking-water and household-water are potentially important pathways, directly or through their use in food preparation and processing, although dilution, time-delays and water treatment can reduce the contamination levels markedly. Water consumed by livestock and/or used for irrigation purposes can also be a source of radionuclides in foods. Sea-water can be a source of contamination for seafoods such as mussels, shellfish, fish and algae. Water from streams, lakes and ponds should also be considered as a source of contamination. Contaminated soil serves as a direct source of radionuclides leading to the contamination of all agricultural products. Contaminated soil used in greenhouses could add significantly to the contamination of vegetables. Grass is a direct pathway of radionuclides to animals and then to man through meat and/or milk. The radionuclide content of grass can provide a basis for deciding whether cattle can be permitted to graze in a given area. Sediment in all types of water (sea, lake, pond and large or small streams) may be a source of contamination to aquatic organisms. Contaminated sedimentary materials used as fertilisers may also increase the radioactivity levels of soil. In large-scale facilities such as reactors, hot laboratories and accelerators, area monitors are installed in the controlled area to detect abnormal increments of radiation dose rate and to inform the workers of abnormalities by sound or light to prevent unnecessary exposure. Abnormal levels are previously defined as investigation level and/or intervention level. The detectors generally used for the y area monitors are GM counter and NaI(T1) scintillator for lower dose rate regions and ion-chamber for higher dose rate regions. For the neutron, a BF3 counter with or without moderator is used. For quick and effective measures in the case of radiologically abnormal situations and for comprehensive data filling and analysis, these measured values together with the other items of radioactive concentration in air and radioactivities in airborne effluents are centralised and analysed continuously by the computerised data acquisition system. The more important factors affecting the design of the survey are as follows: a. The type of installation and the potential hazard associated with it. b. The nuclides to be released, their activity, their physical and chemical form, and the method and route of release. c. The existing or expected presence of these nuclides from sources.
Monitoring Accidentally Released Radionuclides in the Environment
385
d. The behaviour of the released nuclides in the environment. e. Natural features of the environment which affect the behaviour of released nuclides, e.g., climate, topography, pedology, geology, hydrology and vegetative cover. f. Man-made features of the environment which affect the behaviour of released nuclides, e.g., reservoirs, regulated streams or rivers, and harbour. g. The utilisation of the environment for agriculture, fisheries, water and food supplies, industry and recreation. h. The population distribution and habits. An example of the operational environmental monitoring program presented in the "Guide for Environmental Radiation Monitoring" by the Japan Atomic Energy Commission is shown in Table 9.5. Following a release of radionuclides from a uranium-fuelled reactor to the environment the most important radionuclides to be assessed for internal exposure from the ingestion of food and water, and for the contamination of environmental materials which are parts of the immediate pathways leading to contamination of food, are 134Cs, 137Cs(137mBa), 131I, and other gamma emitters, the beta emitters 89Sr, 9~ and tritium, and the alpha emitters 238pu,239+24~ 241Amand 2421--, ~m. The levels of radionuclides in the environment and food have been extensively compiled by UNSCEAR. In general, the radionuclides of major importance in the contamination of food and environmental samples (materials which are part of the pathways leading to the food) are:
1311, 134Cs, 137Cs 131I, 134Cs, 137Cs
Air Water
3H, 895r, 9~
Milk
895r, 9~
131I, 134Cs, 137Cs
Meat 134Cs, 137Cs Other foods 895r, 9~ 137Cs Vegetation 895r, 9~ 95Zr, 95Nb, l~ l~ 1311, 134Cs, 137Cs, 141Ce, ~44Ce Soil 9~ 134Cs, 137Cs, 238pu, 239+24~ 241Am, 242Cm
This group of radionuclides is most likely to be of concern in terrestrially produced foods. Biological concentration processes in freshwater and marine systems can result in very rapid transfer and enrichment of specific radionuclides. The radionuclides which enter such systems can in certain cases be rapidly accumulated by plankton and algae. These organisms serve as food for higher trophic levels and thus the radionuclides become concentrated in organisms such as oysters, clams and shrimps. Radionuclides of particular concern in freshwater and marine food chains include: 54Mn' 55Fe' 59Fe ' 6~ ' 65Zn ' 95Zr ' 95Nb ' l~ ' l~ ' lJ~ ' 1255b' 1311, 134Cs' 137Cs' ]44Ce and some of the transuranic elements. Many other radionuclides would be present in debris from a nuclear accident, and their potential contribution to human exposure depends on the type of accident and the circumstances when it occurred. Since there are several types of fuel, the spectra of radionuclides that would be present in accidental releases could be somewhat different.
Chapter9
386 Table 9.5 Environmental monitoring program Items
Objects
Frequency
Methods of measurement
Exposure
Exposure rate Integral exposure
Continuously Quarterly
NaI(T1),I.C., GMT 4-8 TLD elements TLD are used for one point measurement
On occasion Quarterly
Nuclide analysis Nuclide analysis
On occasion Every 6 months At the time of harvest
131I analysis Nuclide analysis Nuclide analysis
Quarterly
Nuclide analysis
Radioactive in Dust in air land sample Land water (drinking water) Cow milk Soil Agricultural products; green vegetables, edible roots, rice Index plants
Fallout, rain water, dust Monthly Radioactive in Sea water marine sample Sea soil Marine foods Index plant Meteorology
Wind speed Wind direction Temperature Solar radiation Net radiation Precipitation etc.
Every 6 months Every 6 months At the time of fishing season Quarterly
Note
Surface soil
Nuclide analysis
Mugwort, pine needle etc. Basin method
Nuclide analysis
Surface water Surface soil
Continuously
Gulfweed etc. To estimate the dispersion factor, wind speed, wind direction and atmospheric stabilities are statistically analysed according to the JAEC's Guide.
Note 1: One to two years after the operation, more detailed and frequent monitorings are preferable. Note 2: Nuclide analyses are done using instrument as a rule.
T h e f o l l o w i n g f o u r n u c l e a r a c c i d e n t s c e n a r i o s are c o n s i d e r e d in detail in I A E A - R e p o r t 195 (1989): 1. r e a c t o r m e l t d o w n , w i t h or w i t h o u t failed c o n t a i n m e n t ; 2. r e a c t o r m e l t d o w n w i t h particle c o n t a i n m e n t ; 3. n u c l e a r f u e l - r e p r o c e s s i n g plant release; 4. p l u t o n i u m f u e l - f a b r i c a t i o n plant release. T h e r a d i o n u c l i d e s r e l e a s e d in e a c h of the a b o v e scenarios are:
Monitoring Accidentally Released Radionuclides in the Environment
387
(i) Reactor meltdown with or without failed containment A. Of importance in the first day: a. Radionuclides with noble gas precursors; b. Volatile radionuclides; c. Less volatile and refractory radionuclides (fine particles, aerosols). Radionuclides with half-lives of 6 hours and greater. 90y, 91Sr ' 93y, 96Nb ' 97Zr ' 99M0 '
l~
' l~
' lllAg ' ll2pd ' 115Cd' 121Sn'
1255n ' 1265b ' 1275b ' ~31i, 132i,
132I, 131roTe' J32Te' 133i, 135i, 140La' 143Ce' 143pr' 146Ba' 147Nd' 149pm' ~5~pm' 153Sm' 156Sm' 157Eu' 239Np" The presence of high levels of the radionuclides of cerium, zirconium, ruthenium and transuranic elements in foods and environmental materials indicates the presence of hot particles which may be of special importance in considering exposure by inhalation and/or ingestion. B. Of importance in the first week: a. Volatile radionuclides; b. Less volatile or refractory radionuclides. Radionuclides with half-lives of about 1 day and greater: 86Rh ' 895r ' 9Oy, 91y, 95Nb ' 95Zr' 99Mo, l~ ~~ ~Ag, 112pd' 115Cd' llsmCd ' 121Sn ' 1245b ' t25Sn ' 127Sb' 131i, 131roTe' ~32Te' 133i, 136Cs' 14~ ' 14~ ' 141Ce' 143Ce' 143Ce' 143pr' laVNd' 149pm' 15~pm' 153Sm' 239Np
C. Of long-term importance: 3H, 895r ' 90Sr ' 91y, 93mNb ' 95Nb '
l~
' ~~ ' ll~ 121mSn ' 1245n ' 124Sb ' 141Ce ' 144Ce ' 147pm ' 24~ ' 241Am ' 241pu '
' ll3mCd ' ~lSmCd ' 129i, 134Cs ' 137Cs ' 16~ ' 238pu ' 239pu ' 242Cm ' 242pu ' 243Am '
244Cm
(ii) Reactor meltdown with particle containment A. Of importance in the first day: 3H, 88Rb' 895r ' 90Sr ' 90y, 91Sr ' 91y, 103Ru ' ~05Ru ' 106Ru ' ~2~i, 123i, 123I, 134I, 136Cs ' 138Cs ' 139Cs ' 139Ba ' 14~ ' 140La
Chapter9
388
B. O f i m p o r t a n c e in the first week: 3H, 898r ' 9~ ' l~ ' l~ ' l~ 1311, 1331, 14~ ' 14~
'
C. Of long term importance: 3H, 898r ' 9~ ' 99Tc ' l~ 131I, 137Cs"
' ~~
(iii) Nuclear fuel reprocessing plant release 9~ ' 95Nb' 95Zr' 99Tc' l~
' l~
'
1291, 131I, 134Cs ' 137Cs ' 141Ce ' ~44Ce' 238pu ' 239pu ' 24~ Z41Am ' 241pu ' 242Cm 242pu ' 243Am ' 244Cm"
(iv) Plutonium fuel fabrication plant release: 238pu ' 239pu ' 24~
' 241Am ' 241pu ' 242pu
Other nuclear accidents which may result in major atmospheric emissions are: 9 Plutonium fuelled reactor meltdown 9 Breeder reactor meltdown 9 High flux radionuclide production reactor meltdown 9 Fast flux reactor meltdown 9 Nuclear powered ship/submarine reactor meltdown 9 Satellite re-entry and burn-up of satellite nuclear power source 9 Nuclear weapon destruction by chemical explosion 9 Criticality at nuclear materials processing plant 9 Fusion reactor fuel loss. Each of these possible accidents may release a unique spectrum of radionuclides and this should be considered in developing radioanalytical capabilities.
9.3 GUIDELINES FOR RADIOLOGICAL MONITORING OF THE ENVIRONMENT Based on the general guidelines discussed earlier, many countries have established a set of national guidelines for radiological monitoring of the environment. This type of standard usually provides guidelines for establishing an environmental program covering a. sampling and analysis protocols; b. analytical techniques and sensitivity; c. statistical treatment of monitoring results; d. quality assurance; e. methods for expressing results; and f. record keeping.
389
Monitoring Accidentally Released Radionuclides in the Environment
Although such a set of rules applies to the environment of a nuclear facility operating mainly under normal conditions, certain of these requirements may also apply to situations involving abnormal releases. An environmental monitoring program should be established if there is a need for it as discussed in the BSS (Basic Safety Standards). The major objectives of a monitoring program are to: a. permit the estimation of actual or potential doses to critical groups and populations from the presence of radiation fields or radioactive materials in the environment; b. provide data to confirm compliance of the facility of source with release guidelines and regulations and to provide public assurance of compliance; and c. provide a check, independent of effluent monitoring, on the effectiveness of containment and effluent control. Additional objectives include a. establishing and maintaining the capability for environmental monitoring so that an effective response can be made to emergency conditions; b. maintaining a database to facilitate the detection of trends; c. verifying or refining the predictions of environmental models; d. determining the fate of released radioactive materials to show whether any significant pathway to man has been overlooked. Special attention should be paid to the design of an environmental monitoring program. The design of an environmental monitoring program requires sound professional Immersion Inhalation External
Vegetated soil
Y
Atmosphere
Forage and crops
Ingestion
Animal produce
Ingestion
Aquatic animals
Ingestion
Source
Dose
~ Ir
"1
....
Surface water
Aquatic plants
~I
Sediment
Ingestion
External Ingestion Immersion
Fig.9.3.Environmental
transfer model.
_
390
Chapter 9
judgement in conjunction with thorough knowledge of the facility and the local environment. The environmental monitoring program should be designed to allow the assessment of the most significant radionuclides and pathways resulting in doses to humans. In selection of samples, measurements and monitoring locations, one should consider an environmental transfer model (as shown in Fig. 9.3, after CAS 1990).
9.3.1 Objectives of environmental monitoring Environmental monitoring means the measurement of radiation and radioactivity outside the boundaries of installations operating nuclear power plants, research reactors, fuel reprocessing plant, accelerators or handling radioactivity materials including nuclear fuels or radioactive sources. General objectives that would justify setting up an environmental monitoring program, described in CRP publication 43, are as follows: a. to assess actual or potential doses to critical groups and populations from the presence of radioactive materials or radiation fields in the environment from normal operations or accidents, b. to demonstrate compliance with authorised limits and legal requirements, c. to check the condition of the source, the adequacy of operation of the plant or containment and the effectiveness of effluent control, to provide a warning of unusual or unforeseen conditions and, where appropriate, trigger a special environmental monitoring program. There is a need for monitoring to fulfil one or more of the basic objectives, then different programs can be implemented to satisfy objectives. Subsidiary objectives are as follows: a. to provide information to the public, b. to maintain a continuing record of the effect of the installation or practice on environmental radioactivity levels, c. to distinguish the contribution from the operator's installation practice from the contribution from other sources, d. to obtain data on the behaviour of the local environment that may be required in assessment of the consequences of accidents, e. to identify changes in the relative importance of transfer pathways and mechanisms including the emergence of new pathways, and therefore to enable the environmental monitoring program to be revised in the light of experience and in response to changing conditions, f. to verify or refine the predictions of environmental models, so as to improve the structure of the model and to reduce uncertainties in the parameters, g. to conduct more general, scientific studies aimed at improving knowledge of the transfer of radionuclides in the environment. In addition, the testing of nuclear bombs in the atmosphere in the late 1950s and the early 1960s has led to widespread contamination of air, soil, water, and the biosphere.
Monitoring Accidentally Released Radionuclides in the Environment
391
Many countries have started surveillance programs to monitor contamination. It can be assumed that the high contamination that was found, for instance, in the Scandinavian countries prompted efforts to enact the 1962 treaty of the ban on nuclear tests in the atmosphere. Since then the concentration levels of radionuclides in the environment had declined considerably, partly due to decay and partly due to their removal to sinks where they are strongly bound and cannot be recycled into the biosphere. The basic idea of surveillance networks has changed from monitoring fallout, more or less as a means to follow the decline of the artificial contamination in the environment, to monitoring discharges from nuclear power plants and to preparing for the possibility of widespread contamination following a severe reactor accident or even nuclear warfare. Of special interest is radiation monitoring of the workplace, the objectives of which are as follow: 1. to confirm that no abnormal radiation level has occurred in the workplace, 2. to assess an upper limit of the dose equivalent of the workers due to the external and internal exposure based on the measurement of radiation and radioactivity level, working condition and other information in the workplace, 3. to obtain the information available for deciding the appropriate protection and the operational procedure by knowing the radiation level in the workplace before operation, 4. to obtain the data necessary for another design of monitoring program such as individual monitoring for the internal exposure. In general the radiation monitoring can be subdivided as follows:
External monitoring Personal __j-monitoring ~_ Internal monitoring Radiation monitoring
Environmental monitoring
Workplace monitoring External monitoring Surface contamination Air contamination Radioactivities in the waste (gaseous and liquid) Surface contamination of carry-out materiales Environmental monitoring ~ External radiation Radioactivities in environmental samples Meteorological observation
392
Chapter 9
In environmental radiation monitoring it is important to know the geographic location of the sample point on the earth's surface and in the past this could only be determined by the use of a theodolite, often a laborious and time-consuming procedure. Various other systems have been used such as celestial navigation and various radiobased systems but these suffered from the drawbacks of limited accuracy, restricted coverage or high cost. The dynamic nature of modem warfare make it essential that geographic locations can be determined quickly and accurately. The US Department of Defence spent a lot of money ($12 billion) to solve this and the result was the Global Positioning System (GPS). This is a satellite-based system which uses radio waves to determine the position of a receiver on the earth's surface. A constellation of 24 satellites, orbiting at about 18,000 kin, circle the Earth and constantly broadcast radio signals to the Earth. The orbits are designed so that at least four satellites are visible from any point on the Earth's surface at any given time. By measuring the time it takes the signals from at least three satellites to reach a receiver on the Earth's surface, it is possible by triangulation to determine the position of the receiver. This has all been made possible by the development of modern computers and highly accurate clocks. The time taken for a signal to reach a receiver from a satellite is proportional to the distance between the receiver and the satellite, given that the speed of radio waves is constant, within certain limits. Thus time and distance can be used interchangeably in describing the triangulation process. If it takes 8 seconds for a signal to reach a receiver then the receiver must lie on a sphere with a radius of 8 seconds from the satellite. If the signal from a second satellite takes l0 seconds to reach the receiver then, similarly it must lie on the surface of a sphere 10 seconds in radius. If the position of both satellites are known then the receiver will be on a circle defined by the intersection of the two spheres. Now if we have a reading from a third satellite then this will define a third sphere which intersects with the other two at two points in space. Thus for an unequivocal answer we actually need to read the signals from four satellites. However this is not necessary because only one of these points will be realistic and GPS receivers use various programs to decide which is the correct point. Three satellite measurements are all that are needed to get a reasonable fix on your position, but it is desirable to measure at least four satellites for more accurate and reliable positioning, particularly where height determination is important. For triangulation to work, it is obvious that the position of the satellites must be known accurately at the time that the readings are taken. But how do we determine exactly the position of something which is 18,000 km out in space. The satellites are placed in very precise orbits and it is possible to calculate exactly where a satellite is in space at any given point in time. Even so, minor variations in the orbit do occur depending on the relative positions of the sun and the moon and these are monitored by earth stations which record these variations and send the corrections back up to the satellite. The satellites then broadcast these corrections to receivers on the ground which use them to calculate the precise positions of the satellites when the readings are taken.
Monitoring Accidentally Released Radionuclides in the Environment
393
The accuracy of a GPS is dependent on a number of factors, such as clock speed, receiver error and atmospheric effects. These errors limit the accuracy of better GPS receivers to about six metres. A more serious error is one induced by the US Department of Defense called selective availability (SA), whereby the significant errors are introduced into the satellite clock. This is to prevent enemy access to the system during times of conflict. This error causes an inaccuracy of some 7.5 m. The combination of these errors with a geometrical principle of "Position Dilution of Precision" will give a minimum average horizontal error for a good receiver of about 30 m but this can be as high as 100 m depending on the relative positions of the satellites being monitored and the degree of selective availability. Readings from four satellites are needed to obtain an estimate of the elevation of a point, and in general terms the error in the height estimate is usually about two to five times that of the horizontal position. This gives an accuracy level perfectly suitable for maritime navigation and other general uses, but is not much use in environmental radiation monitoring surveys. There is however a way of circumventing much of these errors so as to make GPS a useful tool for the type of work we are considering. This is by a technique called "differential GPS". This is based on the idea that two GPS receivers which are within a couple of hundred kilometres of each other will be subject to the same errors, be they natural or induced. If one of the receivers is placed on a known point it can be used as a reference receiver to calculate the error in its estimated position at any given time. If a roving receiver is used to locate points on a survey grid at the same time, the calculated errors of the reference receiver can be used to correct the estimated grid positions. This can be done in real time where the reference receiver transmits the correction directly to the roving receiver, or both receivers can store the data in databases. These are then downloaded onto a PC where the corrections are applied to the roving receiver's data. Using this technique, the geographic locations are determined by a roving receiver with an accuracy of between 2.5 and 5 m, which is perfectly acceptable for most surveys. The time saved in conducting the survey using GPS rather than conventional survey techniques more than offsets the minor loss in accuracy of the grid positions. Let us now describe GIS before going on to discuss how these two technologies complement each other and the benefits to be derived from combining them in the field of environmental radiation monitoring. Information presented in a map can be of two types, spatial and spatially distributed. Spatial data are those which have a form which can be represented by lines on a map such as geological boundaries, roads, rivers and so on. Spatially distributed data are the descriptive data related to the spatial data. An example would be the lithology of the rock type contained within a geological boundary. A shortcoming of conventional maps is the limited amount of spatially distributed data that can be presented on a map. A lithological type could be assigned a particular colour or pattern of shading and described briefly in a legend. A fuller description could then be written up in a report which could be referred to by the person studying the map. An alternative would be to create an electronic database in which is stored the lithological information for all the rock types in the map. This database can then be interrogated to get more detailed
394
Chapter 9
information regarding the rock types shown on the map. A problem here is that one must know the relationship between the spatial data in the map and the spatially distributed data in the database in order to access the appropriate information in the database. A GIS overcomes this problem by establishing a link between the graphical information in the CAD system and an electronic database containing the spatially distributed information. It also allows one to interrogate one or more databases simultaneously in order to highlight relationships between different sets of data for the same area. The greatest advantage of using CAD systems for map presentation is the ability to modify the map quickly and efficiently. Changes to the map can be done on screen and saved immediately without time-consuming redrawing as is the case with paper-based maps. The concept of layers allows one to change the amount of detail shown on a map at will. An analogy with paper based maps is that of having a simple base map with a number of transparent overlays which show various sets of more detailed information. The example of a house plan can demonstrate how this works. A house plan is made up of a number of discrete data sets which are all drawn on the same plan, which can be quite confusing to the viewer. In a CAD system each set of data would be put on a separate layer. The foundations on one layer, the walls on another, the roof structure on another, the electrics on another, the plumbing on another and so on. These layers can be selected in any combination for plotting; thus if an electrician needs a plan, only the electrics and walls need to be plotted because these are the only parts that he needs. The discipline to which this is applied is immaterial. Highly complex digital maps can be created on the computer and the output can be designed such that it meets the specific needs of a particular user. Maps can be created in a CAD system in a number of ways. Drawing directly into the system is the quickest but most inaccurate method. This is only really suitable for sketch maps and maps which contain largely straight lines or fixed curves where the coordinates of the start and end points are accurately known. A more effective method is by digitising, which is an electronic method of tracing. In this method a map containing the required information is placed on a digitising tablet and by means of control points is referenced to the map in the CAD system which the information is to be transferred to. The required information is dependent on the scale of the source map. Digitising can be carried out consistently to within about one millimetre, thus if the source map is at a scale of 1:50 000 then the information captured will be accurate to within 50 m. The required accuracy of the end product will determine the scale of the maps from which the information is captured. Electronic databases are extremely useful and flexible tools for storing, retrieving and analysing a diversity of data sets. The only negative aspect of electronic databases is the time it takes to input data into the system. However, once the system is in place, with careful planning it is possible to maintain and update the system on a routine basis. There is also a growing move by various organisations to provide historical data in digital form which can be input directly into an electronic database. The use of electronic mapping and databases can make one's work quicker and more efficient when handling spatial and spatially distributed information. Using GIS
Monitoring Accidentally Released Radionuclides in the Environment
395
one can create a direct link between a point or area on a map and the related information in a database. It is thus possible to retrieve that information by pointing to that feature in the map on the screen. All that is needed is to specify the information required would be to point at a particular area of a geological map and retrieve the lithological information associated with it. An alternative and more useful approach used when analysing information is to ask the GIS to display those areas which meet or exceed certain criteria. In the field of environmental radiation monitoring a GIS can be of great help in all aspects of a project from planning through to final reporting. If data relative to the parameters of the survey such as population densities, land usage and map information are available they can be input into the GIS, which can then be queried to determine the extent and amount of detail required to achieve the desired result. The CAD part of the GIS can be used to design and lay out optimum grids with coordinates for each grid point. Plots of the grid can be output to guide the field collection of data. Once the data has been collected it can be stored in the databases of the GIS where it can be interpreted and presented. Problematic areas can be highlighted by designing suitable queries and the results used to plan further work where necessary. The GIS can be particularly useful where on-going monitoring takes place. Historical data can be interrogated to provide information regarding trends in measurement with time. In the case of a release such as at Chernobyl, the GIS can be used to monitor the decrease in radiation with time and thus quantify the effectiveness of the remedial actions that are taken. It can also identify areas of concern where further remediation needs to be undertaken. The use of a GIS can thus be seen to increase the effectiveness and efficiency of any environmental radiation monitoring project. If will also increase the professionalism and clarity of the final reporting. From the previous discussion it must be becoming clear how the two technologies of GPS and GIS will complement each other in conducting an environmental radiation monitoring survey. Most of the points have already been made and here we will pull them all together. In the planning stages of a project, the GIS can be used to define the extent of the survey and determine the density and distribution of the sampling to be undertaken. The GPS is not really suitable for laying out detailed grids although it is possible using real-time differential GPS. GPS is more useful for determining the positions of existing grids and for locating sample positions of sample points in more regional surveys. It is also useful in locating positions of infra-structural features which can be added to the map where required. Once the positional data have been collected by the GPS, the data are dumped to a PC and differential correction can be applied if a base station was operating during the survey. The corrected data are then converted to the appropriate projection system and output as a DXF file which can be read by the CAD part of the GIS. In addition, if the radiation measurements are stored digitally, they can be merged with the GPS grid coordinates and loaded into the GIS database. A map of the study area is thus created in
396
Chapter 9
the GIS and the radiation survey data in the database related to it by means of the GPS coordinates. The data can then be plotted on the map in numeric form or contoured if a contouring module is included in the GIS. An image processing module in the GIS allows the generation of sophisticated images of the data distribution which can be enhanced, filtered and sunshaded by various techniques where appropriate. The relationship with other variables such as geology, soil type and land usage can be investigated. Areas where further study is indicated can be identified on the GIS and the coordinates of these areas determined. These coordinates can be fed into the GPS and the areas located in the field by the use of the GPS. The coordinates of further samples can be determined by the GPS and again loaded into the GIS to update the map and database. Geographically accurate maps at any scale, showing specific selected information can be generated for the final report, as can tabulations of the survey results. The maps can be of the raw data or any combination of interpreted spatial or temporal distributions of that data. The use of a GIS allows one to analyse and interpret the data more comprehensively than the use of a more conventional paper-based system, thus enhancing the overall reliability and validity of the results. Many countries have more or less elaborate monitoring networks. Long after the Chernobyl accident several countries installed or extended automatic networks, obviously owing to political and public pressures. It is to be hoped that monitoring on a discontinuous and nuclide-specific basis, which is necessary for assessing the contamination of the environment, food, and the public, will not be neglected in favour of costly automatic networks (which in the case of a severe accident may provide a quick warning, but no information on isotopic composition, chemical form, and contamination of the food chain). Many countries have made provisions for routine monitoring of environmental radiation and for emergency response to a major release of radioactivity to the environment. Not surprisingly, specific provisions vary from country to country according to the country's needs and capacity to fulfil them. A common feature, however, is that most countries with developed monitoring activities have two programmes. 1. A country-wide programme designed to monitor the environment for radioactive contamination, irrespective of its origin, national or otherwise. This programme is usually the responsibility of a government authority on public health (radiation protection) or environmental protection, and frequently involves major inputs from specialised agencies in meteorology and agriculture. 2. A programme which is site-specific to each major establishment in the country dealing with substantial quantities of radioactive material, such as a nuclear power plant, nuclear research reactor, fuel reprocessing plant or radioactive waste facility. This type of programme is designed to monitor the environments of each site for radioactive contamination generated by the establishment operating there. It is usually the responsibility of the establishment at the site and comprises an essential part of its safety system.
Monitoring Accidentally Released Radionuclides in the Environment
397
In many countries, both programmes measure the environmental radiation field and radioactivity in air, water and the various elements of the food chain; the objective is to enable continuous assessments to be made of the impact of radioactivity on health, and on the environment. In some countries, the two programmes are complementary so that country-wide monitoring provides an independent check on site-specific monitoring. National public health and environmental protection authorities are making a significant contribution to controlling environmental radiation. These bodies are expected to play an even greater role in the future. Radiation monitoring varies widely between countries in the adequacy of population coverage, and environmental depth. There are large geographical areas, of great importance when worldwide coverage is sought, that are sparsely monitored, or not monitored at all. The monitoring networks in countries with nuclear power plants have been in operation for a longer period of time. In many other countries attention to environmental radioactivity monitoring has been a result of the Chernobyl accident in 1986. Here we shall describe the situation in some of the countries. France The environmental measurements around French nuclear power plants are described by Le Corre and Bourcier (1996). Electricitd de France generates 75% of its electricity in nuclear power plants with pressurised water reactors (PWR). These plants comprise 34 units of 900 MW and 20 units of 1300 MW, the first of which was connected to the grid in 1977, and the last in 1993. Three other units of 1400 MW are under construction. The environmental measurements are performed in two complementary ways: 1. Routine regulatory monitoring carried out by the operator according to a programme and procedures drawn up by the Central Ionising Radiation Protection Service (SCPRI), which is the State monitoring authority in France. The SCPRI checks the results against those obtained with its own samples. This organisation also monitors radioactivity throughout France (in particular through the TELERAY network). Around each power plant, the following are monitored: 9 ambient 7-radiation (continuously) at 8 points around the site within a radius of 5 km; 9 aerosols in the air (once per day) at 4 points within a radius of 1 km; 9 rainwater and groundwater (monthly); 9 surface waters (whenever there is a liquid radioactive discharge); 9 milk and vegetables (monthly) at 2 points in the area close to the site. The plant has an off-site laboratory, two specially-adapted vehicles and a team of three chemists. These measurements are quite separate from those conducted on radioactive wastes. 2. Annual and ten-yearly radioecological measurement campaigns around the sites in order to improve scientific knowledge of the environmental impact of the plants. A ten-yearly campaign consists of "radioecological photographs" which are compared with the "initial zero point". About 40 samples are taken and
398
Chapter9
various analyses performed (total 15, ct and T spectra, strontium, carbon-14, free and organic tritium). The choice of samples and the places where they are taken depend on the zero point and the special features of the region. Items sampled include drinking water, irrigation water, ground moss, vegetables, fruit, field crops, field soil, humic gely soil, meadow grass, milk, sewage sludge, wine, sediments, water or marine plants, fish or shellfish. The annual campaigns around each site enable a picture to be built up of the radiation situation in time and space from ),-spectrometry measurements on 27 samples selected as being the most suitable indicators (sediments, bryophytes, fish, moss, fruit, drinking water, milk, wine, soil). The programme involves around 600 samples and 800 measurements per year and is conducted by the Institute for Radiation Protection and Nuclear Safety (INPS) which has established the methodology for the sampling and measurements and has the capacity to ensure the continuity indispensable for this type of measurement.
Japan The evolution of environmental radioactivity and radiation measurements in Japan originated essentially in the survey of widespread radioactivity contamination due to the nuclear explosion tests at Bikini atoll on 1st March, 1954. Today, environmental radiation monitoring in the vicinity of nuclear power facilities has become more important than radioactive fallout surveillance, since 38 nuclear power plants are already in operation or under construction in Japan. Environmental radiation monitoring in Japan mainly comprises two different systems. One is radiation measurements and radioactivity analyses of various samples taken from the vicinities of 14 nuclear facilities and the other is a nationwide network consisting of 47 public hygiene institutes of local (prefectural) government with countermeasures against the radioactive fallout due to nuclear explosion tests or a severe accident in a foreign country, such as the Chernobyl accident. For radiation monitoring in the vicinity of nuclear facilities, the monitoring programmes of local governments include the measurements of external T-ray dose and dose rate using thermo-luminiscence dosimeters (TLD) and NaI(T1) detectors, and measurements of T-ray-emitted radionuclides using Ge semiconductor detectors for soil, vegetables, milk, airborne dust, rain water, seawater, seaweeds and shellfishes. In addition to the ordinary environmental radiation monitoring mentioned above, there is a system for prediction of environmental emergency dose information, that is related to the emergency, due to the release of large amounts of radioactive material from a nuclear power plant. This system can carry out the diffusion calculation for radioactive materials based on data from the discharged quantity, meteorological and radiation monitoring data and provides a prediction of the local radiation doses and concentrations to "Local Emergency Headquarters". The system installed in the Nuclear Safety Technology Center (Tokyo) is linked up with STA, local governments and radiation monitoring centres where nuclear power plants are located. In the event of an emergency involving radioactive release from a nuclear power plant, the appropriate
Monitoring AccidentalLy Released Radionuclides in the Environment
399
local governments must take protective action in accordance with the "Local Emergency Plans". About 40 nuclear power plants are presently in operation in Japan and local governments where these are located routinely monitor environmental radioactivity and radiation around sites as a part of the regional environmental safety programs. Results of those analyses and measurements have to be assured in compliance with the principle based on ICRP publication 43 (1984), Chapter "Quality Assurance". The Science and Technology Agency of Japan (STA) commits JCAC to perform maintenance and improve the reliability of monitoring data by means of implementation of the following items. a. Education and training of procedures of environmental radioactivity measurements for personnel of the hygiene institutes of local governments and monitoring center around nuclear power plant. b. Development and publication of 23 manuals of environmental radioactivity measurements including analytical procedures and TLD measurement. c. Establishment of the traceability for radioactivity measurements of local laboratories by the use of standard samples which are prepared by JCAC and Japan Isotope Association Corp., under the reference of standard of Electro Technical Laboratory of Japan. d. Inter-comparison analysis (cross check analysis) with 14 local governments having nuclear facilities and with 33 non-nuclear facilities mentioned below. e. Management and provision of data for nationwide fallout surveillance and radiation monitoring around nuclear facilities. The inter-comparison analysis adopted consists of the "sample dividing method" and the "reference standard sample method". The former method is to divide various environmental samples collected by monitoring laboratories or the hygiene institutes of local government into two parts. One of them is then analysed by such laboratories and the other by J C A C for comparison. The latter method uses reference standard samples which are prepared by JCAC and Japan Isotope Association Corp., by adding appropriate radioactive nuclides to the materials such as aluminum oxide powder or agar. These mock-up samples are analysed by both JCAC and the monitoring laboratories. The sample dividing method is designed to reconfirm the results obtained by monitoring laboratories in regard to sampling, pretreatment (ashing, evaporation, sample dividing, storage), chemical separation and measurements carried out by such laboratories. The reference sample method is designed to evaluate analytical method, level of calibration of equipment and instruments as well as calculations which are employed by such laboratories. The items which are the subjects of inter-comparison between JCAC and local laboratories are relevant to environmental radioactivity and radiation dosimetry that are recommended by the "Guideline for Environmental Radiation Monitoring" (NRC, July, 1983).
400
Chapter 9
The evaluations of both methods are carried out by an evaluation committee to compare the results obtained by the monitoring laboratories with those of JCAC and the amounts of radioactivity added. The criteria of judgement for the sample dividing method are mainly based on whether the difference of results between JCAC and the monitoring laboratories is within 10% + 3 cy (counting error) or not. The criterion for reference sample method is within 10%. If the difference between both results is found in radionuclides analyses using Ge detector and radiochemistry, JCAC will try to find the causes due to inappropriate usage of methods such as sampling, pretreatment, measurements, calculation. Until 1987, numbers of samples for the inter-comparison program were 10-20% of the total samples collected by the local laboratories but were unified to 20 samples for each laboratory. As a result of the 15-year inter-comparison, it has been recognised that inter-comparison is a useful tool for resolving technical problems sometimes encountered and for upgrading the quality of analysis and measurements performed by prefectural hygiene institutes/laboratories. They might also serve to ensure the safety of nuclear power plants.
Finland The description is based on studies done by the Finnish Centre for Radiation and Nuclear Safety (STUK, 1986a, 1986b). In Finland a radiation-monitoring network consisting of approximately 270 measurement stations is run by the Ministry of the Interior and the Finnish Defence Forces. They are equipped with simple Geiger counters and measure every second day. An aerosol measurement network consisting of 10 stations is run by the Finnish Meteorological Institute. It is not nuclide-specific but acts as a warning system. For environmental samples and foodstuff the Finnish Centre for Radiation and Nuclear Safety (STUK) also routinely runs a monitoring program. Aerosols are collected with high-volume samplers in Konala (Helsinki) and north of Helsinki as well as in the vicinity of the nuclear power plants of Loviisa and Olkiluoto. Precipitation is collected with high-surface samplers normally at four stations in the south and west of the country, but there are small samplers at an additional 24 stations. Samples from five major rivers are analysed four times a year; the tritium content of some lakes is also measured. From the surrounding sea nine samples are taken, usually once a year. In addition, bottom sediments and fish samples along the coast are collected. Concerning foodstuff, emphasis is on measuring the radioactivity of milk: milk and dry milk is controlled from several parts of the country and more intensively near Loviisa and Olkiluoto. Samples of wheat and rye as well as beef and pork are gathered from the main production areas; vegetables and fruits are sampled as well. Whole body counting is performed on control groups from Helsinki, Liviisa, and Olkiluoto yearly. Lapps who are a risk group for radiocaesium incorporation are monitored in cooperation with the University of Helsinki.
Monitoring Accidentally Released Radionuclides in the Environment
401
Austria This description is based mainly on a study by Sch6nhofer et al. (1986). In 1957 the first station for aerosol surveillance was installed. In 1986 eight stations were operating. At these sites precipitation is collected on a monthly basis. In 1986 the rivers Danube, Thaya, and March in northern Austria were monitored mostly on the basis of monthly grab samples. Originally gross-activity measurements were used in environmental monitoring, but since 1979 high-volume samplers and nuclide-specific high-resolution gamma spectrometry were introduced as the routine method. In 1986 only a small program of food surveillance was in operation, and it was undergoing reorganisation. Because no nuclear power station is operating in Austria, this surveillance system was mainly to monitor the environmental levels of radiation and to detect discharges from foreign nuclear power stations and from nuclear medicine. It also had the task of preparing for nuclear accidents. In 1975 construction of another system, the Early Warning System, was started. It consists of 336 stations across Austria, which measure the gamma dose rates continuously. Its measuring range is from natural background radiation (approximately 10 ~tR/h) to more than 30 R/h and is divided into eight warning levels. The actual level is reported on-line to centres in the respective federal state and also to the federal warning center. It is intended to provide information for immediate action after explosion of nuclear warheads when external radiation is of much concern. The stations are therefore in populated areas, and no information is possible on the situation in the mountains. The system is not coupled to meteorological systems, which is a drawback. Because no nuclide-specific data can be provided by this system, only the external doses to the population can be estimated.
Sweden The description is based on reports from the Swedish National Institute of Radiation Protection (SSI, 1986a and 1986b). Since the end of the 1950s, 25 stations equipped with ionisation chambers 2.5 m above ground have been in operation by the Swedish National Institute of Radiation Protection (SSI). They register continuously the gamma radiation from both ground and cosmic rays. Only three stations transmit data automatically via telephone to a computer at SSI. The Swedish National Defence Research Institute (FOA), which from 1978 until 1983 was connected to the SSI, runs a system of high-volume aerosol samplers that normally detects very small amounts of radionuclides by high-resolution gamma spectrometry. The FOA also has access to army airplanes and helicopters to take air samples at different heights, to record measurements from the air, and to transport equipment and personnel to remote areas quickly to perform in situ measurements with portable germanium detectors. At the SSI, routine measurements of milk were run before the Chernobyl accident. Routine programs concerning environmental surveillance of nuclear power plants currently exist.
402
Chapter 9
Switzerland The description is based on material from Bundesamt ftir Gesundheitswesen (1986) and Bundesamt fiir Energiewirtschaft (1986). Three warning systems exist. One consists of six early warning stations (FWP) positioned near the border, which measure continuously the aerosol activity. If a preset level is exceeded, an alarm is automatically sounded locally. Seven more stations without automatic alarms are distributed over the country. The second system is NADAM (network for automatic dose alarm and measurement). Twelve NADAM stations were operating at the time of the Chernobyl accident; the operation of all 55 stations was scheduled for the end of 1986. Also in this case an automatic alarm is given if a preset dose rate is exceeded. In the case of high contamination, 111 atomic warning stations (AWP) operated mainly by the police can be activated, but the dose-rate meters used by AWP can only measure dose rates higher than 1 mR/h. Besides these stationary alarms, three cars at different organisations contain the necessary measurement equipment. Additional cars can be equipped to do surveillance. For measurement of foodstuff, drinking water, or fodder, specialised laboratories exist, which also in "normal times" record measurements and regularly take samples in the region, to measure and to communicate the results to the National Alarm Center (NAZ). The army provides, in the case of an alarm situation, personnel, an army laboratory, and a surveillance helicopter.
9.4 EARLY W A R N I N G AND EMERGENCY RESPONSE SYSTEMS
In the case of heavy contamination, early warning is essential. Gamma radiation, which in nearly all cases will be associated with radioactive material emitted in an accident, can be easily measured. Many instrument systems can measure dose rates caused by gamma radiation from environmental levels (and therefore well below any critical dose rate) up to extremely high levels. The higher the level, the faster and easier it can be measured. Therefore, on the instrumentation side, the requirements for fast early warning can be met. Another question is the density of the measuring station network, which obviously depends not solely on the geographical parameters of a given country but also on political and financial considerations. This is easily demonstrated by the fact that Austria regarded 336 stations on 83,855 km 2 as necessary (Schrnhofer et al., 1986), while in Sweden 25 were in operation on an area of 449,964 km 2 at the time of the Chernobyl accident. The combination of dose-rate measurements with meteorological parameters, linkages of these data via computers to produce isolines of contamination, and forecasts of the contamination situation are regarded by most experts not only as easily achievable by appropriate systems, but also as absolutely necessary. Modem environmental monitoring systems of the powerful nuclear power industry and technology enterprises fulfil three functions:
Monitoring Accidentalh, Released Radionuclides in the Environment
403
1. Direct registration of ionising radiation in the environment in the dosimetric points disposed in the location around the enterprises, these points being supplied with microprocessing equipment controlled by sensors, having telephone or radio communication with the central computer. 2. Registration of meteorological parameters and mathematical description of radionuclide transfer in the atmosphere by calculating in real time. 3. Transmission of the on-line information for users. All this enables us to obtain reliable and full information under the normal work regime of a nuclear power plant and in the case of increased radionuclide emission into the environment (incidents, emergencies). The necessity and the role of Emergency Response Systems in a nuclear reactor accident can be understood from the following arguments. The protective measures which are available to avoid or reduce radiation dose can be taken by 9 sheltering, 9 stable iodine administration, 9 evacuation, 9 relocation, 9 control of access, 9 decontamination of individuals, land and property, and 9 control of distribution of foodstuff and water. It is clear that the risks, difficulties and disruption which follow the implementation of these various protective measures are widely different and depend on many factors, including the location of the site and the meteorological conditions at the time of the accident. Thus, if the projected dose information can be provided by a real-time Emergency Response System (ERS) by using various information on the accident plant and meteorological conditions around the site, it must be effective information for countermeasures. In particular, it will be important in the early stage when the information about the nuclear reactor and the environmental aspects is insufficient to grasp the accident situation. In the event of an accident at a nuclear facility, the role of ERS is to assist the protective measures taken. There are three convenient time phases in accident sequences, which provide for different considerations to apply to decision-making about off-site action (Clark, 1986). These are termed early, intermediate and late (recovery) phases. These basic principles are adopted in IAEA, ICRP and WHO. The early phase is defined by the time period during which there is the threat of a significant release. The time interval between the recognition of an accident sequence and the start of the release can extend from about half an hour to about a day and the duration of the release may be between half an hour and several days. In this phase the information based on the analysis of data and predictions being from the nuclear installation and some limited environmental measurements of off-site exposure rates and airborne concentrations from the plume may become available. Thus, the
404
Chapter 9
prediction calculation in the early phase is required especially for decision-making about the protective measures to be taken with the public. The intermediate phase covers the period which starts from the first few hours after the commencement of the release and could extend for several days or weeks. During this phase, environmental measurements of radiation levels from deposited radioactive materials, as well as levels of radioactive contaminants in food, water and air, will become available. It is also during the intermediate phase that the plant is expected to be restored to a safe condition and the protective measures, based on the environmental measurements, will be implemented. In this phase the results of the prediction calculations by ESR will be very important to design and check the environmental monitoring plans and to carry out the appropriate evaluations of environmental consequences in conjunction with environmental monitoring activities. The late phase may extend from some weeks to several years after the accident, the duration depending upon the nature and magnitude of the release. During this phase the data obtained from environmental monitoring can be used to make decisions on returning to normal living conditions. After the late phase, ERS will be useful for the detailed analysis of environmental consequences and doses, using real meteorological information and source term and the release information. After the accident at TMI-2 reactor in the United States in March 1979, many emergency response systems to assess environmental consequences have been developed in various countries; moreover accuracy and response speed of these systems are remarkably improved, supported by the computer technology which has been considerably developed in the last ten years. These systems are divided into two types, i.e., centralised and localised systems. The centralised systems are generally owned by national governments and have a role to serve and advise all of the plants in the country, while the localised systems are located in the vicinity of specific sites. The data communication network, computer and software of the centralised system are larger than those of the localised one. This results from the difference in the complexity of computational models for the diffusion of radioactivity in each system. Since it is necessary for the centralised system to simulate the transport-diffusion of plume on any type of site, the models must account for the effect of topography, sea and so on. Corresponding to this, the amount of data treated in the centralised system becomes larger. Meteorological data used by the centralised system are supplied from the national weather centre which provides data observed all over the country in addition to the site data, while those used by the localised system are supplied from the site. Furthermore, the centralised system needs topographical data all over the country. In order to manage this software, the large-scale computer is generally used in the centralised system. On the other hand, the localised system must use simplified models, because the localised system uses a small-scale computer or minicomputer from the viewpoint of cost-effectiveness. Therefore, the major characteristics of the localised system are the low cost and the high-speed response.
Monitoring Accidentally Released Radionuclides in the Environment
405
Let us outline a centralised emergency response system in Japan called SPEEDI (Imai et al., 1985). When an accident happens at a certain nuclear facility in Japan, the operator inputs data for site specification, such as the site location, reactor type and the computational domain. The computational domain can be selected from local scale (25x25 km) and regional scale (100xl00 km). Then sequential calculations of wind field, concentration of radioactive materials and dose are carried out under the direction of the operator. In these calculations, the time-dependent source term information, such as release height and release rate of each nuclide, can be specified as input data. If the source term data are insufficient, the concentration and dose calculation are carried out with the assumption of continuous unit release. In the case that only the gross information of source term, such as the total amount of the released radioactivity, is obtained, the fraction of each nuclide is deduced on the basis of burn-up rate. The meteorological data used in these calculations are collected routinely through a data communication system. The invariable data, such as the site characteristics, nuclide data and topographical data are stored in the database. The results of the computation, such as the wind field, concentration and dose distribution, are stored in the magnetic disc and are displayed on the graphic display together with some map elements. The user can select several kinds of map elements, such as administrative boundaries, coast lines, road/railroad, topography and locations of towns. The system can display the following information: 9 observed wind data, 9 horizontal or cross section of the computed wind field, 9 concentration distribution at the selected height, 9 surface deposition of radioiodine, 9 external gamma-dose-equivalent, 9 thyroid-dose-equivalent due to the inhalation of airborne radioiodines, 9 temporal variation of concentration or dose at a fixed point, and 9 area where the dose exceeds a selected value. In order to use SPEEDI effectively in practical emergency countermeasures, it is essential to collect meteorological and radiation monitoring data on time. In Japan, hourly meteorological data are observed automatically by AMeDAS (Automated Meteorological Data Acquisition System), which is operated by the Japan Meteorological Agency. Those data are supplied to SPEEDI through MICOS (Meteorological Information Confidential Online Service) by the Japan Weather Association. The local meteorological data and the radiation data are obtained by the monitoring stations of prefectural government, which are also supplied to SPEEDI. The construction of the SPEEDI network was started in 1986, in order to establish the data communication lines between the SPEEDI operations center, the Science and Technology Agency (STA) and prefectural governments in which nuclear plants are placed, according to a national plan for the improvement of the national emergency countermeasure system in Japan. The data communication network system of SPEEDI consists of five main organisations, which are
Chapter 9
406
9 9 9 9
SPEEDI operations centre, Science and Technology Agency. JAERI computer centre, radiation monitoring and/or emergency operation centre of each prefectural government, and 9 Japan Meteorological Agency. The SPEEDI operations centre is located in Tokyo near the STA. Its basic functions are: 9 collection of meteorological and radiation monitoring data, which are essential for the prediction calculation and source term estimation by SPEEDI, 9 control of data communication network system and SPEEDI calculation carried out in the JAERI computer center, and 9 distribution of calculated results and the collected data to the STA and emergency operation centres established in prefectural government. If an accidental release of radionuclides occurs, intervention may be required to avoid doses (IAEA-1996). At the time of the accident, it may be necessary to implement urgent countermeasures, such as evacuation or sheltering, to avoid short-term, relatively high doses. It is important that a disruptive countermeasure such as evacuation should not be implemented to avoid trivial doses; the advantage of the dose saving should be commensurate with the disadvantages of the disruption (Fry, 1996). For this reason, for example, the National Radiological Protection Board (NRPB) in the UK advise that evacuation is unlikely to be justified unless the dose averted is at least a few tens of millisievert (NRPB-1990). A less disruptive countermeasure, such as sheltering, would be justified for a smaller dose saving, say a few millisievert. Sheltering or evacuation, which are expected to be short-term removal of people from their homes, raises different issues. If relocation is contemplated, it would be to provide protection, not against a short-term hazard, but against a level of exposure that is considered unacceptable because of its long duration. The NRPB of the UK has noted the international guidelines on relocation given by ICRP (ICRP-1991) and being developed by IAEA, and has been consulting widely in the UK with government departments and other interested parties such as local authorities and the police. A number of points have arisen from these discussions and from the experience of the Chernobyl reactor accident. Criteria must be developed, promulgated and accepted in advance of any accident. Relocation is expensive, in monetary terms, and it is stressful. It is necessary to take full account of social factors when making decisions on relocation and the people affected should be involved in the decision. These, and other issues, lead us to the view that relocation is a countermeasure of last resort and would only be taken to avoid doses of around 15 mSv -~ or more. For smaller dose savings, other procedures such as decontamination should be considered so as to reduce doses, avoid restriction on the individual's lifestyle and so promote a return to "normality" (Fry, 1996). Similar considerations apply to situations where contamination from past practices is discovered on land to which the public has access. Simple procedures would be
Monitoring Accidentally Released Radionuclides in the Environment
407
appropriate to avert doses of a few millisievert, but substantial disruption to people's lives would be justified only to avoid doses around 15 mSv y-~. Some perspective can be gained from ICRP's recommendations on radon in homes (ICRP-1993), where it is recommended that the action level for simple remedial procedures should lie in the range 3-10 mSv y-~. It is of interest to mention the possibility of aerial surveying for detailed mapping of a radiation-contaminated area. We shall describe two attempts at aerial surveying, one in France (Bourgeois et al., 1996) and one in Ukraine (Shkvorets et al., 1996). The French Atomic Energy Commission (CEA), Valduc Centre, has developed an aerial system of gamma cartography named HELINBUC. This equipment enables, in a few hours, the establishment of a map of radioactivity over areas several dozen to several hundreds of hectares in size, by identifying radioelements present, with a sensitivity between the level of natural radioactivity and that of artificial radioactivity resulting from a large-scale accident. HELINUC has been operational for about ten years and is part of the French intervention system in the event of a civil or military nuclear accident. About seventy nuclear or industrial sites, civil or military, have been mapped, either in the framework of systematic surveillance measurements, or during intervention exercises during tests and training. The HELINUC system is mounted on a helicopter (equipped with a NaI detector) and linked to ground-based equipment (with a germanium detector). The detection limits by aerial means are (Bourgeois et al., 1996): 241Am 15 kBq/m 2 extended source of 2000 m 2 137C86~ 2-kBq/m 2 extended source of 2000 m 2 137Cs 10 MBq point source Aerial gamma mapping allows us to bring into the evidence, apart from artificial gamma emitters (fission and activation products), the three natural radioelement families, K, U, and Th. It is of interest to mention here the aerial system used in 1992 for the mapping of the radioactive contamination of the region around Chernobyl NPP (Shkvorets et al., 1996). In the survey the data were obtained by the gamma-spectrometer system installed on a helicopter. Measurements were made by scanning the territory at a height of 100 m and a distance between the flights of 250 m. The velocity of the helicopter was about 100 km/h. Calibration coefficients for evaluating of flight data were measured at various heights and on various types of landscapes with different distributions of radionuclides in the soil. The configuration of the equipment used for the aerial survey of radiation is shown in Fig. 9.4. The total count rate of NaI(T1), height of flight and navigation data are recorded once each second. The count rate information is used for location of lost sources and small radioactive spots. Spectral information is recorded periodically. Time of spectra acquisition depends upon sensitivity of HPGe and one was chosen upon 10-20 seconds for condition of intensity of radiation in Chernobyl region. This period allows measurement of a minimum activity of the ground contamination of ~37Cs: 0.5 Ci/km 2
Chapter 9
408
Fig. 9.4. Airborne system for radioactivity measurement.
with accuracy 30%. The operator observes visually the type of landscape (i.e. forest, water, agricultural area) and indices of landscapes are recorded by the acquisition system. 9.5 SAMPLE C O L L E C T I O N AND P R E P A R A T I O N Collection of samples, or sampling, is the method (or procedure) of extracting samples for the purpose of measuring the characteristics which are surveyed. Environmental radiation monitoring is mainly conducted with the aims of estimating an exposure dose for people near nuclear power facilities and of protecting public health and safety. In this case it is necessary to determine monitoring items, emphasising the processes that result in individual exposure. These are based on the behaviour of radionuclides and on the following information: 1. Population distribution by direction, distance, community and age 2. Topography and geology around nuclear power facilities 3. Topography and geology of the seabed around nuclear power facilities 4. Marine phenomena at a waste disposal seaport 5. Atmospheric phenomena at a waste disposal seaport 6. The types and quantities and the seasons food is to be collected near nuclear power facilities 7. Distribution routes and intake of foods in the area 8. Characteristics of foods around nuclear power facilities 9. Information on the behaviour of radionuclides in the environment including in living things 10. Information on exposure evaluation of the human body
409
Monitoring Accidentally Released Radionuclides in the Environment
I I
,
,
Cosmic rays
I
I
'
~
I
m
m
m
I
I
I
L.
II Radioactive gaseous waste
t
/ .I / Natural radioactive substances or radioactive substances due to nuclear testing
"~ 7
ii
// it
Soil
Air ......~I~ ~ ' . External . %radiations . -- DI~ | "~1 ~ [Airborne I -~ ~] gD ('~~)'N,{dust -e] Respiratorytract ..~l / D I Fallout I . ' ~ (~, " ~ agricultural] Plants, [ m ~ 0 Soil '[[ Ab~olpuu. .... :.._"l i . . ;- ~ ~, I ~ [~priaucts (~)-1 M~ ! f
Animals
(~)
te
,wD~!ing
Mouth
~
Mouth
Fig. 9.5. Main paths of exposure of the human body through radioactive gaseous waste and items to be monitored. Figures 9.5 and 9.6 show examples of monitoring items viewed from the path of exposure. There are also some general procedures that one should follow. For example, double identities should be placed on samples at collection time. It is advisable to fill in a standard form with all relevant information (date, location, fresh weight, weather, collector's name, etc.). After collection, the samples must be properly stored to avoid degradation, spoiling, or other decomposition, and to avoid contamination. Proper care must be taken to avoid loss of volatile radionuclides. Short periods of storage before analysis may require refrigeration, or the addition of a preservative such as sodium bisulphite, alcohol or formalin (as in the case of milk) for biological samples. When long periods of storage are needed, it may be preferable to convert the samples to a more stable form immediately after sampling. Drying or ashing of the sample will allow extended storage, but the temperature must be carefully controlled in these operations to avoid loss of radionuclides. Sample containers must be suitable for storage without degradation especially when acids are added to liquid samples. The absorption of radionuclides from solution is less on polyethylene than on glass. With a few exceptions, almost all sorption losses can be eliminated by the addition of acid, a carrier solution containing stable elements, or a complexing agent.
410
Chapter 9
@
O)
Radioactive liquid waste ~
Jcater I ~~ariie
q Edible......I....Mouth life[
S
.... Na~raf~aa~oaci~ve'i S
.
substancesor ~ radioactivesubstances : due to , nuclear testing !
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
Nonedible ~~ ~ ] ._o ~ m a r i n e l i f e ] ~~ ~_ _
~ ~= ]Su~mfarine]~
-
@4
,~-
f
isai entt
External radiations
Fig. 9.6. Main paths of exposure of the human body through radioactive liquid waste and items to be monitored. The sample collection equipment, containers, and sample preparation areas must be kept clean to avoid contamination, disposable containers should be used whenever possible (plastic, bags, aluminium trays, etc.). Drying reduces the weight and volume of the samples and may also permit a longer storage time. Samples may be dried in a low-temperature oven at 105~ or at room temperature without significant loss of any radionuclides except iodine. Samples should be dried for a sufficient period of time, at a fixed temperature, to acquire a constant dry weight. Measurements of fresh or wet weight and the dry weight are required. It is important to prevent contamination during the drying procedure. If necessary, freeze-drying may be used to further reduce the loss of volatile radionuclides from the sample. However, this process is very time-consuming and is therefore not highly recommended. Evaporation is the normal method of concentrating liquid samples. Reasonable care is required when evaporating liquids, particularly milk, with a hotplate in order to avoid spattering and loss of sample. Evaporation lamps usually eliminate the problem of spattering. The evaporation bowl should be made of material that will not absorb the radionuclides. Some radionuclides, such as iodine, tritium and ruthenium, may be lost during the evaporation process. A fast evaporation can be performed satisfactorily using a rotating evaporation system that operates under reduced pressure. Different volumes of the rotating spheres of up to 30 1 are available. Where samples need ashing, low-carbon-nickel trays are adequate for the ashing operations. However, other trays lined with thin-sheet aluminum, which is discarded after each use, may be entirely satisfactory. Trays are easily cleaned with detergents of dilute mineral acids (usually HCI). The temperature for dry-ashing varies but an upper limit of 450~ is recommended. If the sample is not completely dry at the start, an
Moniwring Accidentally Released Radionuclides in the Environment
411
initial drying step at 105~ should be introduced. The ashing time depends on the type and quantity of the material; large samples may require 16-24 h. Dry-ashing should be used only for radionuclides that do not vaporise at the ashing temperature. Significant loss of caesium will occur above 400~ Carrier elements and radioisotope tracers should be added to all sample types before ashing. Measurements of the ashed weight are necessary for calculation of the radionuclide concentrations and yields. We now consider some specific materials in some detail. 9.5.1 Air Air provides an important pathway through which humans are exposed, by inhalation, to a number of radionuclides. Air also conveys airborne radionuclides that were once sedimented in soil or on plants. Radionuclides that then reach humans through the respiratory system, digestive system, or skin cause both intemal and external exposures. For the analysis of radionuclides in airborne dust, the dust is collected on a filter using a dust sampler. Iodine in the air is collected on an active carbon filter using a dust sampler. Tritium exists in the form of vapour (HTO) or gas (HT) in the air. The HTO is absorbed on silica gel: HT is changed to HTO using a palladium catalyst and then the HTO absorbed on silica gel. Radioactive noble gases such as 85Kr are absorbed on an active-carbon trap cooled with liquid nitrogen. Several types of filter material are used for collecting aerosol materials (glass, PVC or Microsorban filters). All commercial filter media, when used properly, have adequate efficiencies. The filters are usually compressed to provide a standard counting geometry and are measured by gamma spectrometry, after which they may be dry- or wet-ashed for radiochemical analysis. Air particulate samplers are usually classified as low-volume air samplers or high-volume air samplers. There are, in addition to these classifications, dust samplers that consist of a combination of a low-volume suction pump and a movable filter-paper system. Characteristics of these samplers are as follows: 1. Low-volume air sampler: A low-volume air sampler is an apparatus having a suction capacity of up to 201/min. It is used for one continuous sampling lasting from several days to 1 week. Filter papers having a diameter of 5 cm and an active-carbon cartridge can be attached as a collecting device. 2. High-volume air sampler: A high-volume air sampler is an apparatus whose suction capacity is between 500 1/min and 2000 1/min. It is used for a sample period of 1 day. A filter paper of dimensions 203x253 mm (8x10 in) can be attached as a collecting device. 3. Dust sampler: A dust sampler has a suction pump with the same volume as a low-volume air sampler. It is capable of continuous sampling during 1 month when using an attached long filter paper. Most dust samplers used by local self-governing bodies are specially made to have an attached active-carbon cartridge in addition to the long filter paper. They are capable of measuring total beta and alpha radioactivities and the iodine content of the air.
412
Chapter 9
In addition to a proper choice of collecting material (filter paper), a reliable measurement of flow rate is required. Flow meters are classified into rotameters and integrating flow meters. The latter are further classified into wet-gas meters and dry-gas meters. A rotameter has a specially graduated vertical tube, whose diameter increases in the ascending direction, containing a spinning top-shaped or spherical float. A gas-stream is admitted into the bottom of the tube and the float is held at a vertical position which varies in proportion to the flow rate of gas. 9.5.2 Water
Tap-water should be collected at the water processing (filtration/purification) plants just prior to discharge into the distribution system. If the water is to be collected from a residence, then the pipes should be flushed sufficiently (2 or 3 min) prior to sample collection. Rain collectors 0. I-1 m 2 in area provide adequate collection of rain-water. Automatic sampling devices are commercially available which protect the collector from dry-deposition prior to the rainfall. These samplers start to open the collection area when rain begins to fall and close it when the rain stops. High-walled vessels with smooth surfaces are equally suitable. Some loss of the less-soluble radionuclides will occur on either of these collectors but the loss can be largely recovered (if desired) by washing with dilute acid (0.1 N HC1). An alternative method is to filter the water directly through a mixed-bed ion-exchange column, after which the water is drained away. Contamination of rain-water samples by airborne soil and surface dust can be minimised by locating the sampling stations on the roofs of buildings. Overhanging vegetation should be avoided. The most suitable size for the collector depends upon the amount and frequency of precipitation in the area, as well as the frequency of collection. If water samples have to be stored for any length of time, hydrochloric acid (11 M) should be added to the sample bottles at the rate of 10 ml per litre of sample either prior to sampling or as soon as possible afterwards to avoid absorption of the radionuclides on the walls of the container. The longer the storage time before analysis the more important it is to acidify the water samples. In addition to the radioactivity analysis of the samples, other information is required, including the: 9 atmospheric conditions (weather and surface air temperature); 9 water temperature, pH, salinity and degree of clarity; 9 location (direction and distance from a navigational mark), latitude. 9.5.3 Soil
It is important to identify the radioactive concentrations in soil because it constitutes a path for radioactivity to humans, animals and plants, and is an indicator of radioactive accumulation in the environment. Soil includes submarine sediment and river-bed soil,
Monitoring Accidentally Released Radionuclides in the Environment
413
in a broad sense, but here it includes only soil from uncultivated and cultivated land. The soil to be measured should consist of particles having diameters of 2 mm or less. Sampling locations should not have obstacles nearby (trees, structures) and the utilisation of the land should be considered. Also, sampling locations should not have unusual soil quality or topography, and should have little vegetation. Locations should be selected for periodic sampling to be possible, in order to determine the accumulation of radioactivity. Samples of earth transported from another place should be avoided, even when the soils have been mixed. Maps of the sampling locations should be sketched or photographed whenever possible. Samples should be collected from the surface layer 0-5 cm deep with a soil sampler having a diameter of 10 cm at five to eight locations. Submarine sediment is important in understanding the accumulation of radionuclides discharged with waste water from nuclear power facilities. Grains analysed should have a diameter of 2 mm or less. Collect samples at the outlet of a facility drainage duct. Also collect supplementary survey samples offshore. Refer to marine charts or consult fishermen familiar with the region because sampling may sometimes be hindered by a bedrock, even though the sampling location may have been selected considering ocean current. 9.5.4 Biota
Plants take in radionuclides discharged into the environment, in turn people eat these plants or take them in through animals that have directly or indirectly eaten the plants. It is therefore important to measure the concentration of radioactivity in plants and animals when evaluating exposure dose of humans. Measurements should be made on milk, a major food for infants and a daily food for many people, to directly estimate internal exposure dose. Measurements should also be made on indicator plants and animals, which are not edible or directly involved in the human food chain. These indicators grow readily, concentrating radionuclides. Thus, these indicators near nuclear power facilities are very useful for monitoring changes in the level of environmental radioactivity. Table 9.6 shows biotic items suitable for sampling. Table 9.7 summarises the recommendations for collecting of samples. Next we shall describe some of the objectives used in sample pre-treatment. The objective of these procedures is to reduce the volume of the samples. Portioning, evaporation concentration, chemical separation, absorption and so forth are techniques used alone or in an appropriate combination for liquid samples. Drying, sieving, pulverisation, mixing, reduction, ashing and so forth are techniques used alone or in an appropriate combination, for preparing samples of solids for measurement. Sample pre-treatment procedures depend on the type of samples and type of activity to be measured. Let us mention some of the pre-treatment procedures for gamma-ray spectroscopy. Figures 9.7 and 9.8 show the sequence of steps to be taken during
414
Chapter 9
Table 9.6 Biotic items suitable for sampling
Agricultural and dairy, products Cereal grain Leaf vegetables Root crops Fruit vegetables Potatoes Tea Grass Milk and Livestock products Indicator plants Daily food
rice, wheat, corn, buckwheat, etc. spinach, Chinese cabbages, cabbages, etc. Japanese radishes, carrots, etc. tomatoes, orange, apple, grape, etc. taros, potatoes, sweet potatoes, etc. green tea, black tea, etc. Italian ryegrass, dent corn, etc. raw milk, milk from the market, cheese, meat, etc. mugworts, pine needles, cedar leaves, etc.
Aquatic life and products (sea and fresh water) Fish Arthropods Echinoderm Prochordates Mollusca-Gastropoda Cephalopoda Seaweeds Indicator plant and animal
barnacles, lobsters, crabs, etc. Holothuroidea, sea urchins, starfish, etc. ascidians ear shells, turban shells, short-necked clams, clams, etc. cuttlefish, octopuses, etc. gulfweeds, Eisenia bycyclis, Undaria pinnatifida, tangles, etc. gulfweeds, blue mussel, etc.
Soil sample Weighing (wet weight) Drying (105~ Weighing dried soil (dried soil weight) Sieving (2 mm and under) Weighing (dried fine soil weight) Pulverisaton (top Grinder) no treatment for sand Weighing Mixing (V blender, etc.) Sample for analysis and measurement Fig. 9.7. Sequence of steps to be taken for preparation of soil samples for y-spectrometry.
Monitoring Accidentally Released Radionuclides in the Environment
415
Table 9.7 Recommendations to obtain good samples Sample
Recommendation
Cereal grain
Collect samples from one location. It is desirable to collect samples of a single species or a single kind.
Vegetables
Purchase them from an agricultural co-operative association or directly from farmers. Collect samples from one location. It is desirable to collect samples of a single species or a kind. Collect individual vegetables of average size grown in an open field, preferably at the centre of it.
Tea
Purchase from an agricultural co-operative association or directly from farmers. Collect samples from one location. It is desirable to collect samples of a single species or kind. Generally, processed tea is bought but flesh tea leaves are sometimes collected.
Dairy products
Purchase them from an agricultural co-operative association or directly from farmers. It is desirable to collect samples from milk cows of a single species at one location. After the Chernobyl accident, the level of radioactivity was high in milk from cows fed with green grass. However, few milk cows are now fed with green grass.
Grass
Purchase it from an agricultural co-operative association or directly from farmers. It is desirable to collect samples from a single
Indicator plants
Select those species that can be collected regularly over a long period. Collect samples from a single location. Individual differences are great in mugworts from season to season. It is difficult to distinguish the leaves from the twigs of a cedar. It is believed that the level of radioactivity in pine needles differs according to age. i.e., one or two years old.
Total diet
Measurements of these foods are not presently made as part of monitoring near nuclear power facilities.
Fish and shellfish Ask a fishermen's co-operative association to collect them. It is desirable to collect samples of almost the same size. Select sedentary fishes. Cuttlefish, octopuses Crabs, lobsters
It is desirable to identify species, though it is difficult to do so. It is often difficult to sample the same species continuously because catches fluctuate.
Other samples
They are expensive when catches are small; this raises purchase costs.
Seaweeds
Ask a fishermen's co-operative association to collect them. It is desirable to identify species, though it is difficult to do so. It is sometimes impossible to sample them during certain periods of the year. Varieties of seaweeds change with water temperature or waste water.
416
Chapter 9
Agricultural Products Weighing Pre-treatment (washing with water, removing (roots or cores, peeling, etc.) Fresh weight measurement
Portioning (put 1 kg of each sample in a 30-ram round-bottomed porcelain vessel)
Drying (in a hot air drier at 105~
Pulverisation (top Grinder)
Weighing ash (dryng rate)
Carbonising and ashing (450~
24 hours)
Ash weight measurement (ash content percent)
Sieving and mixing (screening at 0.35 ram)
Sample for analysis and measurement Fig. 9.8. Sequence of steps to be taken for preparation of agricultural products for y-spectrometry.
preparation of soil samples and agricultural product samples for gamma spectrometry; determination of 131I concentration is often required. There we present procedures for its determination in seawater and milk samples. Iodine in seawater should be collected as silver iodine following the steps shown in Fig. 9.9. In the case of iodine in milk, milk should be collected in ion exchange refill following the procedure shown in Fig. 9.1 0.
Monitoring Accidentally Released Radionuclides in the Environment
417
seawater sample 5 liters Wash the vessel (Cubitainer) with 100 mL of nitric acid (3+ 11) add 20 mL of sulfuric acid (1+5) add 40 mL of hydrazine sulfate (saturated) stir (about one minute) stand (more than three hours) add 1.6 grams of silver citrate (0.45 micrometers) stand (more than 12 hours)
decant Precipitate Supernatant (discarded)
filter using membrane filter (0.45 micrometers)
Filtrate (discarded)
Precipitate
Sample for measurement (vessel U-8)
Fig. 9.9. Pre-treatment of seawater for '3'I y-spectrometry.
9.6 M E A S U R E M E N T S
OF AIRBORNE RADIOACTIVITY
9.6.1 Measurement of particulates Radiation monitoring equipment covers a variety of models designed for a specific purpose. In this section we shall briefly discuss measurement of airborne radioactivity. Radionuclides are released into the atmosphere from operating the various facilities. These radionuclides are dispersed to populated areas where exposure occurs by breathing or swallowing the materials.
418
Chapter 9
Sample (milk) 4 liters
add 10 mL of sodium iodide solution (150 mgI/mL) add 2 or 3 drops of sodium sulfite solution (10 w/v %) stir add 42 mL of anion exchange resin stir for 20 minutes stand (about 10 minutes) decant
!
Supernatant
Resin
add 42 mL of exchange resin stand (about 10 minutes)
stir for 20 minutes
decant
Supernatant (discarded)
Anion exchange resin (mixed into the sample in a 500 mL beaker) wash (warm water) Sample for measurement (Vessel U-8) Fig. 9.10. Pre-treatment of milk for '3~I y-spectrometry. Measurement of airborne radioactivity provides the first opportunity of identifying the spectrum of radionuclides making up the contamination. Radionuclides will appear very rapidly in ground-level air, and air samples can give the first indication of the nature of the contamination. Radioactive materials in the air may result in exposure to man by inhalation, by ingestion of particulate matter deposited on vegetation, or by ingestion of products derived from animals which were exposed to radioactive materials through inhalation or ingestion. The most probable internal exposure pathway for the workers in a radiation controlled area is inhalation of radioactivities in air. The objectives of an air monitoring program in a radiation controlled area are as follows:
Monitoring Accidentally Released Radionuclides in the Environment
419
1. to assess the probable upper limit of the inhalation of radioactive materials by workers; 2. to monitor the unexpected airborne contamination for the prevention of undue inhalation; and 3. to provide information needed for the planning of individual monitoring programmes to prevent internal exposure. Monitoring of the workplace for air contamination will almost always be needed on a routine basis in the following circumstances: 1. when gaseous or volatile materials are handled in quantity, e.g. tritium and its compounds in large scale production processes and as the oxide in heavy water reactors; 2. the handling of any radioactive material (including reactor fuel fabrication and reprocessing and the machining of natural and enriched uranium) in conditions of frequent and substantial contamination of workplaces; 3. the processing of plutonium and other transuranic elements; 4. uranium mining, milling and refining; 5. the handling of unsealed radionuclides in hospitals in therapeutic quantities and the use of hot cells and reactors and critical facilities. The most common form of monitoring for air contamination is by using samplers at a number of selected locations intended to be reasonably representative of the breathing zone of the workers. Usual sampling and measuring methods of airborne radioactivity are shown in Table 9.8. A dust monitor or a dust sampler with a sampling filter paper is used for the monitoring of airborne particulate radioactivities. Filter paper of cellulose glass fibre and of glass fibre is widely used. A dust monitor is composed of a dust sampling part with a filter paper, an air suction part with a pressure and a flow gauge and a measuring part with a detector and a warning system. There are two types of dust monitor; one is the fixed filter type and another is the moving filter type. As the filter paper is incapable of sampling the volatile radioactivities such as iodine, activated carbon is used for the sampling of such radioactivities. Filter paper impregnated with activated carbon and a cartridge filled with layers of activated carbon granules are available. An ionization chamber is commonly used as a gas monitor for the monitoring of gaseous radioactivities such as noble gases and tritium (vapour) in air. The radioactivity concentration in air is estimated from the ionizing current caused by the radioactivities in air flowing through the ionization chamber. As a gas monitor sometimes gives an enhanced indication when it sucks air-containing ions such as cigarette smoke, an ion trap is provided at the inlet of the chamber. For the sampling of tritiated water vapour in air, a cold trap or water bubbler is used. The sampled water containing tritium is measured with a liquid scintillation counter. Although gaseous and liquid wastes originating from the operation of nuclear facilities and radioactivity treatment facilities etc. are allowed to be released in
420
Chapter 9
Table 9.8 Sampling and measuring methods of airborne radioactive materials Sampling method (media used)
Properties of the material
Nuclides sampled
Measuring method
Filter paper
Particulate
6~
Gross measurement and/or spectrum analysis of a, [3, y radiation. Fluorescence analysis.
Filter paper impregnated activated carbon
Gaseous (volatile material)
131I, 35S, e~
Gross 13-and y-ray measurement "/-ray spectrum analysis
Activated carbon cartridge
Gaseous (volatile)
131I,2~
Gross ),-ray measurement ),-ray spectrum analysis
Silica-gel
Vapour
3H
Measured by liquid scintillation counter
Ionizing chamber for gas sampling Sampling chamber
Gaseous
Noble gas 3H, Noble gas
Cold trap
Vapour
3H
Water bubbler
Vapour, CO 2 mist
3H,
U, Pu
14C
Gross o~-and 13-ray measurement (Measurement of ionization current) Gross y-ray measurement '/-ray spectrum analysis Measured by liquid scintillation counter
14C
Measured by liquid scintillation counter
compliance with the environmental regulatory standard (1 mSv/a), it is very important to control and restrict the release from the point of view of environmental safety. The objectives of radioactive effluent monitoring are: 1. to know the released amount of radioactive materials and then to confirm that it does not exceed the authorized limit for release; 2. to detect the abnormal release; and 3. to provide information for estimating the dose equivalent around the facility due to the released radioactive materials. The ventilated air and gaseous waste from the controlled area are released through the stack after dealing with high efficiency particulate air filters. There are two methods of continuously monitoring the concentration in exhaust gas. One is to measure the gas directly with a detector inserted into the stack or duct and the other is to measure a portion sampled from the exhaust air with a dust monitor or a gas monitor. The purpose of the air radioactivity on-line measurement in the environment is to quickly detect a too-elevated activity concentration of ot and/or [~ artificial radionuclides. In the case of a nuclear accident, this measurement will help the authorities to take a decision accounting for the health aspect.
Monitoring Accidentally Released Radionuclides in the Environment
421
Fig. 9.11. Particle deposition in pulmonary tract as a function of particle diameter.
To perform this objective, the instruments must: 9 be very sensitive for artificial radionuclides, 9 give an accurate measurement of inhalation risk, and 9 avoid false alarms due to natural radioactivity. The penetration of inhaled particles in human airways depends on their size. As defined by new standards (European EN 481 and International ISO 7708), the cut-off aerodynamic diameter of the total thoracic fraction is 10 ~tm; it is related to the smallest particles penetrating beyond the larynx. Because these particles are strongly responsible for the inhalation risk, their on-line measurement must be representative. The variations in intensities of deposited fractions as a function of particle diameter is shown in Fig. 9.11. The characteristics of aerosols carrying natural and artificial radionuclides are generally quite different. The size distribution of radon daughters is bi-modal. The median diameter of free atoms is about 10-3 ~tm, while that of attached radon daughters
422
Chapter9
(by brownian diffusion on atmospheric dust) ranges from 0.1 to 0.5 ~tm (AMAD). The AMAD of artificial radionuclides is generally found to be larger than 1 ~tm in facilities of the nuclear fuel cycle. But in the environment, in the case of nuclear accident, the aerosol size varies with the distance from the source (removal process) and with the time (aging process). A few days after the Chernobyl accident, 70% of all the radionuclides sampled in Austria and Germany were concentrated in the size range of 0.1 to 1 ~tm (except for 131i). The activity concentration of radon daughters is much higher than the Derived Air Concentration value of artificial radionuclides. For example, the 239pupublic DAC (2.4 10 -3 Bq m -3) is lower than four orders of magnitude of a mean radon daughters concentration (30 Bq m-3). One of the most impressive programmes of airborne radioactivity is the one implemented by the Environmental Measurement Laboratory (EML) in New York, USA. This started in January of 1963 as a continuation of a programme that was initiated by the U.S. Naval Research Laboratory (NRL) in 1957, and was continued by them until the end of 1962. The primary objective of this program is to study the spatial and temporal distribution of specific natural and anthropogenic radionuclides in the surface air. Surface Air Sampling Program (SASP) provides data that can be used to test the accuracy of model predictions of the trajectories that are followed by natural or artificial aerosols, such as radioactive debris from nuclear weapons tests or other nuclear events, which can serve as tracers for point source injections into the atmosphere. The SASP data also provided some of the earliest evidence on the extent of the large-scale lateral distribution of pollutants injected into the troposphere; for example, the tropical troposphere at mid- or high-latitudes. The occurrence of seasonal cycles of TBe concentrations in the surface air at many sites in the sampling network was also observed from the data. Some of the factors that cause these seasonal variations, such as atmospheric transport and removal processes, were identified by Feely et al. (1989). Larsen (1993) used data from the program to indicate that a global decrease in the production rate of cosmic-ray products, such as 7Be, had accompanied the recent increase in solar activity. The extensive 7Be and 2~~ database continues to provide the scientific community with tracer data which are used to verify global climate model simulations (Brost et al., 1991; Rehfeld and Heimann, 1994). It was suggested by Brost et al. (1991), that the simulation of 7Be and 2z~ might establish the standards for how well a model can represent the concentration and deposition of an aerosol species. As part of this research, EML initiated the simulation of the global distributions of 222Rn and 2~~ using EML's three-dimensional global transport model (Lee et al., 1993a). Comparing model simulations against measurements of 2~~ and other natural tracers will provide useful information necessary to validate and add, remove or modify the existing model modules that describe the various physical processes of the atmosphere. Following the Chernobyl accident, the data were used to characterize the Chernobyl debris which was transported across North America (Feely et al., 1988; Larsen et al., 1986; Larsen and Juzdan, 1986). Larsen et al. (1989) also used the data in a study of the
Monitoring Accidentally Released Radionuclides in the Environment
423
transport processes associated with the initial elevated concentrations of the Chernobyl debris in the surface air in the United States. In 1993, the program was used to detect minute quantities of debris from the April 6th accidental release of radioactivity from the Tomsk-7 nuclear complex in Russia, demonstrating the long-range dispersion of radioactivity in the atmosphere from this accident, and the capability of the SASP network to detect it (Larsen et al., 1994; Lee et al., 1993b). Data from SASP are periodically reported in EML reports (Larsen and Sanderson, 1991, Larsen et al. 1995). The data resulting from this program constitute one of the most extensive and detailed records on atmospheric radioactivity in the world. The resulting data are distributed to scientific organizations throughout the world and have been used by such groups as the United Nations Scientific Committee on the Effects of Atomic Radiation (1993), the United Kingdom's Monitoring and Assessment Research Center (1987), the Max Planck Institute for Chemistry, and more recently the data has been selected for incorporation into the National Implementation Plan for American Participation in the International Arctic Monitoring and Assessment Program. Measurements of the concentrations of specific atmospheric radionuclides in air filter samples collected for the Environmental Measurement Laboratory's Surface Air Sampling Program (SASP) during 1990-1993, with the exception of April 1993, indicate that anthropogenic radionuclides, in both hemispheres, were at or below the lower limits of detection for the sampling and analytical techniques that were used to collect and measure them. The occasional detection of 137Csin some air filter samples may have resulted from resuspension of previously deposited debris. Following the April 6, 1993 accident and release of radionuclides into the atmosphere at a reprocessing plant in the Tomsk-7 military nuclear complex located 16 km north of the Siberian city of Tomsk, Russia, weekly air filter samples from Barrow, Alaska; Thule, Greenland and Moosonee, Canada were selected for special analyses. Traces of radioactive debris from the accident were detected in some of these samples. The naturally occurring radioisotopes that were measured, 7Be and 2~~ continue to be detected in most air filter samples. Variations in the annual mean concentrations of 7Be at many of the sites appear to result primarily from changes in the atmospheric production rate of this cosmogenic radionuclide. Short-term variations in the concentrations of 7Be and 2~~ continued to be observed at many sites at which weekly air filter samples were analyzed. These short-term fluctuations probably resulted from variations in meteorological factors. The data from quality control samples indicate that in general the reliability of the air filter measurements are acceptable for their intended application. In addition, EML has developed the Remote Atmospheric Measurement Programme (RAMP) to measure gamma-ray emitting radionuclides that have been collected by drawing air through highly efficient filters. The gamma-ray spectrum is transmitted to polar orbiting ARGOS satellites, transferred to ground station, and recovered via a telephone link by EML's computer. The recovered NaI gamma-ray spectrum is automatically resolved using a linear least squares program. Several thousand NaI spectra have been received from these remote sites. These spectra
424
Chapter9
provide information on isotopes of current interest and serve as a data base for the future studies of other radionuclides which may become of interest (Sanderson et al., 1994). It is of interest to describe here EML's sampling system, as well as the procedure used in sample collection, processing and analysis. Three air sampling systems are currently used in SASP: a Roots 24-AF or 24-URAI blower connected to a 1 HP electric motor by a fan belt and a Fuji ring compressor directly connected to either a 0.5 or 1 HP electric motor. The Roots sampler accommodates a 20.3 cm diameter filter, which has an effective exposure area (the area of a filter exposed to the airstream) of about 266 cm 2, while the Fuji sampler accommodates a 20.3 cm by 25.4 cm rectangular filter with an effective exposure area of about 407 2 c m . In general, the samples are collected at weekly intervals. The typical range in the flow rates through an air filter using the Fuji and Roots samplers are about 0.8-1.0 (Fuji 0.5 HP), 1.5-1.8 (Fuji 1 HP) and 0.9-1.5 (Roots 1 HP) m 3 min -~, respectively. Microsorban filter material was primarily used in SASP until 1988 (termination of its manufacture). Dynaweb DW7301L filter material (Web Dynamics, Ironia Road, Flanders, NJ 07836) is currently used at all sites in the program. Dynaweb DW7301L is composed of three layers of 100% polypropylene web sandwiched between two sheets of 100% polyester protective scrim. The collection efficiency of this material as a function of particle size and face velocity has been identified (Larsen, 1990). Intercomparison data on the collection of 7Be and 21~ using Microdon and Dynaweb indicate no significant differences in the collection efficiency of these two filter materials. Detailed descriptions of these filter materials, the air samplers and the techniques used to calibrate the samplers and determine air flow rates through the filters are presented in the EML Procedures Manual (1992). At most SASP stations the filters are changed on the 1st, 8th, 15th, and 22nd of the month, or more frequently if the filter becomes clogged. At RAMP stations the filters are changed once a week. The air filter samples that are collected on approximately a weekly basis are referred to as "weekly samples". The weekly samples collected at all SASP sites are composited to form monthly samples. Monthly samples, which consist of weekly samples that represent a minimum of 14 days of exposure during any given month, are referred to as "monthly composite samples". Frequent readings of the pressure drop across the filter or, at RAMP stations, across a fixed orifice, and of the temperature are submitted to EML along with the filters to permit the calculation of the volume of air that was sampled. The filters from most sites are returned to EML for analysis at the end of each month. Because of transportation difficulties, the samples collected at the South Pole Station, Mawson, Marion Island, Palmer and Marsh Antarctica during the winter months are retained at the sites until they can be shipped to EML. This adversely affects the detection and the precision of measurements of short-lived radionuclides in these filters. During 1990-1993, the weekly samples from most of the sites that use a 20.3 cm diameter filter were cut in half; one half of the filter was included in a monthly composite sample, while the other half was archived. The monthly composite samples
Monitoring Accidentally Released Radionuclides in the Environment
425
are compressed into 45-cm 3plastic planchets and are analyzed for gamma-ray-emitting radionuclides using either n-type low energy coaxial, high-purity germanium (HPGe) detectors or p-type coaxial high-resolution, germanium lithium (Ge(Li)) or HPGe detectors. All weekly samples from sites using 20.3 cm by 2.4 cm rectangular filters are treated differently. A section (80.6 cm 2for Microdon, 65.3 cm 2 for Dynaweb) of each of these filters is removed and compressed into a 1.2 cm 3 cylinder, which is analyzed by gamma-ray spectrometry using a HPGe detector with a 1.5 cm diameter well. The remainder of the filter is archived. These filters are not composited into monthly composite samples. The activities of specific isotopes (TBe, 95Zr, 137Cs, 144Ceand 21~ are determined by computer analysis of the spectral data from both monthly composite and weekly samples. The total gamma-ray activity of each monthly composite sample is determined by summing the total counts obtained with germanium detectors between 100 keV and 2.0 MeV, without any correction for detector efficiency or radioactive decay. To monitor the quality of the data from this program, four types of quality control samples (reference, duplicate, replicate, blank) are regularly submitted to the analysts together with routine monthly composite and weekly samples. These quality control samples are submitted "blind" (i.e., in such a way as to be indistinguishable from the routine samples by the analyst) insofar as this is possible. Reference samples, spiked with known amounts of radionuclides, are used to test the accuracy of the gamma-ray spectrometric analysis. For monthly composite samples, weighed portions of reference solutions are added to halves of four clean filters (to duplicate as closely as possible the characteristics of monthly composite samples). For weekly samples, the reference solutions are added to filter sections which are then compressed into 1-2 cm 3 cylinders. If reference solutions are available, 7Be, 95Zr, 137Cs, 144Ce and 21~ are routinely added to the reference samples. The % deviations reported for these quality control samples are influenced by a number of factors besides the accuracy of the gamma-ray spectrometric analysis. Errors in the calibration of the reference solution, weighing errors during the application of the standard solutions to blank filters and a nonhomogeneous distribution of the reference solution in the sample all contribute to the overall reported deviation. Therefore it is believed that these quality control results represent the minimum accuracy obtained in the program. Reference samples with % deviations less than 20% are thus considered acceptable. In addition, the accuracy of the data is considered acceptable if the mean of the % deviations obtained over long-time periods (one year or longer) is less than 10%. Two problems are of importance: radioactive gases and particulates, the airborne particulate monitoring in the critical stage is by air-particulate sampler. Here we shall describe some of them being used at different laboratories. The Aerosol Sampling Station ASS-500 (produced by the Central Laboratory for Radiological-Protection, Warsaw, Poland) is used for routine environmental air monitoring. The station is a stand-alone, all-weather instrument for continuous air aerosol collection. The high air-flow rate (up to 550 m 3h -t) through a chlorinated vinyl
426
Chapter 9
polychloride filter (option polypropylene filter) allows representative sample taking. Collection of aerosols from air volume from 10,000 to 100,000 m 3 enables the performance of accurate spectrometrical measurements of natural and artificial radionuclides in wide range of their concentration, starting from several ~tBq m -3. The station is accommodated for continuous operation in different meteorological conditions. Special systems assure approximately stable air flow through the filter, the G-M counters system (optional) installed above the collection spot allows for on-line air radioactivity control. The main body of the station is made from aluminium sheet. The dimensions are 885x885• 1840 mm. The suction tube penetrates the roof of the station. This tube is also a supporting structure for the filter head. The filter is installed on a supporting mesh in the filter holder. The dimensions of the filter are 440x440 mm. The filter is locked to the mesh by the frame and rubber sealing to avoid by-passing. To dry the filter, two halogen heat-emitters are installed above the filter. The filter, if wet, causes the air flow to decrease. The air flow rate has a range from 90 to 550 m 3 h -~. The air flow is forced by the fan which gives Ap = 4000 Pa. The filtered air is ejected off the station by the 2 m long outlet tube. The aerosol sampling station ASS-500 components are shown in Fig. 9.12. A company in Finland (Senya Oy, Rekitie 7a, 00950 Helsinki, Finland) is manufacturing several types of air samples. JL- 150 The Hunter is a medium volume sampler, 150 m 3 / h . It is meant for continuous sampling and has both a glass fibre filter and an activated carbon cartridge. It is movable; two persons can lift it in a van and choose another place if so desired. GM-tubes can be mounted individually or as a network that can be centrally controlled by one computer. It has the same microprocessor-based panel-meters as with The Dwarf. Pressure difference, air velocity air volume time are parameters one can read out of the panel-meters, one for filter and one for carbon. It can also be equipped with a timer so one can choose sampling times, thus making it possible to lengthen the filter change period. Another possibility is of course to reduce air velocity, but it has been considered that it is better to use sufficient air velocity and have a good sampling result and then have a total break, etc. By using a calibrated pressure difference flange and a pressure difference sender there are no moving parts. A microprocessor-based panel-meter reads the message from PDS during the whole sampling period and counts the total air volume. This means that possible changes in sampling volumes are registered and the reading of total sampled air volume is better compared to the system when only start and stop readings are used. JL-150 Hunter is a modernized version of the older "Little Boy" which has served well and is still going strong after eight years of service. JL-900 Snow White is a high volume air sampler. It is meant to be used when trying to find even the smallest amounts of radiation in the atmosphere. Air volume is calculated from the pressure difference over the calibrated flange, also in the carbon line, which is equipped with a valve. Vacuum is set to 100 mbars which lies in the middle of the pump' s efficiency of 200 mbars and leaves lots of reserve power to hold the air volume quite stable even when the filter gets dirty. The filter is set in a cassette
Zl
..~
i
\
~1
OL~
~ l
[
ii
~0~!
!7
OOEI
,'" t
j
.
.
I........
!
,i
:
I
'.
~t,17
.
. . . . .
.
i[
i
.......
0~8I
4-
.........
-'7
Monitoring Accidentally Released Radionuclides in the Environment
9
i
0
!
!I
I I
+j
.
'
E L
r
,~
<1
tt~
,,.,a
0 . ....,
! r.~
< =
9 0
,,.,a = . ,.,,~
0
.,,,a
E
. ,,..~ ,,,,a
427
t'-i
r ra~
Chapter9
428
which can be loaded in laboratory, thus avoiding problems with rain and wind outside. The filter and carbon holders and frames are made of aluminium anodized after fabrication. The bottom frame is constructed so that the whole unit can be moved on the ground, for instance on grass or on rolling bars. The sampler does not necessarily need a concrete platform--the frame is so stiff that it will even stand on four rocks if necessary. The covers are of GRP, glass-reinforced plastics. They are thick enough to withstand even small flying stones. GM tube(s) can be installed to monitor the glass-fibre filter. They are connected to a network and can be remotely checked by computer operators in the network. Infrared heaters or driers to dry the glass-fibre filter have been tried a couple of times, but do not seem to have much influence and so have not been permanently coupled. GM tube systems measuring the amount of radiation on filters are custom built, and set in three samplers in Finland. As another example let us describe an environmental particulate monitor designed for continuously monitoring radioactive contamination of air, as developed by MGP Instruments Inc., 5000 Highlands Parkway, Suite 150, Smyrna, Georgia 30082, tel. (404) 432 2744, fax (404) 432 9179. The schematics of the instrument installation are shown in Fig. 9.13. The RADAIR instrument develops readouts on 4 continuous measuring channels of activity concentrations of artificial ~, ]3, emitters and natural radon in Bq m -3 and of the ambient 7 dose in pGy h -~. These activity concentration readouts are divided from countings in 1000 s cycles, of the activity deposited on the filter taken from samples of the surrounding air. The filter automatically advances after remaining 24 hours in front of the stack of two semiconductor detectors. The first detector located above the filter, delivers a net counting rate in proportion to the activity deposited on the filter. The
ARGOS ~
POWER SUPPLY ] 220 V. 50 Hz - 1 KW i
3 Ill RADAIR
RS
]TRENCH DEPTH:
20
or
30
cm --
SHEATH
c 9m
I !
"-i
,
Fig. 9.13. Schematic of RADAIR
,
instrument installation.
Monitoring Accidentally Released Radionuclides in the Environment
429
second detector, located above the other, corrects the counting rate of the first detector, from the influence of the ambient y radiation, and in this way can perform an ambient "f compensation. The countings from the second detector, in 1000 s cycles, are converted to a Y dose rate reading display. After separation of the largest particles to an extra-thoracic channel, a good discrimination between artificial and natural radionuclides is achieved by using the following. 9 A silicon detector and the selection of pulses amplitude. Three 222radon daughters can be accurately determined: 21Spo(~), and 214Bi (13) in pseudocoincidence, 9 A membrane filter (AW 19 Millipore) superficially collecting the aerosol particles. A fibrous filter is not used because it involves a variable decrease of the c~ measurement efficiency, and a poor resolution of ot spectrum. 9 A collimating grid placed between the filter and the detector to reduce the widening of ~ spectrum due to a variable length of ~ rays. The RADAIR may be used in a stand-alone configuration or integrated in the environmental control network. The monitor must be installed inside the station (shelter, bungalow, etc.). Through the sampling head, the pump unit feeds air to the detection head which contains a sequentially advanced filter cartridge that captures alpha and beta emitters for detection. The detection type is gross counting for beta, gamma and spectrometry for alpha particulates using dual passivated silicon detectors. The instrument continuously monitors the following: 9 performs the o~ and 13 radon compensation and also dynamic compensation of'/ background through software processing 9 measures and displays: the level of contamination activity caused by ~ and 13 emitting particulates present in the air; the radon natural concentration or background activity; the ambient "/dose rate; and, optionally, the concentration level of noble gas in the air. 9 manages the alarms 9 manages the output data for local display and for communication with a supervisor. Yet another radiation monitoring system developed in Finland is described here: The Alnor (Rados Technology Oy, P.O. Box 506, FIN-20101 Turku, Finland, tel. +358-21-4684 600, fax +358-21-4684 601). The AAM-90 monitors gamma and X-ray radiation in the environment, research establishments and nuclear power stations, wherever isotopes and radiation are used. It allows one to start in a small way and expand as the need arises. The latest technology used together with long-term experience in custom design radiation measurement guarantees a system that keeps pace with needs. Because the RD-02 probe contains two GM-tubes, it measures radiation from a normal background to catastrophically high levels. The operation temperature range is from -40~ to +70~ Enclosure classification is IP67 which means the probe is water-resistant. Power is supplied through the probe cable and auxiliary stand-by
430
Chapter 9
batteries allow long-term continuous operation. Measurements are, therefore, possible under almost any conditions. The RD-02 is a totally independent probe. It does not need control from the area centre or any other device. It measures the dose rate and saves the measurement data in its memory. There is space for 80 measurement results. These results can be sent out through the RS-232 serial connection on request. The excellent linearity in the whole dose rate range is accomplished by the Time Interval Method (TIM) developed by Alnor in the early 1980s. This method calculates dose rate from the time interval measured between pulses, thus cancelling the need to take the dead time of the GM-tubes into account. Calibration factors are saved in the RD-02's EEPROM memory. Probes can easily be removed and sent to a calibration laboratory. Thus, there is no need for on-site calibration. Let us next describe the system which is in operation in Switzerland (Ferreri and Surbeck, 1994) with the aim of monitoring airborne radioactivity. Ten low-volume samplers (approx. 40 m3/h) are placed along the Swiss border, close to the four nuclear power plants and in Fribourg (Fig. 9.14). The 3-in filters are changed weekly and are measured once a month by gamma spectrometry in the lab in Fribourg. The low air volume sampled and the bad counting efficiency for such large filters lead to high 137r~ detection limits. For t_.s,with a collection period of one week and a measuring time of 160,000 s, one typically gets a detection limit of 10 ~tBq/m~. This is largely sufficient to detect intolerable releases from one domestic nuclear plant or from major events abroad like the Chernobyl accident, but cannot be classified as "low-level". By orders of magnitude worse are the detection limits of a new early warning air monitoring system
/ Stein'o.----l~KKL~ /----L_//
Romanshorn
9 S
/
/
Saingnel6gier
.F rib~
L o s ~
rges
rges
Locarno
Fig. 9.14. Air radioactivity monitoring stations in Switzerland (after Ferreri and Surbeck, 1994). Circles: low volume samples; squares: high volume samples.
431
Monitoring Accidentally Released Radionuclides in the Environment
with step filters and simple detectors (RADAIR). The advantage however of this system is that data from the detectors are transmitted continuously to a central computer. The detectors' ability to discriminate between alphas and betas also allows for some compensation of the natural background. They also operate two high-volume samplers with nominally 500 m3/h (ASS-500). One station is north of the Alps close to the lab in Fribourg, the other is south of the Alps between Locarno and Lugano (Fig. 9.14). Because of failures of either the pump or the flow measuring system, there are holes in the accumulated data. The Petrianov-filters are changed weekly. In the lab the glass-fibre part of the filter is separated from the underlying fabric and pressed with 15 tons to a disc having a diameter of 63 mm and a thickness of approx. 4 mm. The disc is measured with an n-type HP-Ge detector (20% rel efficiency) for 160,000 s. Being an n-type, the detector still has a good efficiency down to the 21~ line. The measurement starts at 3-4 days after the filter has been changed. As an example data for 7Be, 2~~ and 137Csare shown as measured for two stations. For the periods without flow measurements only relative activities are shown (Figs. 9.15a-c). The 137Cs activities at Oberschrot, north of the Alps are considerably lower than at Mte. Ceneri, south of the Alps. At first sight this would correspond well to the differences in the deposition of 137Cs during the Chernobyl accident. But a look at the correlations shown in Figs. 9.15a and 9.15b puts some doubt on the view that the 137Cs at Mte. Ceneri is mainly from local resuspension. There is quite a good correlation between changes in the 137Cs concentration and changes in the 7Be concentration (Fig. 20
8000
18 16 6000 ~
14 ~.~
"~ 4000-
10 ~ "7
c...) <
6 2000
0
4
'
i
'
i
",
i
'
i
'
i
0
Jan 1 Feb 1Mar 1 Apr 1 May 1 Jun 1 Jul 1 Aug 1 Sep 1 Oct 1 Nov 1Dec 1 Jan 1
1992 Fig. 9.15a. Radionuclides in aerosol samples, 1992, Station Oberschrot/FR.
~ <
Chapter 9
432
20
8000
18 16 6000
E
r
=I. ,-~
m
14
f
tT'
12 C~ :~L t--.
4000
10 O 8
9
O
<
6
,, [..., O <
2000
o
.~.,
Janl
Fe
. , . , . , 1MarlAprlMaylJunl
. ~ . , . ,
.,~,
Ju 1 A u g l S e p l O c t l
.,. NovlDecl
Janl
1992 Fig. 9.15b. R a d i o n u c l i d e s in aerosol samples, 1992 (until August), Station Monte Ceneri/TI
0.010
1.0'
0.9-
-- 0.009
0.8-
- 0.008 - 0.007
0"71 =I..---,
0.6-J
-0.006
-
0.5-
>. r..)
<
0.005
~ t--! =I.o ~
-0.004
~
- 0.003
~ r..) r..)
0.40.3-
<
-
0.002
-
0.001
0.20.1-
- 0.000
0.0 , , , , ,, ,, ,, Janl FeblMarlAprlMaylJunl
,, ,, 9 ,, Jull AuglSeplOctl
,, , , NovlDecl
1992
Fig. 9.15c. The same as Fig. 9.14b after the month of August.
, Janl
433
Monitoring Accidentally Released Radionuclides in the Environment
9.15b). This points to a common origin. If this origin were the soil surface, the activity ratios in the aerosols should represent--at least to some extent--the activity ratios at the soil surface. From the data on activities in precipitations one can estimate the steady-state surface concentrations. This leads to about 200 Bq/m 2 for 7Be and at least 5000 Bq/m 2 for 2~~ From soil profiles near the Mte. Ceneri station it was estimated that the 137Cs concentration in the topmost part of the soil was around 200 Bq/m 2. Therefore the activity ratios in the aerosol samples do not represent by far the activity ratios in the top layer of the soil: soil surface: 2~~ >> 7Be--137Cs aerosols: 7Be > 2~~ >> 137Cs Generally, twice a year, they get aerosol samples from filters flown by the Swiss Army at altitudes between 5000 and 12,000 m above sea level. Sample preparation and measurement conditions are the same as described above for the high-volume-sampler filters. Figure 9.16 shows that not only the 7Be activity increases sharply across the tropopause (around 11,000 m) but also the Cs activities. The 137Cs/134Cs ratio clearly shows that the Cs in the stratosphere now mainly consists of Cs from the Chernobyl accident. From the data for the filters flown in February 91 and April 93 at altitudes of 5000 m and 7000 m, respectively, one can conclude that there is a considerable downward
~
m
< i 10 5
-
10
-
4
E
i
.,,1
~,
i 0
9 9169
o
r ::::I.
[.., >
10 3
-
102 -
< 10' - 9
10~
-
~7
2600
40'00 60'00
8000
i O(]O0 12000
ALTITUDE / meters above sea level Fig. 9.16. 7Be, 137Csand 134Csat different altitudes in Switzerland. The lowest altitude data are from Mte Ceneri, higher altitude data are from filters flown by the Swiss Army. Error bars represent 2(5 counting errors.
434
Chapter 9
transport of Cs from the stratosphere to the troposphere; at 5000 to 7000 meters above sea level one does not expect a contribution from resuspension. The network of radioactivity measurement stations in Germany is complex, see Fig. 9.17 (after Steinkopff et al., 1995). We shall discuss one location. In the report by Wershofen and Arnold (1995) the activity concentrations of the fission products 134Cs and ~37Cs and some natural radionuclides (TBe, =Na, 4~ and 2~~ contained in Braunschweig' s ground-level air were determined by 3t-ray spectrometry. Mean weekly activity concentrations were measured, mean monthly and mean annual activity ~List
~
o
e
s
t
ort
Fig. 9.17. Field-network of the Deutscher Welterdienst (after Steinkopffet et al., 1994).
Monitoring Accidentally Released Radionuclides in the Environment
435
~. 800 lo' E 7.00
10:
6.00
10:
7Be
-Z
5.00 103 4.00 103 3.00
103
2.00
103 m
[-.
m
l ooo3 k
<
0.0010
~
, ; I I ' ', '. '. : '. I I I I I I I I I : 11 : I : I I I I I I I I I I :
1
2
5
7
9
', I I
' ', '. I '. I ', : I '. , .
11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 WEEKS
Fig.
1994
Mean weekly 7Be activity concentrations in ground-level air in 1994 in Braunschweig.
9.18a.
1.00 104 tzr ::t. Z 9 [., < [., Z r,.) Z 9 o >, [> [.,
21Opb
1 . 0 0 103
1.00
102
1.00 10'
< 1.00
10 ~
1
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
2
5
7
9
11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 WEEKS
Fig.
9.18b.
Mean
weekly
activity concentrations of 4~ and 2'~ Braunschweig.
1994
in ground-level air in 1994 in
concentrations were calculated. The seasonal fluctuations of the mean weekly activity concentrations of 7Be, 4~ 21~ ~34Cs and ~3VCsare given in Figs. 9.18a-d. In 1993 and 1994 no traces of fresh fission products or activation products were detected in ground-level air dust. The mean annual 1 3 7 C s activity concentration due to resuspended soil dust has decreased only slightly compared with the 1992 value. It was less than 0.05% of that in 1986 but still twice as high as in 1985. In 1994 it was less than 0.04% of that in 1986 but still 1.6 times higher than in 1985 (see also Arnold et al., 1994).
Chapter 9
436
3.00 10-'
=I.
2.50 10-' 134Cs
Z
_o
2.00 10-'
1.50 10" Z 0
[-.
1.00 10 -~ 5.00 10 -2
"< LLD Cs-134>
1.00 10 ~
1
2
5
7
9
11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 W E E K S 1994
Fig. 9.18c. Mean weekly '~TCs activity concentrations in ground-level air in 1994 in Braunschweig.
6.00 10 ~ _ m m n
tzr
::I. Z O [.[.. Z t.rq Z O
5.00 10 ~
m --
137C5
m m
4.00 10 ~ m
3.00 10 ~
2.00 10 ~
[..
_> 1.. <
1.00 10 ~
0.00 10 ~
'
1
', ', ', ', ~ ~ ', '
2
5
7
9
~ ~ ~ ~ ', ', ', ~ ' I I I:
', I ' ', ', ', I ', ', ~ '. I ! ! I !.!.
', ', ~ ', , ', ', ', ! I I
11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 W E E K S 1994
Fig. 9.18d. Mean weekly 134Cs activity concentrations in ground-level air in 1994 in Braunschweig.
The activity concentrations of uranium and plutonium isotopes in air-dust ashes in Braunschweig were determined by ix-spectrometric measurements after a radiochemical separation and purification procedure. Mean activity concentrations of 234U, 235U and 238U were measured in monthly samples in 1992 and 1993 and in one semi-annual sample (July-December 1993) and annual mean activity concentrations were calculated. Mean monthly activity concentrations of plutonium isotopes were measured in 1993 and 1994 and the mean annual activity concentrations were calculated.
Monitoring Accidentally Released Radionuclides in the Environment
437
Samples of atmospheric dust were collected about 1.5 m above ground in Braunschweig (52~ 10~ The station in Berlin (52~ 13~ was operated from 1983 to the end of 1992 (Wershofen and Arnold, 1993). Organic fibre filters were exposed on high-volume air samplers with average flow rates of between 390 m3/h and 1100 m3/h. The filter holders are horizontally mounted and covered by a wide weather shield allowing the aerosol fraction of interest to reach the filter. The filters are slightly irradiated by an IR heater to avoid wetting by air humidity. The average filter efficiencies under the operational conditions of the turbine-driven samplers had been previously measured and found to be in the range of 90-95% (by mass) of aerosols. A more detailed investigation of commercially available filter materials on another air-dust sampler which maintains flow rates of more than 850 m3/h over the whole sampling period recently revealed nuclide-specific collection efficiencies of more than 95% for one single filter layer (Arnold et al., 1994). The Finnish centre for radiation and nuclear safety (STUK) has seven sampling stations collecting ground air samples (Leppanen, 1994). Two types of samplers are in routine use: the 900 m3/h sampler (JL900) and the 150 m3/h sampler (JL150). The JL900 type samplers are located in Helsinki, Kotka and Rovaniemi, and the JL 150 type samplers are located in Imatra, Viitasaari, Kurhutunturi and Ivalo. The particle filter in all routinely used samplers is Whatman GF/A glass fibre filter. All samplers are equipped with a 0.5 1 charcoal cartridge in order to collect also the radioactive substances that penetrate the particle filter. One part of the air that has passed through the glass fibre filter goes through activated carbon (type Sutcliffe Speakman 207B K1, mesh 8-12). In carbon sampling, the air flow rate is adjusted to 12 m3/h to obtain retention time of 0.2 s. The JL900 type samplers are equipped with an alarm system with two GM-tubes above the filter. The readings from the GM-tubes are updated in 1 min intervals and an alarm level of 0.5 ~tSv/h for a 15 min average reading is in use. For preparedness purposes there are in addition three samplers of type JL 150 in reserve. Two of them are located in Helsinki and one is located in Rovaniemi. If needed, they can readily be brought into use and operated anywhere in Finland with the standard 230V (1 kW) power supply. One of the reserve samplers is equipped with a timer. At present STUK has a third type of sampler, type JL40, in use. The JL40 type sampler is portable, and can also be operated in any place with the standard 230 V (1 kW) power supply. The JL900 and the JL150 samplers are developed in STUK and the JL40 sampler is developed by Senya OY. More detailed technical data and pictures of the samplers are available from manufacturers. In the French laboratory in Monthl6ry jointly operated by Minist~re de la DEfense (Ministry of Defence) and Commissariat ?a l'Energie Atomique/C.E.A., under the authority of Direction des Centres d'ExpErimentations Nucl6aires/DIR.C.E.N. (Nuclear Experimentation Centre Directorate), radioactive aerosols particles are collected in surface air, using cellulose filters and high output pumps (110 m3ha). Equipment operates on a permanent, 24-hour a day basis, and filters are replaced daily
Chapter 9
438
throughout the year. The volume of filtered air is 2600 m 3 per day. After removal, individual filters are subjected to alpha counting, using a ZnS counter, and to beta counting, using a proportional counter. Counting takes place five days after collection, to allow sufficient decay of natural background radioactivity. Later, filters collected at one location during any one-month period (representing a filtered air volume of 80,000 m 3) are assembled, calcined and dissolved in acid for the purpose of measuring gamma emitter activity by gamma spectrometry using a HP Ge detector. Chemical extraction is then performed to isolate Plutonium isotopes (see Fig. 9.19), which are subjected to alpha spectrometry using a grid ionization chamber following electro-deposition. To improve I34c9 and 239+24~ detection and quantity determination, certain samples are [ Atmospheric sampling filter [ I Alpha and Beta counts at D+5
i
,=
I Collection of filters for the month 1 ! Calcination at 550 ~ C for 36 hours !
[
|
Addition of 236putracer Dissolving ashes in fuming HNO3, then in HNO3+HF+HC10~ Recovery of diluted HC1 Gamma spectrometry using HP Ge detector (4000 mn of measurement) Evaporation. Recovery of HCI 6N Elimination of iron by 4-metylpentan-2-one extraction Evaporation. Recovery of concentrated HNO3. HNO3+H202,then HC1 1ON 2 cycles Fixation on AG lx8 resin in HCI 10N HNO38N flushing Elution of Pu in HC1 1N+Hydroxylammonium chloride Electrodeposition on stainless steel disc (NH4CI electrolyte) Alpha spectrometry in grid chamber (4000 mn of measurements) Fig. 9.19. Radiochemical process sequence for separating plutonium collected in atmospheric filters (after Milles-Lacroix et al., 1994).
439
Monitoring Accidentally Released Radionuclides in the Environment
assembled on a six-month (filtered air volume of 480,000 m 3 approx.) or yearly basis (in the latter case, filtered air volume is approximately 960,000 m3). Gamma spectrometry measurements are then made in Modane, and alpha spectrometry measurements in MonthlEry. It was thus possible to determine 134Cs and 239+24~ quantities (MilliesLacroix et al., 1994). The ground level air concentrations of lead-210 have been measured at numerous locations all over the world (Rangarajan et al., 1976). The vertical distribution of this nuclide in the atmosphere was determined by Burton and Steward (1960), Rama and Honda (1961), Feely et al. (1965) and Peirson et al. (1966). The results of these measurements were used for the study of the air mass transport and the residence time of aerosols in the atmosphere (Machta, 1965; Karol, 1970; Moore et al., 1973, 1980; Martell and Moore, 1974; Rangarajan et al., 1975). Air concentrations of radium-226, lead-210 and uranium near ground level at different locations are shown in Table 9.9. There were only a few measurements of radium-226 and uranium in the ground level air. Preliminary data on the vertical distribution of radium-226 and uranium in the troposphere and stratosphere measured during periods of 3-11 years have been reported by Jaworowski and Kownacka (1976), Kownacka (1980) and Kownacka et al. (1985), and measurements of uranium in six samples of high-altitude aerosols have been reported by Krey et al. (1979). In the paper by Kownacka et al. (1990) results of the long-term measurements of radium-226, lead-210, uranium and stable lead at several altitudes in the troposphere Table 9.9 Some of the reported air concentrations of radium-226, lead-210 and uranium near ground level (after Kownecka et al., 1990) Radium-226 (~tBq m-3)
Brunswick (Germany) Berlin (Germany) Boulder (Colorado) New York City (New York) Teheran (Iran) Gorce National Park (Poland) Chorzow (Poland) Warsaw (Poland)
1.8 1.2 0.74 0.003 10-840 0.7-2 4--45 0.8-32.2 (av. 12.4)
Lead-210 (~Bq m-3)
Northern Hemisphere Warsaw (Poland)
0.01-1 <0.04-0.71 (av. 0.32)
Uranium (ng m -3)
Sutton (Gt. Britain) Atlantic Antarctica New York City (New York) Mol (Belgium) New York State Tokyo and Tsukuba (Japan) Warsaw (Poland)
0.01-0.4 0.002-0.004 0.002-0.004 0.4 0.04-0.2 0.1-1.5 0.005-0.065 0.08-1.44 (av. 0.43)
Chapter 9
440
km 15
12
200 I 20
10 I
I,
0.2
I
0.4
....I
0.6
I 0.8
Pb / nq / m ~ STP ~26Ra / l.tBq / m ~ STP " 21opb
U / nq / m-' STP .. / mBq / m 3 S T P "
Fig. 9.20. Vertical distribution in the a t m o s p h e r e of average concentrations of radium-226, lead-210, u r a n i u m and stable lead (after K o w n a c k a et al., 1990).
over Poland have been presented. Vertical distributions of radium-226, lead-210, and uranium were observed in the troposphere and lower stratosphere over Poland at several altitudes between 0 and 15 km in the period 1973-1987. The reported results are shown in Figs. 9.20, 9.21, and 9.22. Greatly increased concentration of radium-226 was observed at all altitudes for several years after the Fuego volcano eruption in 1974, and also after the Nevado del Ruiz eruption in 1985. The volcano eruptions in 1980--1982 contributed to the radium-226 and uranium levels at the higher altitudes. The annual flows of radium-226, lead-210, and uranium into the global atmosphere, estimated from their long-term average contents in the 0-15 km air layer, are 2.3x10 TM Bq, 8.4x10 Is Bq, 8 . 3 X 1 0 9 g respectively. These estimates are similar to those based on concentrations of these nuclides in widely dispersed glaciers in both hemispheres, and on radon-222 exhalation measurements. However, they are higher than estimates based on particulate emissions. The anthropogenic contribution to the total flow of radium-226 into the global atmosphere is-3.7%, for lead-210 0.25%, uranium 17%. A summary of airborne radioactivity measurements in Monthl&y, France is presented in the report by Milles-Lacroix et al. (1994). Radioactive aerosols particles are collected in surface air, using cellulose filters and high output pumps (110 m3/h). Equipment operates on a permanent, 24-hour a day basis, and filters are replaced daily throughout the year. The volume of filtered air is 2600 m 3 per day. After removal, individual filters are subjected to alpha counting, using a ZnS counter, and to beta
Monitoring Accidentally Released Radionuclides in the Environment
441
mBq
100
80
60
40
20
I
i
! "!
L
' I
19,75 Fuego
I
I
1980S Mt. St. Helens
I
I
I'
I
i
s E1 Chicon
I
I
I
I
1985 llt Nevado DelRuiz
I
I
|
I
e i
i i
li~ pr
1990 I Mt. Pinatubo
Fig. 9.21. Temporal changes of content of radium-226 in the 1 m 2 column of air between ground level and 15 km (after Kownacka et al., 1990).
106"o
~
10 5
I !
CHINESE
! I I !
10 4
l
~
CHERNOBYL ACCIDENT
t ! I I I t
~
STRATOSPHERIC AIR
103! I
102:
\
10 ~, GROUND LEVEL AIR
10 ~
10-'
,
'
1975
.
.
.
.
1980'
'
'
'
1985'
'
'
'
1990
.
.
.
.
' ......
Fig. 9.22. Average monthly concentrations of radiocaesium in the ground level air and daily concentrations at 15 km altitude (after Kownacka et al., 1990).
442
Chapter 9
counting, using a proportional counter. Counting takes place five days after collection, to allow sufficient decay of natural background radioactivity. Later, filters collected at one location during any one-month period (representing a filtered air volume of 80,000 m 3) are assembled, calcined and dissolved in acid for the purpose of measuring gamma emitter activity by gamma spectrometry using a HP Ge detector. Chemical extraction is then performed to isolate plutonium isotopes, which are subjected to alpha spectrometry using a grid ionization chamber following electro-deposition. To improve ~34Cs and 239+24~ detection and quantity determination, certain samples are assembled on a six-month (filtered air volume of 480,000 m 3 approx.) or yearly basis (in the latter case, filtered air volume is approximately 960,000 m3). Gamma spectrometry measurements are then made in Modane, and alpha spectrometry measurements in Monthl6ry. It was thus possible to determine 134Cs and 239+24~ quantities. The details of the radiochemical process are given in the flowchart showm in Fig. 9.19.
9.6.2 Radon gas measurement methods
All Rn gas measurement methods make use of a sensitive volume into which Rn gas is introduced. The gas may be introduced by active means, such as pumping, or the gas may diffuse into the sensitive volume. In the latter case the methods are said to be passive. A detecting element is mounted inside the sensitive volume, which detects the radiation from the radon in the sensitive volume directly, or absorbs it for later detection of the radiation. The detection means is the main distinguishing factor between the various Rn gas measurement methods. The measurements may be in the form of grab samples of the gas, i.e. the concentration is sampled at a specific instant in time. Alternatively the sampling may be continuous, and a time spectrum of the radon concentration is available. Lastly, the sampling may result in an integration of the concentration over a certain time period. The latter method is the most common, and is also preferred because it provides the best way to obtain average radon levels at the typically low concentrations found in environmental and indoor situations. It also incorporates the temporal fluctuations that occur in the radon concentrations (see ICRP65 and ICRP66). A typical scintillation cell is shown schematically in Fig. 9.23. It consists of a hemispherical or fight circular cylinder (George, 1976; Lucas, 1957). The walls of the chamber are optically clear, and are coated with ZnS (Ag) scintillator powder. Air is introduced into the chamber passively (after evacuation of the chamber with a pump, and opening to atmosphere) or by means of a pump. Inside the chamber, the Rn decays into the RnD. The alpha particles from the decay causes the scintillation powder on the walls to scintillate. Four hours after sampling, the radon is in equilibrium with its daughters. The container is then placed on a photomultiplier tube and the scintillation pulses (light pulses) from the ZnS layer are counted with associated electronics. With calibration, the number of light pulses can be related to the concentration of Rn in the sampled air.
443
Monitoring Accidentally Released Radionuclides in the Environment
Valve
~
9 Radon decay K/Alpha Radiation
",,.. \"..... ' .....Light "-":.,,....,
n
entry
~l--ZnS (Ag) coated walls
...~ PHOTOMULTIPLIER
Fig. 9.23. Scintillation cell.
The passive means of air introduction into the cell leads to a grab sample of the Rn concentration, similarly for a short sample period by pumping. To obtain a sample over a longer time, the pump can be run for a longer period. However, the RnD attaches to the walls of the chamber and the counts obtained in any time interval during or after sampling, is a complex function of the growth and decay of the various isotopes of Rn. In principle, this may be solved by complex mathematical software, to provide continuous readings of the radon concentration. This way of monitoring is however, tedious, and open to many sources of error, e.g. the pump flow speed. The method of scintillation cells is capable of measurement of Rn concentration down to very low levels, however, it becomes necessary to use cells of very large volume to overcome the problems of counting statistics that occur at the low concentration levels. For long-term monitoring, scintillation cells are not considered convenient. A disadvantage of the scintillation cells are that sophisticated, and expensive, electronic counting equipment is required. This is often not mobile which severely restricts the measurement regimes of this kind of monitor. It is known that about 88% of all 2~8po atoms formed carry a single, positive, electric charge (Wellisch, 1913; Porstendorfer and Mercer, 1979; Dua et al., 1983; Chu and Hopke, 1985). The positive electric charge on the decay product provides a means of collecting the atoms, after formation, on a sensitive surface where the subsequent radioactive decay can be recorded. The collection is done with a relatively high electrostatic field. A number of such electrostatic monitors have been developed, and are available commercially (Albrecht and Paul, 1967; Wrenn et al., 1975; Spitz and Wrenn, 1977). A typical electrostatic collection monitor is shown schematically in Fig. 9.24. The air is introduced into a chamber, with conducting walls at ground potential. In the chamber, there is a sensitive surface, at a high (1-2 kV) negative potential. The 218po atoms are collected on the sensitive surface by the electric field. The sensitive surface can be any of a number of radiation-sensitive detectors, e.g. the surface may be a layer
444
Chapter 9
Rn
Pump E Field .................. ........ 9.... .~-
RaA " ~
.........................
.."'" ..." .....:" -.
:.:" ,
................................"'"'.. '"
f f/
..y........".........................
Detector Fig. 9.24. Electrostatic collection.
of ZnS (Ag) scintillation powder, covered by a very thin conducting material (aluminised Mylar). The layer is mounted on a photomultiplier, which then records the light pulses from the decay of the RnD on the screen. Alternatively, the sensitive surface may be the surface of a silicon barrier detector (e.g. PIPS), connected to its associated electronics. These monitoring methods are capable of continuous monitoring of the radon concentration at low levels. It has been shown, however, that the humidity content of the air being sampled has an effect on the collection efficiency of the monitor (George, 1976; Hopke, 1989). This is explained by the fact that the ionising radiation of the Rn and its RnD in the chamber dissociates the HzO vapour molecules in the air into species which can neutralise the positively charged :~8po atoms. Less atoms are then collected per unit of Rn activity. Monitors must therefore have a means of drying the air, or it must be ensured that the collection electric field is high enough to collect the atoms before they have time to be neutralised. Many commercially available monitors do not dry air, and the collection fields are insufficient to account for these effects, and these instruments must be used with caution. Air can be adequately dried using CaSO4 (Drierite). Other drying agents may absorb the Rn atoms. It should also be noted that the counts obtained from the detector in these instruments are a complex function of the growth and decay of the various Rn isotopes on the detector. To obtain a true time spectrum of the Rn concentration, especially on a scale of hourly fluctuations, these counts have to be corrected using deconvolution algorithms. The same disadvantage of associated electronic equipment is valid for the electrostatic means, although commercial instruments contain these as part of a compact measurement unit. Another means of determining the Rn concentration in the sensitive volume is by collecting or absorbing the Rn atoms in a medium. This can most conveniently be done
Monitoring Accidentally Released Radionuclides in the Environment
445
Fig. 9.25. Charcoal Rn monitoring.
using activated charcoal granules as shown in Fig. 9.25 (Countess, 1970). The sensitive volume is filled with activated charcoal, and radon diffuses into the vessel (or pumped in). The radon atoms are adsorbed into the charcoal, and decay to the various RnD in the charcoal. After exposure, the container with charcoal is left for a period of 4 h for the RnD in the charcoal to reach radioactive equilibrium with Rn and then put onto a radiation detector such as NaI crystal with associated electronics. The y ray activity in the charcoal (from the decay of the RnD) is recorded. With calibration, the count obtained can be related to the Rn concentration in the air. Charcoal canisters can be exposed for periods up to about seven days, and therefore provide a means for long-term monitoring of Rn. However, the charcoal also absorbs water vapour, which makes it less effective toward the adsorption of Rn. Standard methods and procedures exist for the use of activated charcoal canisters (George, 1984), which have been adopted by the US EPA for indoor radon monitoring. The monitor averages the radon concentration over the period of exposure. However, it has been shown that the radon gas can also degas from the charcoal. This implies that the canisters become very sensitive to the time variations in the Rn concentration, and tend to record the changes toward the end of the exposure more realistically than at other instances during exposure. To reduce the effects of the variations by damping, the entrance of the canisters can be covered with a diffusion barrier (smaller open area), or semipermeable membrane (Pritchard and Marien, 1985). Cohen and Nason, (1986) developed a barrier charcoal canister for a seven-day exposure period. It appears that with the proper case and calibration, and correction for humidity effects, charcoal canisters can be used for reliable radon monitoring for periods of 2-10 days. However, this period may be too short to provide the long-term annual average exposure value that people are exposed to. Another disadvantage is again that electronic counting equipment is required for the analysis of the exposed charcoal. A solid-state nuclear track detector is a piece of special plastic which is exposed as the sensitive element in a radon monitor. The alpha radiation from Rn and RnD, which penetrates the surface of the plastic, causes radiation damage along the entrance path, as shown in the schematic in Fig. 9.26. Chemical etching of the plastic after exposure
446
Chapter 9
//"
Alpha radiation
//" ,,~
,," o,O
Chemical etching
Y
Damage
o 50-80 ~tm o
V
O
o
etch pit Top view Fig. 9.26. Track etching mechanisms.
makes these damage trails visible to the eye, and the track density can be determined by counting the tracks. Two types of track-etch monitor occur, open and closed types. In the open type, the SSNTD is not contained in a volume and is exposed to the air as a bare foil. This detector will register the alpha radiation from the Rn and RnD in the air, and the track density on the foil represents the sum of these activities. However, the Rn signal will be much larger than the signal from the RnD, except at very high levels of RnD (high F factor), and the track density has to be interpreted in terms of this ratio, which is typically unknown. In close monitors the SSNTD is enclosed in a closed container into which Rn diffuses through a filter. This prevents the entry of RnD and dust particles into the chamber, and the foil is then sensitive only to the alpha radiation from Rn and RnD formed in the container. There is a repeatable equilibrium between the isotopes in the container, and calibration provides the relationship between the Rn concentration and the track density on the foil. A typical track-etch radon monitor of the closed type is shown in Fig. 9.27. ]~ '
Filter Radon decay
Fig. 9.27. Closed track-etch monitor.
Monitoring Accidentally Released Radionuclides in the Environment
447
There are various kinds of plastic material suitable for use as track-etch detector, of which LR 115 (cellulose nitrate), MAKROFOL (polycarbonate) and CR-39 (allyldiglycol poly carbonate) are the most popular. These materials are described by Fantini and Richard (1981) and Cartwright et al. (1978). LR-115 is produced by Kodak-PathE in France, MAKROFOL by Bayer and CR-39 by a number of manufacturers, viz. American Acrylics, Pershore Mouldings, UK. The materials differ in sensitivity and etching method, as well as in the way the tracks on the foil are read. The materials also differ in robustness and ease of handling. Two etching techniques exist, chemical and electrochemical etching. In chemical etching, the foil is immersed in an aqueous solution of NaOH or KOH (1-12 M) in temperatures that range from 40-80~ The time of immersion varies between 6 and 16 h. The damage trails become visible as spots of 50-80 ~tm. In electrochemical etching an alternating electric field at frequency of about 2 kHz is applied across the foil while in the etching solution. This causes electrical breakdown at the track tips, within large star-shaped tracks or spots on the foil, with diameters of 100-150 ~tm, which are more easily countable. Track counting is realized by various means, ranging from eye counting under a microscope or microfiche machine, to sophisticated automated equipment with video camera, computer etc. When small numbers of foils are being analyzed, manual counting on a microfiche machine is the least expensive option, although tighter quality control measures must be in place for the counting operators. The entire system of preparation, exposure and analysis of the track-etch monitors should actually be very strictly quality controlled through frequent calibration, control foils, standard check counts, etc., since error may easily be introduced which is not easy to detect. A track-etch monitoring service is available commercially in South Africa, which supplies ready-to-measure monitors, plus the full analysis and quality control service. These track-etch systems are relatively unaffected by environmental factors such as humidity and atmospheric pressure changes, and will respond only to the radiation from Rn. The monitors will not record T radiation from external sources. Track-etch devices are considered to yield the best estimate of the annual average exposure of individuals, due to the very long integration times possible with these monitors, and is therefore the instrument of choice for long-term, low-level surveys. The measurement times vary according to the Rn concentration, but are typically one month at high underground levels (100 Bq m -3) to three or more months at environmental levels (10-50 Bq m-3). The commercial system in South Africa has a lower detection limit of 11 Bq m -3 for an exposure time of two months, and about half of that at exposures exceeding three months. A relatively recent development in radon gas monitoring is the electret system (Kotrappa et al, 1988). In effect this is an ionization chamber with an electret foil in the chamber, into which radon gas diffuses through a filter. The electret is a plastic foil which carries a permanent electric charge and therefore an electric potential. The decay of Rn and RnD in the chamber ionises the air, and the electric field caused by the potential on the electret attracts these ions to the electret. The electret loses its charge proportionally to the number of ions collected. The charge, or voltage on the electret is
448
Chapter 9
measured with a custom-made voltage reader, and the radon concentration is calculated by special software from the decrease in potential of the electret. The electret ion chamber for radon measurement or E-PERM (commercial name) can be used to determine the long-term average Rn concentration over periods of one week to months, depending on the levels. The electret system, as any other ionization chamber, is sensitive to pressure variations, and corrections for this effect should be made. A major disadvantage is that the electret will also respond to radiation from external sources. When used in gamma radiation fields (in many situations) a duplicate set of monitors have to be exposed to be able to correct for the effect of the external gamma radiation on the radon measurement. Radon exhalation is the flux of Rn from surfaces of U-containing materials. In many instances, it is necessary to measure this flux to determine the source term of radon represented by the material (e.g. radon exhaled from tailings dams as a radiological impact on the public; or from building materials used for houses, to determine the potential radiation hazard). There are various ways to measure this parameter. In this method a container is upturned onto the surface of the material, and the radon that accumulates inside is measured by any of the methods for radon measurement described in the previous section. The actual means of Rn measurement distinguishes the different available methods. There are, however, two very basic objections to the use of this method. The first pertains to the fact that the measurement is usually done over a relatively short time period (24 h to 1 week). The exhalation rate may vary considerably in this time, due to atmospheric pressure variations, and the monitoring method should adequately, and in a known way, integrate these variations. However, it is possible that the measurement is made when the exhalation is at a peak or valley of the exhalation rate variation, which means that the measurement will either overestimate or underestimate the true annual average exhalation rate. The second objection has been raised by Samuelsson, 1987. When the container is placed onto the material surface, the concentration gradient in the underlying material is immediately disturbed. This affects the exhalation rate, which immediately changes, and the monitoring system measures an erroneous value. The extent of perturbation depends on the type of material, among other factors. The use of methods with containers sealed to the surface of the material should therefore be approached with extreme caution, and the system should be analyzed to quantify the above effects. The exhalation of Rn from material surfaces is controlled by the generation rate of Rn in the material, and the transport by diffusion through the material to the surface. The generation rate is determined by the 226Rn content of the material, and the emanation fraction. The transport through the material is controlled by the diffusion length through the material. The diffusion process is well described mathematically by one-dimensional diffusion theory, so that knowledge of these parameters will allow accurate calculation of the Rn exhalation rate from the material surface. The generation and transport parameters can be measured in diffusion tubes; tubes into which the material is compacted to various depths. The Rn emanating from the
Monitoring Accidentally Released Radionuclides in the Environment
449
material into the headspace of these tubes is measured using track-etch methods. From the measured values, the required parameters are then deduced through mathematical treatment.
9.7 MONITORING FOOD The determination of the radionuclide concentration in the diet or individual food items constitutes an important element of an integrated program of radiological surveillance and assessment. This field has been very active after the Chernobyl accident. Many countries' food importers have applied their own regulations. For example, in Japan the Ministry of Health and Welfare, responsible for food safety, determined The Interim Derived Intervention Limit for screening imported foods. In determining the permissible limit of radioactivity in foods, the Ministry adopted not more than 1/3 of the 500 mrem/year dose (individual dose-equivalent limit to the member of the public) under the Radiological Protection Law of Japan (in 1986) on the basis of recommendation by the International Commission on Radiological Protection (ICRP Publ. 6). From the following formula, one can deduce the permissible limit, A pCi per one kilogram of imported foods. 5.4x10 -5 (mrem/pCi) x 1.4 (kg) x 35 (%) x A (pCi/kg) x 365 (days) < 500 (mrem) x 1/3x66 (%) A < 421 (Bq/kg) where 5.4x10 -5 (mrem/pCi) is the coefficient for the whole body exposure by Cs calculated from fallout 134Cs and ~37Cs concentration ratio of 1:2; 1.4 (kg) is the per capita per daily food intake; 35 (%) is the ratio of imported food to food intake; A (pCi/kg) is the radioactivity concentration in imported foods; 500 (mrem) is whole body dose limit; 1/3 is the allowance for exposure from food of whole body dose limit; and 66 (%) is the dose contribution from 134Cs and 137Cs in food. The formula was derived from the fact that the target is imported food and the major radioactive fallouts are 134Cs and 137Cs. The ministry took account of the level of the European Community (370 Bq/kg for milk and infant food and 600 Bq/kg for general food) and that of the USA (10.000 pCi/kg, 370 Bq/kg) and set up the Japanese interim standard level of 370 Bq/kg) If radioactive contamination of imported foods was over the standard level, the foods were requested to be reshipped as products in violation of the Food Sanitation Law of Japan. Although the level was reviewed afterwards, the level was still suitable for products imported from European countries, even after amendment of the law of Japan in which the annual whole body dose limit of 500 mrem was altered to 100 mrem (= 1 mSv) in 1988 (on the basis of the recommendation of ICRP Publ. 26, in 1985), because the real foods imported from European countries were estimated to account for 5% of all imported foods which were 35% of all food intake in Japan.
450
Chapter 9
Table 9.10 Guideline levels for radionuclides in food
Foods destined for general consumption:
Dose per unit intake factor ( S v / B q )
Representative radionuclides
Level (Bq/kg)
10.6
241Am, 239pu
10
10 -7
9~ 131I, 134Cs, 137Cs 241Am,239pu
1000
10-8 Milk and infant foods:
10-5 10-7 10.8
1311,9~ 134Cs, 137Cs
100
1 100 1000
The governmental inspection of imported foods was taken by food sanitation inspectors at seaports and airports in Japan by a first screening test with a simple scintillation survey meter and also by a minute analysis of Ge detector at four major sea ports (Tokyo, Yokohama, Osaka, Kobe), one major airport (Narita) and in the National Institute of Hygienic Sciences, Tokyo. At the same time, trade companies were requested voluntarily to check imported foods for radioactive contamination under the above Japanese interim standard of 370 Bq/kg at the laboratories designated by the Ministry of Health and Welfare of Japan (Washima and Ohkubo, 1987). At the beginning of the survey, the 100% inspection was conducted for foods imported mainly from USSR, Eastern and Northern European countries and the 10% inspection for ones from the other European countries. Afterwards, the target countries and food items were reviewed to check samples effectively on the basis of information from foreign countries and the results of data which had already been checked. It was the Joint FAO/WHO Food Standards Programme and its Codex Alimentaries Commission which published guideline levels for radionuclides in food, following accidental nuclear contamination, for use in international trade. These levels are shown in Table 9.10. Food is analysed to determine: (1) the level of contamination at the point of production; and (2) the level of intake of the contaminant for the consumer or a particular population group. Several factors must be considered in determining the sampling protocol for diet intake studies. These include the number of samples that can be handled, the effect of the lag time between production and consumption on the level of the contaminant, the importance of the relative contribution of a single food or food group, and the effect of processing of the food for consumption on the level of the contaminant. There is no single method that is entirely satisfactory for estimating human dietary intake of contaminants in foods. Several approaches will be discussed in the following section. More detailed information can be found in the guidelines by the World Health Organisation (1986). One method of food monitoring is to sample individual foods at the point of production. This is most useful for relating contaminations to local conditions of
Monitoring Accidentally Released Radionuclides in the Environment
451
fallout, soil content, or farming practice. The geographic area to be sampled is generally relatively small. This sampling system is used, for example, by the Food and Drug Administration (FDA) in their program of food monitoring near power reactors (Stroube and Jelinek, 1985). It is very difficult however to correlate the concentration level of a contaminant in samples collected at the point of production to the dietary intake of any specific group of people. Another sampling procedure often used is to duplicate the actual foods consumed by an individual during a day, a week or some specified time period ("duplicate plate" sampling). This type of sampling is used in metabolic balance studies where the intake and the retention of the contaminant are monitored on an individual basis (e.g., Spencer et al., 1973). This procedure has also been used in the U.S. Environmental Protection Agency's Institutional Diet Sampling Program (USEPA, 1974) and the California (Hospital Standard) Diet Study (USEPA, 1973). The advantage of collecting meals over a week period, and duplicating meals as consumed is that the resulting sample is a better estimate of the average dietary intake. However, since the analyses are performed on the composite of the diet, only the total intake can be estimated. Information on the contribution of the component parts of the diet is not obtained. A third diet sampling procedure is based on analysis of representative food items purchased locally and on food consumption statistics. The intake estimates are then calculated by multiplying the concentration estimate of each food by the respective consumption estimates per day or per year. This approach is used by EML in 9~ in the diet program (Klusek, 1984), Argonne's ~37Cs in the Chicago Foods Program (Karttunen, 1982), and the FDA's Total Diet Studies Program (Stroube and Jelinek, 1985). This sampling method allows a considerable degree of flexibility for consideration of special diet types or differences in consumption by region, age, sex, income or urbanisation. It is possible to reduce the number of foods or to consider more general categories of foods to reduce cost or effort in particular cases. It is recommended that the sampling take place at the retail level; however, if appropriate (i.e., rural farm intake) the sampling can take place at the consumption level. The advantage of this method is that the analysis of a number of foods or food groups gives more information than the analysis of a single composite diet sample. It is also useful in indicating the primary sources of intake for a particular contaminant. However, the intake estimates are limited by the accuracy of the consumption statistics. A summary of estimates is presented in Kramer et al. (1973). Food consumption statistics are available through national food sheets (e.g. FAO, 1984) or local specialists. Food consumption statistics for U.S. population groups have been published by the U.S. Department of Agriculture (USDA) in a series of surveys conducted since 1936. The "Household Food Consumption Survey" presents data on the consumption of 200 food types by income group, by region in the U.S. and by degree of organisation (USDA, 1982). Food intake summaries for 22 sex-age categories, by racial group and by season are also available through the USDA (1983). In cases where it is necessary to compile other surveys of consumptions statistics,
Chapter 9
452
several reviews of the methods of measuring dietary intake are available (Block, 1982; Marr, 1971; Morgan et al., 1987). The EML (1990) diet sampling program was developed to provide estimates of the intake of 9~ by man. These dietary intake estimates are then correlated with fallout measurements and the 9~ c o n t e n t of human bone. The sampling protocol or archived samples from the diet sampling program have been used by other researchers in studies of the dietary intake of other natural (Fisenne and Keller, 1970; Fisenne et al., 1987; Petrow et al., 1965; Welford and Baird, 1967) and artificial radionuclides (Bennett, 1976; Rivera, 1967) and some stable elements (Bogen, 1972; Rivera, 1962). At EML, the diet sampling is grouped into 19 food categories representing the five basic food groups. On a quarterly basis, 42 food items are purchased in New York City markets. The foods purchased within each category approximate the distribution of food consumption within that category. The food items sampled are shown in Table 9.11. Table 9.11 EML diet sampling program food purchase list Food group
Food category
Food item purchased
Dairy products
Milk
Whole milk
Vegetables
Fresh vegetables Canned vegetables Root vegetables Potatoes Dry beans
Lettuce, cabbage, spinach, peas, beans, tomatoes
Fresh fruit
Oranges, apples, bananas, melons, berries Peaches, pears, pineapple, applesauce Orange, pineapple, tomato, grapefruit
Fruit
Canned fruit
Grain products
Fruit juice Bakery products Flour Whole grain products Macaroni Rice
Meat, eggs and fish
Meat Poultry Eggs Fresh fish Shell fish
Peas, beans tomatoes Onions, carrots, turnips Whole potatoes Navy beans
White bread White flour Whole wheat bread Spaghetti, elbow macaroni White rice Pork, beef Chicken legs, chicken breasts Fresh white eggs Halibut, fillet Shrimp, clams or oysters
Monitoring Accidentally Released Radionuclides in the Environment
453
9.8 M O B I L E R A D I O L O G I C A L UNIT A significant number of mobile radiological units are in operation worldwide aiming to provide reliable radiological data. They have mainly been designed and constructed on a national basis according to the particular needs and commitments of the specific laboratory or country. In most cases, these units are intended to be used in emergency situations for in-situ radiological measurements of accidentally released radioactivity, and sometimes for monitoring environmental pollution. As the purpose of these units is very diversified with regard to the kind of vehicle and its in-built measuring equipment, the varying outfit of these units cannot be adopted in general for other countries aiming to improve their capability for in-situ radiological measurement. In order to achieve harmonisation of equipment and comparability of radiological data being obtained from field measurements, it is necessary to have general guidelines available for designing mobile radiological units, taking into account different scenarios and tasks to be achieved. In the very early stages of an accident, most of the information available on the quantity of radioactive material being released, its radionuclide composition and the likely progression of the accident will come from the operator, and will be based on the conditions in the plant. Few environmental monitoring results from off-side can be expected within the first few hours. In this very early phase, decisions on the application of protective measures will, therefore, be based largely on plant status and forecasts of changes in that status as well as on meteorological data. As time progresses, results will increasingly become available from the monitoring of radionuclides in the environment (e.g. dose rates and concentration of radionuclides in air and particular materials such as water, food etc.). Monitoring results can be used to estimate potential doses to people and the need for further protective measures can thus be determined from a comparison with intervention level of dose. Decision making in an emergency will however be more rapid and effective if the intervention levels of dose are expressed in terms of level of radionuclides present in appropriate environmental materials. The latter are termed "derived intervention levels", (DIL) and are the practical expression of the intervention level of dose. Contamination of an environmental material at the derived level is predicted to result in an exposure at the intervention level of dose. The need for, and extent of, protective measures can be determined by direct comparison of the monitoring results with the derived levels. When designing a mobile radiological unit one has to take into account different scenarios and tasks. Options for the design of mobile radiological units include: satellite, helicopter, boat, train, truck, car, robots, portable devices. Guidelines for a mobile radiological unit should include recommendations on kind of truck/car and its design for either emergency situations and/or routine radiological monitoring. The guidelines should also include recommendations on equipment for sampling, sample preparation, measuring radioactivity, measuring meteorological data, data acquisition and evaluation and data transmission.
454
Chapter 9
Taking into account the underlying basics, a mobile radiological unit for monitoring food and environmental samples should in the first instance have following components: 1. Truck (preferably air-cooled engine, all-wheel drive, suitable for operating independently and self-sustaining even in a hot climate (e.g. air-conditioning, special tyres, generator). It is supposed to house radiological laboratory, working and sample storage area. Design might include a movable laboratory container. 2. Basic equipment including: 9 gamma dose rate devices 9 gamma dose rate probe connected to a telesonde or mounted on a meteorological mast 9 pen dosimeters 9 contamination monitor 9 gamma dose rate instrument: sensitivity 10-v-10 -2 Sv/h 9 digital rate meter with external power supply 9 gamma proportional counter 9 hand monitor for surface contamination monitoring 9 air pollution sampler and filters (glass fibre) 9 flow rate: 80 m3/h. including iodine sampling, charcoal cartridges to be used with the air sampler 3. Other equipment 9 measuring device for iodine cartridges comprising digital rate meter, NaI detector (3"x3") and lead shielding, coupled to a PC, printer and software for evaluation of gamma spectra. 9 measuring device for air filter total beta and gamma measurements. 9 NaI bench-top gamma spectrometric system comprising of NaI detector (8% resolution), lead shielding to house Marinelli beakers and polyethylene bottles (0.5 or 1 litre), amplifier, high voltage devices etc. PC aided system with printer, plotter and software for spectra and data evaluation. 9 (Optional: HP Ge detector; rel. efficiency >20%, minimum resolution: 1.9 keV). 4. Additional equipment 9 provision of meteorological mast comprising of devices for measurement of: temperature, atmospheric pressure, humidity, wind direction, wind speed, and data transmission to indoor recording stations 9 acoustical and optical alarm system for in- and outdoor radioactive contamination 9 protective clothings and masks 9 refrigerator 9 homogenizer (kitchen blender) 9 grinder 9 balance 5. Data handling 9 recording station for meteorological data 9 recording station for radiological data 9 transmission station: telefax (wireless) or modem
Monitoring Accidentally Released Radionuclides in the Environment
455
9.9 M E A S U R E M E N T OF R A D I O A C T I V I T Y IN W A T E R Water contains a small and variable quantity of natural radioactivity from the decay of uranium and thorium and their daughters, together with 4~ The background radiation has been increased during the past three decades as a result of man's exploitation of nuclear fission. The original, and still the major, artificial input to the hydrosphere is from the fallout of fission products from nuclear weapon testing, and the presence, in rain and river waters, of 9~ and ~37Cs (and other radionuclides) has been well documented. Other sources of radioactivity now include the discharge of small quantities of liquid radioactive waste from the operation of nuclear-powered electricitygenerating stations and research establishments, the use of radioactive materials in industry and medicine, and also from the use of tracers for the investigation of water and sediment movement. The control of the uses of radioactivity, should ensure that the levels in water are below limits derived from the International Commission on Radiological Protection (ICRP) recommendations (ICRP, 1977). Where appropriate, the radioactive content of water is measured by the operator who is authorised to discharge radioactivity, and the results are checked by the appropriate authorising government Departments; in addition, tracer experiments to follow water movement are usually carried out by specialist groups with the appropriate measuring equipment. The measurement of the radioactive content of water is carried out by some Water Authorities as a check on trends and natural levels to be expected in the environment (see for example Greenberg et al., 1981). Tritium occurs most commonly in water as water itself (HTO). Except for radiological monitoring purposes, the main reason for determining tritium in water is to assess whether the supply is being replenished by rainfall or not. As mentioned earlier, tritiated organic compounds, usually insoluble, are used commercially mainly in luminous equipment or in medicine and research. Such material may eventually end up in tip leachates, effluents and incinerator gases. Tritium is a weak beta emitter, hence some form of concentration and special counting techniques are required, which will be dependent on the form in which the tritium is suspected of being present. In the most usual method for water samples, chlorine, if present, and radioiodine are removed by thiosulphate and acidity neutralised with sodium carbonate, solutes are removed by distillation to give a relatively pure water. Tritium in organic compounds is usually determined by first oxidising the substance to give tritiated water. For most wastes it is necessary to concentrate the tritium electrolytically, for details see Metson (1969) and Allen et al. (1966). A scintillating substance is added and the sample counted for several successive intervals of time using a liquid scintillation counter. The counts are quantified by comparison with tritium standards, allowance being made for decay and statistical errors. The Marine Environmental Laboratory of IAEA located in Monaco is involved in the measurement and understanding of marine radioactivity. Analytical quality assurance, the use of isotopes as tracers and site-specific radiological assessments are central
456
Chapter9
to IAEA-MEL's programme. The laboratory is involved in (a) independent UN observation, (b) radioanalytical and radiometric measurements, (c) analytical quality assurance and distribution of intercalibration materials, (d) maintenance of analytical databases, (e) measurement of biological concentration factors and of associated biokinetics, (f) computer modelling of possible waste dispersions and transfers and (g) radiological assessments of measured and modelled marine radioactivity data. IAEA-MEL' s recent studies at four locations where radioactive materials have been either intentionally or accidentally introduced into the marine environment are described by Baxter et al. (1996). These projects are (a) the former Soviet high-level waste dumpsites in the Kara Sea, (b) the former European dumpsites for low-level wastes in the North-East Atlantic, (c) the Norwegian Sea location at which the soviet nuclear-powered submarine Komsomolets, armed with nuclear weapons, accidentally sank in April 1989 and (d) the former soviet and Russian dumpsites for high-level nuclear wastes in the Far Eastern Seas. In the Kara Sea, about 4.5 PBq of activity is at present associated with nuclear reactors dumped in shallow waters. Work on this problem is done within IAEA as IASAP (International Arctic Seas Assessment Project) Programme. Sets of marine samples from each dumpsite, from the open Kara Sea and from the mouths of the Rivers Ob and Yenisey, which potentially could carry major inventories of artificial radionuclides from previously contaminated land in the Urals and Siberia, have been collected. Samples have afterwards been analysed for both man-made and natural radionuclides, the latter being important to the understanding of the rates and mechanisms of sedimentation and diagenesis. The results show that the open Kara Sea is relatively uncontaminated, the main contribution being from nuclear weapons test fallout and land-based sources. At certain coastal sites, there is localised evidence of minor waste leakage, contributions from surface and/or underwater nuclear testing and of river inputs. The IAEA-MEL has developed a dual-detector (NaI and HPGe) system for underwater and seabed gamma-spectrometry. High resolution data collected near one Kara Sea dumpsite show the absence of gamma-emitters from the dumped materials and the dominance of naturals. The main implication of these results is that no large-scale leakage has occurred in the past. A box model approach has been used to assess the order-of-magnitude global dosimetry which would follow leakage from the dumped waste. Preliminary estimates indicate that the global consequences of these disposals are minor and only potential exposures to local and regional populations may be significant. Preliminary results suggest that maximum individual dose rates below 1 ~tSv y-~ and 1 mSv y-~ could be delivered through the fish ingestion pathway at regional and local scales, respectively, following unit (1 TBq y-~) ~3VCsrelease rates. Prior to the above-related revelations about former Soviet disposals of radioactive wastes in the marine environment, more than 98% of packaged low-level radioactive material disposed of in the oceans was believed to be dumped at deep sites in the North Atlantic Ocean. 92% of the total activity was dumped in the eastern basin. While, in general, 98% of the total radioactivity disposed of comprised beta-gamma emitters, small quantities of alpha-emitting nuclides were also included. At the two main sites in
Monitoring Accidentally Released Radionuclides in the Environment
457
the North-East Atlantic (46~ 16~ and 46~ 17~ a total activity of more than 30 PBq was disposed of. The alpha-emitting inventory at these sites would be expected to be in the order of 0.5 PBq. These dump-sites were used until 1982 but were previously and subsequently subject to radiological survey, normally on an annual basis. In 1992 IAEA-MEL collected water samples above the sea-bed of the main sites for anthropogenic radionuclides such as 14C, 137Cs, 238pu, 239'24~ and 241Am. Samples were collected from the FRV Walther Herwig at one control site and at four locations in the area of the two main subsites. The results show enhancements of 5-7 times in the 238pu concentrations in sea water collected at the dumpsites relative to those at the control site. 239'24~ 241Am, 137Cs and 14C are also higher in some dumpsite samples. The 238pu/239'24~ activity ratio at the control site is similar to that expected for global fallout in the northern hemisphere and is considerably less than the ratios observed in water at the dumpsites (0.08-0.13). These preliminary results suggest that measurable leakage is occurring at the dumpsites. It should be borne in mind, however, that the highest observed activities (--0.6 mBq 1-1 137Cs, -20 gBq 1-~ 239'24~ are extremely small compared to the naturally occurring radioactivity of open ocean water (12.5 Bq 1-l of beta/gamma emitters, 115 mBq 1-~ of alpha emitters). Thus the localised enhancements represent increases of up to 10-3% and 10-2% in beta/gamma and alpha activities and these are therefore radiologically negligible. On 7 April 1989, a fire broke out in the stern section of the Komsomolets nuclear submarine. The submarine sank to a depth of 1685 m at 73~ 13~ 15'52"E, near the south-west of Bear Island. The site is about 300 nautical miles from the Norwegian coast. The wreck contains one nuclear reactor and two nuclear warheads, one of which was fractured. The radionuclide inventory includes 1.5 PBq 9~ 2 PBq 137Cs, about 16 TBq 239pu in the two warheads and 5 TBq of actinides in the reactor's core. During June/July 1994, an international expedition to the Komsomolets site at the request of the Russian Federation was organised. The objectives of the scientific cruise on board the R/V Mstislav Keldysh were to close nine door holes, including torpedo tubes, by capping them with titanium metal cover caps, and to sample and monitor for ambient radioactivity. A series of 280-600 1 sea-water samples collected in profile, a suite of surface sediments and cores and various biota samples were returned to IAEA-MEL for analysis. The results showed that a very limited leakage of caesium and tritium had occurred from the submarine. The former Soviet Union also disposed of a high, intermediate and low activity wastes in the Far Eastern Seas, although, unlike in the Arctic, no reactors containing fuel were dumped there. In the Sea of Japan region, therefore, the dumped inventory of radioactive waste is relatively low, comprising -440 TBq of solids. Within the framework of a joint agreement between the Governments of Japan, the Republic of Korea and the Russian Federation, IAEA-MEL was invited to participate in the first Japanese-Korean-Russian joint expedition to the radioactive-waste-dumping areas in their common seas. The cruise was carried out on board the R/V Okean during March/April 1994. Preliminary results obtained at the IAEA-MEL have shown that concentrations of ~37Cs, 9~ and 239+24~ in sea water and bottom sediments are very
458
Chapter 9
low and do not differ from global weapons fallout baseline levels observed in the North West Pacific Ocean (Baxter et al., 1996). This type of measurement often requires continuous monitoring. Continuous monitoring of radioactive nuclides in environmental watermfor example, the water sampled from nuclear power plants, radiochemical labs and hospitalsmis an indispensable routine procedure (Cember, 1983). Traditional monitoring methods are based on the measurement of the sampled waters, which in turn are placed in a low-background gamma detecting system for qualitative and quantitative analyses. Very low detection limits for multi-nuclides can be reached by use of a high-resolution, high-purity germanium (HPGe) detecting system. However, it usually involves tedious sampling processes prior to radiation counting, leading to the demand to develop a portable in situ gamma-ray spectroscopic device.
Fig. 9.28. Configuration of the HPGe probe with relative gamma contribution contour indicated in water for 1 MeV gamma rays. (A) Probe holder; (B) liquid-nitrogen dewar; (C) styrofoam; (D) HPGe detector; (E) lead block. (After Chao and Chung, 1992).
Monitoring Accidentally Released Radionuclides in the Environment
459
Chao and Chung (1992) have described an HPGe gamma probe designed to monitor radioactive nuclides in environmental water in situ. The probe is equipped with a 15% HPGe detector and an associated spectrum analyser. Laboratory tests were performed to evaluate its operating depth, detecting sensitivity, detecting volume and the detection limits of radionuclides. A field operation was conducted to measure in situ radionuclide concentrations in a nuclear reactor pool, and the feasibility as well as the disadvantages of this rapid survey are discussed. The configuration of the probe is shown in Fig. 9.28. Such a probe offers comparable detection limits for most of the radionuclides. The water surrounding the probe contains an environmental gamma source on the one hand, but at the same time provides the best gamma shield for the submerged detecting system. Water mass with thickness of 1 m is equivalent to a 9.1-cm lead shield for 1 MeV gamma rays, which is thick enough to reduce the gamma rays emitted from soils, from mud at the bottoms of fiver or ocean, or from the sky. Effective detecting volumes of water, estimated at around 1200 1, are much larger than the biggest Marinelli-type container reported for environmental aqueous sample measurement (Suzuki et al., 1988). Therefore, an in situ measurement probing a much larger sample size is more representative than those samples taken off-line. For an environmental survey in a flowing water mass in situ, on-line response is preferable, and timely, accurate results can be obtained. More common under-sea gamma spectroscopy systems are equipped with NaI probes. In such systems the NaI probe, along with its analogue circuitry (HV, preamplifier and amplifier), is housed in a pressure-proof container, connected to the instrument by a long and strong cable. This cable has the double function of electrical and mechanical connection i.e. withstanding the stress of pulling the probe underwater from a boat moving at considerable speed. In order to prevent breakage of electrical contacts caused by mechanical stress, the cable is mechanically tied to a box which is firmly fixed to the boat deck and electrically connected to the instrument.
9.10 THE C H E R N O B Y L ACCIDENT 9.10.1 Introduction Although much attention has been paid to safety in the various aspects of nuclear technology, some unfortunate nuclear accidents occurred in the history of the development of nuclear power. Major accidents are listed in Table 9.12 (Vinjamuri et al., 1982). Among them, some accidents, in which radioactivity was released into the atmosphere, are briefly described. Windscale-1 in 1957 The Windscale facility was an air-cooled, uranium-graphite plutonium production reactor. When the uranium metal fuel overheated and caught fire, it released fission products to the air stream discharging up a 400-ft stack over a two-day period.
460
-'= [--
O0
o,,,,.
>.
"E"
0
0
~,
Z
.m
.,...
0
,.o
--
m
x
trh
.=_
r
E
,.~
x~
C'4
9
~,~
L~ E
._
~
=
oO
"~
:c.
~
u~
"'
oo
9
~
~
=
,.-
~.~
~.
"~
o
,,,
oO
9. - ~
-
.-
cq
Z3
I=
I,,,.,
I=
,s
2~=
~-'~
--
~
"6
~
,.~
=~
~
o
0
0
I=
,.~
~
oo._,
--"
,.~
cq
"~
,,6
--"
Chapter 9
o ,-.,
',,o
~
;>-,
.~
0
LSL~ 0
7~
c~ 0
0
oo
~-~ k~ ~" .~
Y
"~
c,4
C'I
L. ,,,--, . ..,,,
X
O-O 9,....
C'4
o~ --~ % ~ ~.=
0
0
Z
x ~.=_ ~
=. =
~ ~ ~~
~ ~
= ~
!
oo
L~s
|
I::=
~=
Monitoring Accidentally Released Radionuclides in the Environment
":::t"
E
I=
7b
[-
d
.,..~
0
,_~
;~
~.~
r
.,-,
7,
~-~ "~.~ y
~
~'~
7,
9-
II
~
II
cI~
"r.~
.=-~~
ZZ-~
~o~~~o
..=
II
0
o
.0
. ,...,
o
,
"
~'~ eq
" ~#~~
r.~
~
~
461
462
Chapter 9
The burning uranium heated the graphite, which also burned to release CO and CO 2. About 2x104 Ci of gaseous iodine, which represented 12% of the available iodine inventory, were released into the atmosphere from the stack. The filter removed the particulate iodine (20 to 50x103 Ci). The radioactive plume was detected as far away as Germany and Norway. SL- 1 in 1961 SL-1 was a research reactor of the U.S. Army. In the SL-1 accident, a very large fraction of the core was destroyed and most of the primary water was ejected from the primary system. The reactor building filled with steam, which leaked to the environment from gaps in the doors on the operating floor, from open doors in the control room, and from the exhaust on the fan floor. About 104 Ci of noble gas and 80 Ci of iodine were released into the atmosphere. TM-2 in 1979 The TM-2 accident was initiated by a sudden stop of some pumps in the secondary system. Some erroneous operations and the inferiority of some equipment enhanced the accident. About three hours after the initial incident, the primary water overflowed onto the floor of the auxiliary building, and the radioactivity in the primary water, especially xenon, krypton and iodine, were released into the atmosphere. In total, 107 Ci of noble gases were released into the atmosphere, whereas the total amount of the iodine release was 17 Ci. Chernobyl Reactor 4 in 1986 This is the most serious accident in the history of the development of nuclear energy. It was caused by illegal operations. The reactor core was completely destroyed and about 50 MCi of noble gas was released in the first day, April 26th. Furthermore, about 50 MCi of other fission products were released into the atmospheric environment until May 6th. The radioactivity from Chernobyl was detected at many places in the northern hemisphere. A large area of Europe received significant surface deposition of radioactive materials such as ~3~Iand ~37Cs. The brief chronology of major events related to the Chernobyl accident is as follows: 26 April 1986 Accident occurs 01:23. Governmental Commission formed 27 April 1986 Evacuation of Pripyat takes place 6 May 1986 End of 10 days of atmospheric release of radioactive material from the core 6 May 1986 Evacuation of the population within the prohibited zone completed 31 May 1986 Revision of "temporary permissible levels" May 1986 "Temporary dose limits" for the population set at 100 mSv (internal and external) annual total dose
Monitoring Accidentally Released Radionuclides in the Environment
July 1986
463
First summarised contamination map (not published until 1989) November 1986 Completion of the "sarcophagus" construction 1987 "Temporary dose limits" for the population reduced to 30 mSv annual total dose (subsequently lowered to 25 mSv for 1988) April 1987 Completion of the work begun in May 1986 for protecting the water system December 1987 Revision of the "temporary permissible levels" established 31 May 1986 1988 "Temporary dose limits" for the population reduced to 25 mSv annual total dose September 1988 Council of Ministers of USSR adopts the 350 mSv lifetime dose for relocation to be implemented as of 1 January 1990 March 1989 Contamination maps officially published in the three Republics April 1989 BSSR Academy of Sciences registers disagreement with the 350 mSv life-time dose concept and makes new proposals October 1989 USSR requests the IAEA to organise an international assessment of the consequences of the accident and the protective measures taken The response to the request by the USSR government was a proposal for a multinational team to undertake an assessment of the radiological situation in the three affected Soviet Republics--the Ukrainian Soviet Socialist Republic (UkrSSR), the Byelorussian Soviet Socialist Republic (BSSR) and the Russian Soviet Federated Socialist Republic (RSFSR). The International Chernobyl Project was thus arranged, with the participation of the Commission of the European Communities (CEC), the Food and Agriculture Organisation of the United Nations (FAO), the International Labour Office (ILO), the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR), the World Health Organisation (WHO) and the World Meteorological Organisation (WMO). The Project was formalised at a February 1990 meeting in Moscow at the headquarters of the USSR State Committee on the Utilisation of Atomic Energy. The conclusions and recommendations of the International Chernobyl Project were approved by the International Advisory Committee (IAC) at its meeting in Vienna from 18 to 22 March 1991 and they are based upon the radiological and health assessments carried out by the Project. The technical details of these assessments are to be found in the extensive Technical Report (1991). In addition, valuable information about the accident can be found in the UNSCEAR Report (1998). Many countries submitted scientific data either directly to the UNSCEAR Secretariat or to the data bank set up in Vienna by the International Atomic
464
Chapter 9
Energy Agency. The UNSCEAR team of experts had free access to this data bank for the purpose of deriving data for the assessment. 9.10.2 The accident
On 26 April 1986 at 0123 hours local time an accident occurred at the fourth unit of the Chernobyl nuclear power station. The accident destroyed the reactor core and part of the building in which the core was housed. The radioactive materials released were carried away in the form of gases and dust particles by air currents. In this manner, they were widely dispersed over the territory of the Soviet Union, over many other (mostly European) countries and, in trace amounts, throughout the northern hemisphere. The Chernobyl nuclear power station is located in the Ukrainian Soviet Socialist Republic in the western USSR, near the boundary with the Byelorussian soviet Socialist Republic. It lies about 100 km north-west of Kiev and 310 km south-east of Minsk, on the River Pripyat, which flows into the Dnieper (Fig. 9.29), Poland (eastern part) and Romania (northern part), are 450 km away.
Fig. 9.29. The site of the Chernobyl nuclear power station.
Monitoring Accidentally Released Radionuclides in the Environment
The Chernobyl Unit 4 reactor had the 1986): Thermal power Fuel enrichment Mass of uranium in fuel assembly Fuel burn-up Maximum design channel power Isotopic composition of unloaded fuel
235U 236U
465
following principal specifications (IAEA, 3200 MW 2.0% 114.7 kg 20 MW d/kg 3250 kW
4.5 kg/t 2.4 kg/t 239pu 2.6 kg/t 24~ 1.8 kg/t 241pu 0.5 kg/t The radionuclide composition of the Chernobyl Unit 4 core is shown in Table 9.13. The accident happened while a test was being carried out on a turbine generator during a normal, scheduled shutdown of the Unit 4 reactor. The test was intended to ascertain the ability of a turbine generator, during station blackout, to supply electrical energy for a short period until the stand-by diesel generators could supply emergency power. Written test procedures that were unsatisfactory from the safety point of view, and serious violations of basic operating rules put the reactor at low-power [200 MW (th)] operation in coolant flow rate and cooling conditions that could not be stabilised by manual control. In view of the design features (the positive power coefficient at low power levels), the reactor was being operated in an unsafe regime. At the same time, the operators, deliberately and in violation of rules, withdrew most control rods from the core and switched off some important safety systems (IAEA-1986). The subsequent events led to the generation of an increasing number of steam voids in the reactor core, which enhanced the positive reactivity. The beginning of an increasingly rapid rise in power was detected, and a manual attempt was made to stop the chain reaction (the automatic trip, which the test would have triggered earlier, had been blocked). However, there was little possibility of rapidly shutting down the reactor as almost all the control rods had been completely withdrawn from the core. The continuous reactivity addition by void formation led to a prompt critical excursion. It was calculated that the first power peak reached 100 times the nominal power within four seconds. Energy released in the fuel by the power excursion suddenly ruptured part of the fuel into minute pieces. Small, hot fuel particles (possibly also evaporated fuel) caused a steam explosion. The energy released shifted the 1000-tonne cover plate of the reactor, cutting all the cooling channels on both sides of the reactor cover. After two or three seconds, another explosion occurred, and hot pieces of the reactor were ejected from the damaged reactor building. The damage to the reactor permitted the influx of air, which then caused the graphite to burn. Damage to the reactor containment and core structures led to the release of large amounts of radioactive materials from the plant. The release did not occur in a single
466
Chapter 9
Table 9.13 Core inventory and estimate of total release of radionuclides (after IAEA-1986) Radionuclide
Half-life
Inventory ( E B q )
Percentage released
8~Kr J33Xe 1311 132Te 137Cs 134Cs 89Sr 9~ 95Zr 99Mo l~ l~ 14~ 141Ce
10.72 a 5.25 d 8.04 d 3.26 d 30.0 a 2.06 a 50.5 d 29.12 a 64.0 d 2.75 d 39.3 d 368 d 12.7 d 32.5 d 284 d 2.36 d 87.74 a 24065 a 6537 a 14.4 a 163 d
0.033 1.7 1.3 0.32 0.29 0.19 2.0 0.2 4.4 4.8 4.1 2.1 2.9 4.4 3.2 0.14 0.001 0.0008 0.001 0.17 0.026
~100 ~100 20 15 13 10 4 4 3 2 3 3 6 2 3 3 3 3 3 3 3
144Ce 239Np 238pu
239pu 24~ 241pu 242Cm
massive event. On the contrary, only 24% of the materials released escaped during the first day of the accident; the rest escaped over a nine-day period. The estimated percentages of various radionuclides released from the total in the inventory are shown in Table 9.13. Only two earlier reactor accidents caused significant releases or radionuclides: the one at Windscale (United Kingdom) in October 1957 and the other at Three Mile Island (United States) in March 1979 (UNSCEAR-1982). While it is very difficult to estimate the fraction of the Windscale radionuclide core inventory that was released to the atmosphere, it has been estimated that the accident released twice the amount of noble gases that was released at Chernobyl, but 2,000 times less 131I and 137Cs(DOE-1987). The Three Mile Island accident released approximately 2% as much noble gases and 0.00002% as much 131Ias the Chernobyl accident. At the time of the accident, surface winds at the Chernobyl site were very weak and variable in direction. However, at 1500 m altitude the winds were 8-10 m/s from the south-east. The initial explosions and heat from the fire carried some of the radioactive materials to this height, where they were transported by the stream flow along the
Monitoring Accidentally Released Radionuclides in the Environment
467
western parts of the USSR toward Finland and Sweden. The arrival of radioactive materials outside the USSR was first noted in Sweden on 27 April (Devell et al., 1986). The transit time of 36 hours over a distance of some 1200 km indicates transfer at an average wind speed of 10 m/s. According to aircraft measurements within the USSR, the plume height exceeded 1200 m on 27 April and on subsequent days, the plume height did not exceed 200-400 m. The volatile elements iodine and caesium, were detectable at greater altitudes (6-9 km), with traces also in the lower stratosphere (Jaworowski et al., 1988). The refractory elements, such as cerium, zirconium, neptunium and strontium, were for the most part of significance only in local deposition within the USSR. Changing meteorological conditions, with winds of different directions at various altitudes, and continuing releases over a 10-day period resulted in a very complex dispersion pattern. The plumes are shown in a simplified manner in Fig. 9.30, along
Fig. 9.30. Plume behaviour and reported initial arrival times of detectable activity in air. Plumes A, B and C correspond to air mass movements originating from Chernobyl on 26 April, 27-28 April, and 29-30 April, respectively. The numbers 1 to 8 indicate initial arrival times: 1 (26 April), 2 (27 April), 2 (28 April), 4 (29 April), 5 (30 April), 6 (1 May), 7 (2 May), 8 (3 May).
468
Chapter 9
with the reported initial arrival times of radioactive material. The initial plume arrived on 27 April in Sweden and Finland. A portion of this plume at lower altitude was directed southward to Poland and the German Democratic Republic. Long-range atmospheric transport spread the released activity throughout the northern hemisphere. Reported initial arrival times were 2 May in Japan, 4 May in China, 5 May in India, and 5-6 May in Canada and the United States. The simultaneous arrival at both western and eastern sites in Canada and the United States suggests a large-scale vertical and horizontal mixing over wide areas. No airborne activity from Chernobyl has been reported in the southern hemisphere. The effective dose equivalents received by individuals (adults) during the first year following the accident show rural-urban differences. Contributions to dose from the ingestion pathway also include committed doses from caesium in the body following the first-year intake of caesium in diet. The highest average first-year committed effective dose equivalent in subregions was 2 mSv in the Byelorussian Soviet Socialist Republic. Subregions where effective dose equivalents were 1-2 mSv were located in Romania and Switzerland and 0.5-1 mSv in Austria, Bulgaria, Federal Republic of Germany, Greece and Yugoslavia. The effective dose equivalent in the Byelorussian Soviet Socialist Republic approached the yearly effective dose equivalent due to natural radiation sources. The mean values for each country are listed in Table 9.14 (UNSCEAR- 1988).
9.10.3 Environment contamination
The Chemobyl accident involved the largest short-term release from a single source of radioactive materials to the atmosphere ever recorded. Of the materials released from the reactor core, four elements have dominated the short-term and long-term radiological situation in the affected areas of the USSR: iodine (primarily ~3~I),caesium (~34Cs, 137Cs), strontium (primarily 9~ and plutonium ,,239r.~ t r'u, 240pu ) . In addition, highly radioactive fuel fragments (hot particles) were released. The destroyed reactor released a very large amount of radioactive material into the environment: 1019 becquerels. Although the discharge included many radioactive chemical elements, just two of them--iodine (in the short term) and caesium (in the long term)--were particularly significant from a radiological point of view. About 10 TMbecquerels of iodine- 131 were released in the accident. Iodine is mainly absorbed by a person' s thyroid gland after inhalation or after consumption of contaminated foodstuffs such as milk products; its short-range beta particles irradiate the gland from the inside. Uptake of iodine by the thyroid is very easy to prevent, for example by banning consumption of contaminated food for a few weeks until the iodine- 131 decays sufficiently or by administering small amounts of non-radioactive iodine prophylactically to block the thyroid gland. About 1017becquerels of radioactive caesiums were released, and precipitated over a vast area. Exposure to caesium is difficult to prevent. Once it is deposited in the soil,
Monitoring Accidentally Released Radionuclides in the Environment
469
Table 9.14 Country average of first-year dose equivalent Country
EUROPE Bulgaria Austria Greece Romania Finland Yugoslavia Czechoslovakia Italy Poland Switzerland Hungary Norway German (Dem. Rep.) Sweden Germany (Fed. Rep.) Ireland Luxemburg France Netherlands Belgium Denmark United Kingdom Spain Portugal USSR ASIA Turkey Israel Cyprus Syrian Arab Rep. China Japan India NORTH AMERICA Canada United States
Thyroid dose equiv.
Effective dose equiv. (~tSv)
Infant (~tSv)
Adult (~tSv)
25000 9400 20000 18000 1800 14000 2200 3400 8100 15000 6000 1000 5100 1000 1700 2500 2700 1600 940 2300 160 710 110 9 5000
29000 1800 5000 2800 1200 5500 2700 1500 1400 2300 1000 570 970 340 440 540 580 360 390 460 64 130 24 4 1400
2300 1500 4700 1400 390 210 69
480 1100 1200 74 47 100 5
75 110
11 15
1.4
Ratio to result reported from country (N5) Thyroid dose
Effective dose
Infant
Adult
760 670 590
1.2 3.6
1.0 2.6
1.0 1.6
570 460
1.0
1.7
0.9
390 35O 300
0.5
0.5
0.6
270 270
9.3
2.1
1.2
230 230
0.8
1.5
1.4
210 150 130 120 98 63 58 41 30 27
2.0 0.6 0.2 3.5 1.8 0.6 1.7 0.6 0.3
0.9 0.5 2.3 1.7 4.1 1.3 2.2 1.3 0.8
0.7 0.4 1.1 0.8 2.6 0.8 1.0 1.1 0.7
4.2 1.8
0.1
0.4
0.3
190
0.7
1.2
2.2
92 68 8.3 7.8 7.6
1.4
2.1
1.2
4.2
6.5
0.6
260
2.1
1.5
470
Chapter 9
its long-range gamma rays can expose anybody in the area. To clean the surfaces is difficult and, if the concentration of caesium is high, often the only feasible countermeasure is to evacuate the inhabitants. Caesium in the soil can also be transferred into agricultural products and grazing animals. For iodine-131, there is no clear information on where the release went, who was exposed to it and to which levels, or whether iodine uptake was effectively prevented. Indirect estimations gave a firm indication that very high thyroid doses were incurred by some population groups. Children, who are particularly sensitive because of their normally high ingestion of milk products and their small thyroids, received higher doses. Aerial radiation measurements and environmental sampling begun shortly after the accident showed that the highest level of environmental contamination was in the area around the reactor; that would eventually become the prohibited zone. Elsewhere in the Soviet Union, changing wind conditions and sporadic rainfall over the ten-day release phase resulted in a very uneven pattern of radioactive fallout within areas of the BSSR, the RSESR and the UkrSSR. Heavy rainfall combined with local conditions to create pockets ("hot spots") of exceptionally high surface radioactivity levels resulting in external dose rates that were as much as five thousand times the dose rate due to the natural background. Once releases had been halted, changes in contamination patterns resulted from radioactive decay (primarily of 131I, which decays almost totally within three months) and normal weathering processes which caused the migration of contamination into the soil and the dispersion of soil particles through the runoff of surface waters. Information from continuing aerial surveys and environmental sampling has been used to derive official surface contamination maps which display the ranges of surface concentration of caesium, strontium and plutonium. Officially published in 1989, the maps have stirred controversy among scientists and residents. About 25,000 km: and 2225 settlements in the three Republics are officially defined as having a 137Cs surface contamination in excess of 185 kBq/m: (5 Ci/km2). An intercomparison exercise organized by the IAEA Laboratory at Seibersdorf provided a yardstick for judging the validity of official data (see Cooper et al., 1992 for details). The 13 institutes that took part are reported to be the most heavily engaged in sampling and laboratory analyses of environmental materials and foods. The institutes analyzed "blind" samples of (radionuclides measured): soil (9~ e39pu, 137Cs, 226Ra); milk powder (9~ 134Cs, 137Cs,4~ simulated air filters (9~ 137Cs,6~ 133Ba, :~~ and vegetation (9~ 134Cs, 137Cs, 4~ and reported the results together with the associated numerical uncertainties. The IAEA Laboratory compared their results with the recommended (i.e. "reference") values. The reported results for ~3VCsin soil agreed well with the recommended values (see Table 9.15). On the other hand, results for strontium and plutonium in soil showed a tendency for overestimating (by as much as a factor of four). A similar tendency for overestimation was noted for strontium in milk (by as much as a factor of nine) and for caesium in milk (by as much as a factor of three). While results for strontium in
Monitoring Accidentally Released Radionuclides in the Environment
471
Table 9.15 Comparison of performance of the two groups of laboratories" worldwide vs. Soviet Union Range of reported values for milk(H) (Bq/kg)
Radionuclide
Worldwide
USSR
137Cs
469.3-2491.3
173-3070
134Cs
58.0-652.5
184.7-542.5
4~
103.6-3650.0
429-4959
9~
5.53-8.54
1.43-68.8
vegetation appear generally reliable, there was an observed slight tendency to underestimate caesium. In simulated air filters the results for caesium were in agreement with recommended values while the results for strontium deviated by 30-50%. Figure 9.31 shows the performance of different laboratories investigated in the case of high 13VCs concentrated in milk powder. The analytical capabilities of Soviet laboratories appeared to be adequate. There is an extensive infrastructure for the analysis of environmental and food samples. The 3500-
+
3000"
2500" -+..
exl)
+
2000-F-- +
"-+-~
..--_
i..~=._ _ ~ . , . + _ .
i
_,__ ~
t
...-
-+-
cw-_
--I--~
1500-4---
Recommended value 9 2065 Confidence interval 9 1991 - 2143
1000-
500 -
0
I '
I
11 22
'
!
"i
7
15 3
i .....
l
!
16 23
....
2
1'7 14
!
;
2~1 5 2;
!
"
1'3 '~
!
1
I ""
10
!
I
I
9
18
6'
!
12
L A B O R A T O R Y CODE Fig. 9.31. Intercalibration exercise of SSSR laboratories in determination of '37Cs in milk powder.
-I
8
472
Chapter 9
range of performance of the Soviet laboratories that participated in the inter-comparison exercise was broad, but similar to that found in previous international comparison exercises. The few problems identified, including the tendency to over-estimate strontium, did not significantly affect the use of data for conservative dose assessment purposes.
9.10.4 Hot particles Following a nuclear accident, deposited radionuclides may be present in different physico-chemical forms, ranging from mobile low molecular mass (LMM) ionic species to inert high molecular mass (HMM) colloidal forms or particles. Even in areas far from the actual site, the relative fraction of radionuclides associated with HMM formed in rain-water may be substantial (Salbu, 1988). The size distribution patterns of radionuclides deposited, the composition of the fallout, level of activities and the activity ratios, will depend on the accident scenario, course of event, distance from the source, wind dispersion and climatic or microclimatic conditions. Spatial and temporal variations in the behaviour of deposited radionuclides with respect to mobility and bioavailability are to be expected and may in part be attributed to differences in the physico-chemical forms of radionuclides in the fallout, at least during the first years after deposition (Salbu et al., 1994). The original distribution of deposited species will change owing to interactions with naturally occurring components. Sorption of LMM species to clay materials or complexation with organic ligands may reduce the mobility and bioavailability, while radionuclides associated with inert fuel particles may be mobilized with time, owing to weathering. Most models assessing the long-term behaviour of 137Cs and 9~ from fallout include processes relevant to LMM ionic species only. However, radionuclides may be associated with particles due to (a) release of fuel matrix or clusters, (b) condensation of volatiles on available particle surfaces after the release of (c) interactions with aerosol 137~--~ particulates during atmospheric transport. For volatile radionuclides (e.g., t..s, 9~ all three mechanisms may be equally relevant, while the deposition of nuclides of refractory elements (e.g., 144Ce, 95Nb, 95Zr) indicates the release of fuel particles. For fuel particles, depletion of volatiles relative to refractory elements would be expected to depend on the temperature reached during the releases, whereas the activity ratios for refractory elements should reflect the reactor fuel burn-up. Radioactive particles have been identified in connection with accidental releases from nuclear installations under high- and low-temperature conditions, in particular in releases from the accident in Unit 4 at Chernobyl (Loshchilov et al., 1992; Devell et al., 1986; Raunemaa et al., 1988) in 1986 and releases from the Windscale piles both during the fire in 1957 (Arnold, 1992) and earlier during the normal operation of the plant (Jakeman, 1986). Following the Chernobyl accident, relatively large fuel particles containing refractory elements (e.g. 95Zr, ~44Ce)having activity ratios close to those estimated for the
Monitoring Accidentally Released Radionuclides in the Environment
473
reactor fuel burn-up of Unit 4 have been identified as a major contributor to the total activity deposited within 60 km. Based on activity ratios, the relative distribution of fuel and condensed particles has been estimated. Among particles identified within 10 km, more than 95% were attributed to fuel and only less than 3% can be attributed to condensed particles. Owing to the high temperature involved, fuel particles were largely depleted in Cs isotopes and to a lesser extent in 9~ However, larger fuel particles were less depleted in Cs isotopes than smaller particles (Loshchilov et al., 1992). The accident at Chernobyl Nuclear Power Plant and considerable release of radionuclides in particulate fraction renewed interest in "hot particles" (HPs)~tiny objects of ~tm dimensions, having density of activity comparable with the one of irradiated nuclear fuel. They pose radiological risk, especially when inhaled with the air after resuspension from the soil. Studies (Osuch et al., 1989; Piasecki et al., 1990), performed on a quite large set of HPs (over 200 species) collected in Autumn 1986, indicated the existence of two, roughly equally populated groups of HPs: 9 group A ("ruthenial"; 88HPs)~showing exclusively 7-activity of l~ and I~ isotopes, 9 group B ("fuel-like"; 114 HPs)mcontaining different radioisotopes in relative amounts approximately the same as in the reactor core. According to by Ter-Saakov et al. (1991), the population of hot particles prevailing within the exclusion zone around the reactor comprises mostly "fuel-like" objects (=90%) with some contribution from "condensational" component. By the latter term the cited work denominates HPs enriched in volatile fission products (T, Cs, Te), probably formed in a process of condensation of their vapours on inert carriers. It is interesting that "ruthenial" HPs are quite scarce in the exclusion zone. "The Chernobyl's HPs" database of the Ukrainian Institute of Agricultural Radiology (UIAR), Kiev, Ukraine, CIS, collecting data on about 1200 HPs, mostly from 10 km zone, comprises less than 5% of particles of this type. For comparison, about half of 206 HPs collected in North-East Poland (HP Data Bank of Warsaw University (WU) belong to the considered group (Dabrowska et al., 1987). Regarding the origin of the Chernobyl "fuel-like" HPs, there is wide acceptance of an idea that essentially they consist of the reactor fuel pulverized and dispersed into the atmosphere during the accident. The origin of "ruthenial" HPs is still unknown. According to some hypothesis, particles of this type belong to the class of so-called "white inclusions" common in irradiated fuel in normal reactor operation (Schubert and Behrend, 1987; Antonov et al., 1987; Piasecki et al., 1990). Tcherkezian et al. (1994) have described an interesting experimental investigation of Chernobyl hot particles. In their study hot particles (HP) were picked out from soil samples collected during the 1986-1990 radiogeochemical expeditions in the contaminated zone (within 30 km of the Nuclear Power Plant). A number of hot particles were studied to estimate their contribution to the total activity, investigate their surface morphology and determine the size distribution. The contribution of hot particles to the total activity in the 30 km zone was found to be not less than 65%. Investigation of HP
474
Chapter 9
element composition (by neutron activation analysis and EPMA) and radionuclide composition (direct alpha- and gamma-spectrometry, including determination of Pu and Am in HP) revealed certain peculiarities of HP, collected in the vicinity of the damaged Nuclear Power Plant. Some particles were shown to contain uranium and fission products in proportion to one another, correlating with those in the partially burnt nuclear fuel, which proves their "fuel" origin. Another part of the HP samples has revealed elements fractionation as well as the presence of some terrestrial components.
9.10.5 A review of the accident--ten years later
Many international initiatives followed the International Chernobyl Project, including those highlighted here (after Gonzales, 1996). An agricultural countermeasures project was sponsored by FAO and IAEA. Following a specific request by Belarus at the 1994 IAEA General Conference, the IAEA engaged in a mainly environmental project on "prospects for the contaminated area". The project has been financed mostly by IPSN, which was heavily involved in its technical implementation that extends beyond the general conclusions of the ICP to cover the general environment. Referring to the forested biocoenosis--the environmental system that had reportedly suffered most from the Chernobyl accident--the project concluded that the radioactive contamination was not on a massive scale and affected mainly pine forests: the death of the pine plantations, although severe in the immediate vicinity of the plant, amounted to less than 0.5% of the forested area of the exclusion zone. The WHO International Programme on the Health Effects of the Chernobyl Accident (IPHECA): The results of the IPHECA project were recently published and discussed at the WHO International Conference on the Health Consequences of the Chernobyl and other Radiological Accidents, held in Geneva, 20-23 November 1995. IPHECA generally confirmed the conclusions of the ICP and provided additional information on the increase in child thyroid cancer incidence foreseen by the ICP. The IPHECA conclusions can be summarized as follows: 9 Psychosocial effects, believed to be unrelated to radiation exposure, resulted from the lack of information immediately after the accident, the stress and trauma of compulsory relocation to less contaminated areas, the breaking of social ties, and the fear that radiation exposure could cause health damage in the future. 9 A sharp increase in thyroid cancer was reported, especially among children living in the affected areas. By the end of 1994, 565 children aged 0-14 years were diagnosed as having thyroid cancer (333 in Belarus, 24 in the Russian Federation, 208 in Ukraine). 9 There was no significant increase in the incidence of leukaemia or other blood disorders. 9 Some evidence was found to suggest retarded mental development and deviations in behavioral and emotional reactions in a small number of children
Monitoring Accidentally Released Radionuclides in the Environment
475
exposed to radiation in utero; however, the extent to which radiation might have contributed to such mental changes cannot be determined because of the absence of individual dosimetry data. 9 The types and distribution of oral diseases observed in the residents of "contaminated" areas were the same as those of the residents of "uncontaminated" areas. Projects supported by the European Commission (EC): The EC supported many scientific research projects on Chernobyl's consequences. The results were summarized at the First International Conference of the European Union, Belarus, the Russian Federation and the Ukraine on the Consequences of the Chernobyl Accident, held in Minsk, on 18-22 March 1996. The projects produced valuable information that can be used for future emergency planning, dose assessment and environmental remediation as well as in the treatment of highly exposed individuals and in screening for thyroid cancer in children. Other initiatives: These include several UNESCO-supported studies, mainly on psychological consequences; special reports from UNSCEAR and the Nuclear Agency of the OECD; and individual studies in the affected States and in other countries, e.g. comprehensive monitoring of the affected people carried out by Germany, an extensive study sponsored by Japan's Sasakawa foundation, a major USA project and a large Cuban assessment on the intake of caesium-137, covering about 15,000 children. April 1996: the International Conference on One Decade After Chernobyl- Summing up the Accident's Consequences. The main organizations involved in assessing the Chernobyl accident' s consequences, namely the IAEA, WHO and EC, united their efforts in co-sponsoring that Chernobyl Conference. They organized the event in co-operation with the UN itself (through its Department of Humanitarian Affairs), UNESCO, UNSCEARX, FAO and the Nuclear Energy Agency of OECD. The Chernobyl Conference was attended by 845 scientists from 71 countries and 20 organizations and covered by 280 journalists. It was presided over by Germany's Federal Minister for the Environment, Nature Conservation and Nuclear Safety and attended by high-level officials and members of government, including the President of Belarus, the Prime Minister of Ukraine, and the Russian Federation' s Minister for Civil Defence, Emergencies and Elimination of Consequences of Natural Disasters, as well as by France's Minister for the Environment. Three national reports, 4 addresses by intergovernmental organizations, 11 keynote presentations, 8 background papers, 181 detailed poster papers and 12 technical exhibits provided the basis for this summing up of the Chernobyl accident's consequences. Some of the calculations of the conference are summarized by Gonzalez (1996). Health effects attributed to the accident have commanded the most concern on the part of the public, decision-makers and political authorities, and the Chernobyl Conference devoted a great deal of time to the topic. Clinically observed (and individually attributable) effects were discussed separately from long-term effects which can only be attributed to radiation after long studies of the statistical epidemiological nature of large populations. Among the latter, thyroid effects is a special case that was treated separately from other longer-term health effects.
476
Chapter 9
Clinically observed effects: The number of people who suffered clinically observed health effects individually attributable to radiation exposure due to the Chernobyl accident was relatively modest, given the accident's dimensions. A total of 237 persons, all of them workers dealing with the accident, were suspected of suffering clinical syndromes of radiation exposure and were hospitalized, and 134 of them were diagnosed with acute radiation syndrome. Of these, 28 died of the consequences of radiation injuries (three other persons died at the time of the accident: two due to non-radiation blast injuries and one due to a coronary thrombosis). Some years after the accident, 14 additional persons in this group died; however, their deaths were found to be not necessarily attributable to radiation exposure. The situation in relation to thyroid effects is serious. Up to the end of 1995, there were more than 800 cases of thyroid cancer reported in children, mainly in Belarus. Thyroid cancer may be induced by causes other than radiation, but all these cases seem likely to be associated with radiation exposure due to the accident. They represent a dramatic increase in the normal incidence of this rare type of cancer and the increase seems not to persist among children born after 1986. Thyroid cancer is usually non-fatal with early diagnosis, treatment and attention. At the time of the Chernobyl Conference, three of the children affected had already died. The prospects cannot be precisely predicted; the high incidence is expected to continue for some time and the number of reported cases may be in the thousands; the mortality will depend very much on the quality and intensity of the treatment given to the affected children. There is no evidence to date of any increase in the incidence of any malignancies other than thyroid carcinoma or of any hereditary effects attributable to radiation exposure caused by the Chernobyl accident. This conclusion, surprising for some observers, is in accordance with the relatively small whole body doses incurred by the populations exposed to the radioactive material released. The lifetime doses expected to be incurred by these populations are also small. In fact, the risks of radiation-induced malignancies and hereditary effects are extremely small at low radiation doses and, as the normal incidences of these effects in people are relatively high, it is not surprising that no effects could be detected. An exception to the lack of evidence of long-term effects might have occurred in the group of liquidators" taking into account the relatively high doses reported in this group, an increase in the incidence of leukaemia might have been detected. For all other malignancies and hereditary effects, the theoretically predicted number of causes due to radiation exposure from the accident are so small in comparison with the background incidence as to be impossible to confirm statistically (Gonzales, 1996). The Chernobyl Conference found that social, economic, institutional and political impacts were also important consequences of the Chernobyl accident. Large economic losses attributed to the accident were reported in this official document and also in the national statements delivered at the Chernobyl Conference. Certainly, a major social problem lies in the significant psychological symptoms detected among the population, such as anxiety, depression and various psychosomatic disorders attributable to mental distress.
Monitoring Accidentally Released Radionuclides in the Environment
477
9.11 C O N V E N T I O N ON EARLY N O T I F I C A T I O N OF A N U C L E A R ACCIDENT The Convention was adopted on 26 September 1986, during the IAEA 8th plenary meeting and pursuant to Article 12.3 of the Convention, entered into force on 27 October 1986. Below is the full text of the Convention. The states party to this Convention, Aware that nuclear activities are being carried out in a number of States, Noting that comprehensive measures have been and are being taken to ensure a high level of safety in nuclear activities, aimed at preventing nuclear accidents and minimizing the consequences of any such accident, should it occur, Desiring to strengthen further international co-operation in the safe development and use of nuclear energy, Convinced of the need for States to provide relevant information about nuclear accidents as early as possible in order that transboundary radiological consequences can be minimized, Noting the usefulness of bilateral and multilateral arrangements on information exchange in this area, Have agreed as follows:
Article 1. Scope of application 1. This Convention shall apply in the event of any accident involving facilities or activities of a State Party or of persons or legal entities under its jurisdiction or control, referred to in paragraph 2 below, from which a release of radioactive material occurs or is likely to occur and which has resulted or may result in an international transboundary release that could be of radiological safety significance for another State. 2. The facilities and activities referred to in paragraph 1 are the following: (a) any nuclear reactor wherever located; (b) any nuclear fuel cycle facility; (c) any radioactive waste management facility; (d) the transport and storage of nuclear fuels or radioactive wastes; (e) the manufacture, use, storage, disposal and transport of radioisotopes for agricultural, industrial, medical and related scientific and research purposes; and (f) the use of radioisotopes for power generation in space objects.
Article 2. Notification and information In the event of an accident specified in article 1 (hereinafter referred to as a "nuclear accident"), the State Party referred to in that article shall: (a) forthwith notify, directly or through the International Atomic Energy Agency (hereinafter referred to as the "Agency"), those States which are or may be physically affected as specified in Article 1 and the Agency of the nuclear accident, its nature, the time of its occurrence and its exact location where appropriate; and
478
Chapter 9
(b) promptly provide the States referred to in sub-paragraph (a), directly or through the Agency, and the Agency with such available information relevant to minimizing the radiological consequences in those States, as specified in Article 5. Article 3. Other Nuclear Accidents With a view to minimizing the radiological consequences, State Parties may notify in the event of nuclear accidents other than those specified in Article 1. Article 4. Functions of the Agency The Agency shall: (a) forthwith inform State Parties, Member States, other States which are or may be physically affected as specified in Article 1 and relevant international intergovernmental organizations (hereinafter referred to as "international organizations") of a notification received pursuant to sub-paragraph (a) of Article 2; and (b) promptly provide any State Party, Member State or relevant international organization, upon request, with the information received pursuant to subparagraph (b) of Article 2. Article 5. Information to be Provided 1. The information to be provided pursuant to sub-paragraph (b) of Article 2 shall comprise the following data as then available to the notifying State Party: (a) the time, exact location where appropriate, and the nature of the nuclear accident; (b) the facility or activity involved; (c) the assumed or established cause and the foreseeable development of the nuclear accident relevant to the transboundary release of the radioactive materials; (d) the general characteristics of the radioactive release, including, as far as is practicable and appropriate, the nature, probable physical and chemical form and the quantity, composition and effective height of the radioactive release; (e) information on current and forecast meteorological and hydrological conditions, necessary for forecasting the transboundary release of the radioactive materials; (f) the results of environmental monitoring relevant to the transboundary release of the radioactive materials; (g) the off-site protective measures taken or planned; (h) the predicted behaviour over time of the radioactive release. 2. Such information shall be supplemented at appropriate intervals by further relevant information on the development of the emergency situation, including its foreseeable or actual termination. 3. Information received pursuant to sub-paragraph (b) of Article 2 may be used without restriction, except when such information is provided in confidence by the notifying State Party.
Monitoring Accidentally Released Radionuclides in the Environment
479
Article 6. Consultations A State Party providing information pursuant to sub-paragraph (b) of Article 2 shall, as far as is reasonably practicable, respond promptly to a request for further information or consultations sought by an affected State Party with a view to minimizing the radiological consequences in that State. Article 7. Competent Authorities and Points of Contact 1. Each State Party shall make known to the Agency and to other States Parties, directly or through the Agency, its competent authorities and points of contact responsible for issuing and receiving the notification and information referred to in Article 2. Such points of contact and a focal point within the Agency shall be available continuously. 2. Each State Party shall promptly inform the Agency of any changes that may occur in the information referred to in paragraph 1. 3. The Agency shall maintain an up-to-date list of such national authorities and points of contact as well as points of contact of relevant international organizations and shall provide it to State Parties and Member States and to relevant international organizations. Article 8. Assistance to State Parties The Agency shall, in accordance with its Statute and upon a request of a State Party which does not have nuclear activities itself and borders on a State having an active nuclear programme but not Party, conduct investigations into the feasibility and establishment of an appropriate radiation monitoring system in order to facilitate the achievement of the objectives of this Convention. Article 9. Bilateral and Multilateral Arrangements In furtherance of their mutual interests, States Parties may consider, where deemed appropriate, the conclusion of bilateral or multilateral arrangements relating to the subject matter of this Convention. Article 10. Relationship to Other International Arrangements This Convention shall not affect the reciprocal rights and obligations of State Parties under existing international agreements which relate to the matters covered by this Convention, or under future international agreements concluded in accordance with the object and purpose of this Convention. Article 11. Settlement of Disputes 1. In the event of a dispute between State Parties, or between a State Party and the Agency, concerning the interpretation or application of this convention, the parties to the dispute shall consult with a view to the settlement of the dispute by negotiation or by any other peaceful means of settling disputes acceptable to them. 2. If a dispute of this character between State Parties cannot be settled within one year from the request for consultation pursuant to paragraph 1, it shall, at the request of
480
Chapter 9
any party to such dispute, be submitted to arbitration or referred to the International Court of Justice for decision. Where a dispute is submitted to arbitration, if, within six months from the date of the request, the parties to the dispute are unable to agree on the organization of the arbitration, a party may request the President of the International Court of Justice or the Secretary-General of the United Nations to appoint one or more arbitrators. In cases of conflicting requests by the parties to the dispute, the request to the Secretary-General of the United Nations shall have priority. 3. When signing, ratifying, accepting, approving or acceding to this Convention, a State may declare that it does not consider itself bound by either or both of the dispute settlement procedures provided for in paragraph 2. The other State Parties shall not be bound by a dispute settlement procedure provided for in paragraph 2 with respect to a State Party for which such a declaration is in force. 4. A State Party which has made a declaration in accordance with paragraph 3 may at any time withdraw it by notification to the depositary. Article 12. Entry into Force 1. This Convention shall be open for signature by all States and Namibia, represented by the United Nations Council for Namibia, at the Headquarters of the International Atomic Energy Agency in Vienna and at the Headquarters of the United Nations in New York, from 26 September 1986 and 6 October 1986 respectively, until its entry into force or for twelve months, whichever period is longer. 2. A State and Namibia, represented by the United Nations Council for Namibia, may express its consent to be bound by this Convention either by signature, or by deposit of an instrument of ratification, acceptance or approval following signature made subject of ratification, acceptance or approval, or by deposit of an instrument of accession. The instruments of ratification, acceptance, approval or accession shall be deposited with the depositary. 3. This Convention shall enter into force thirty days after consent to be bound has been expressed by three States. 4. For each State expressing consent to be bound by this Convention after its entry into force, this Convention shall enter into force for that State thirty days after the date of expression of consent. 5. (a) This Convention shall be open for accession, as provided for in this article, by international organizations and regional integration organizations constituted by sovereign States, which have competence in respect of the negotiation, conclusion and application of international agreements in matters covered by this Convention. (b) In matters within their competence such organizations shall, on their own behalf, exercise the fights and fulfil the obligations which this convention attributes to State Parties. (c) When depositing its instrument of accession, such an organization shall communicate to the depositary a declaration indicating the extent of its competence in respect of matters covered by this Convention.
Monitoring Accidentally Released Radionuclides in the Environment
481
(d) Such an organization shall not hold any vote additional to those of its Member States.
Article 13. Provisional Application A State may, upon signature or at any later date before this convention enters into force for it, declare that it will apply this convention provisionally. Article 14. Amendments 1. A State Party may propose amendments to this Convention. The proposed amendment shall be submitted to the depositary who shall circulate it immediately to all other State Parties. 2. If a majority of the State Parties request the depositary to convene a conference to consider the proposed amendments, the depositary shall invite all State Parties to attend such a conference to begin not sooner than thirty days after the invitations are issued. Any amendment adopted at the conference by a two-thirds majority of all State Parties shall be laid down in a protocol which is open to signature in Vienna and New York by all State Parties. 3. The protocol shall enter into force thirty days after consent to be bound has been expressed by three States. For each State expressing consent to be bound by the protocol after its entry into force, the protocol shall enter into force for that State thirty days after the date of expression of consent. Article 15. Denunciation 1. A State Party may denounce this Convention by written notification to the depositary. 2. Denunciation shall take effect one year following the date on which the notification is received by the depositary. Article 16. Depositary 1. The Director General of the Agency shall be the depositary of this Convention. 2. The Director General of the Agency shall promptly notify State Parties and all other States of: (a) each signature of this convention or any protocol of amendment; (b) each deposit of an instrument of ratification, acceptance, approval or accession concerning this Convention or any protocol of amendment; (c) any declaration or withdrawal thereof in accordance with Article 11; (d) any declaration of provisional application of this Convention in accordance with Article 13; (e) the entry into force of this Convention and of any amendment thereto; and (f) any denunciation made under Article 15. Article 17. Authentic Texts and Certified Copies The original of this Convention, of which the Arabic, Chinese, English, French, Russian and Spanish texts are equally authentic, shall be deposited with the Director
482
Chapter 9
General of the International Atomic Energy Agency who shall send certified copies to State Parties and all other States. In witness whereof the undersigned, being duly authorized, have signed this Convention, open for signature as provided for in paragraph 1 or Article 12. Adopted by the General Conference of the International Atomic Energy Agency meeting in special session at Vienna on the twenty-sixth day of September one thousand nine hundred and eighty-six.
REFERENCES Aarkrog, A., The radiological impact of the Chernobyl debris compared with that from nuclear weapons fallout. J. Environ. Radioact., 6 (1988) 151. Aarnio, P.A., Routti, J.T., Sandberg, J.V., and Winberg, M.J., Adapting gamma-spectrum analysis program SAMPO for microcomputers. Nucl. Instrum. Meth., 219 (1984) 173. Allen, R.A., Smith, D.B., Otlet, R.L., and Rawson, D.S., Nucl. Instrum. Meth., 45 (1966) 61-71. Antonov, A., et al., Proc. of Int. Workshop on Hot Particles from the Chernobyl Fallout, October 28-29, 1987, Teuern, Germany. Arnold, L., Windscale. Macmillan, London, 1992. Jakeman, D., Notes of the Level of Radioactive Contamination in the Sellafield Area Arising from Discharges in the Early 1950s, UKAEA, AEEW Report 2104, Atomic Energy Establishment, Winfrith, 1986. Atkinson, Jr., D., Whitworth, J.B. and Gage, S.J., Computer-assisted analysis of gamma-ray spectra. Comput. Phys. Commun., 2 (1971) 40. Baeza, A., Del Rio, M., Miro C., Paniagua, J.M., Moreno, A. and Navarro, E., Radiocesium concentration in migratory birds wintering in Spain after the Chernobyl accident. Health Phys., 55 (1988) 863. Baeza, A., Del Rio, M., Mir6, C. and Paniagua, J., Radiological impact of the Almaraz nuclear power plant (Spain) during 1986 to 1989 on surrounding environment. J. Radioanal. Nucl. Chem., 152 (1991) 175. Baxter, M.S., et al., Marine radioactivity studies in the vicinity of sites with potential radionuclides releases IAEA-SM-339/131, p. 167, IAEA, Vienna, 1996. Bennett, B.G., Fallout Pu-239+240 in the dietmresults through 1975, USERDA Report HASL-306, pp. 115-125, July (1976). Block, G., A review of validations of dietary assessment methods. Am. J. Epidemiol., 115 (1982) 492. BOE, Reglamento sobre protecci6n sanitaria contra radiaciones ionizantes, Boletin Oficial des Estado, No. 241 del 8/10/82, (1982) Madrid. Bogen, D.C., Stable lead concentrations in New York city foodstuffs, USAEC Report HASL-246, pp. 45-52, January 1972. Bourgeois, Ch., Fourgoux, M. and Bergey, C., Areal gamma mapping system, an efficient and rapid means for characterising the impact of radioactive deposition on the environment, Report IAEA-SM.339/91, p. 49, 1996. Bundesamt for Energiewirtschaft (Switzerland), Der Unfall Chernobyl (in German, The Chernobyl Accident), Report HSK-AN- 1816, Oct., Wtirenlingen, Switzerland, 1986. Bundesamt ftir Energiewirtschaft (Switzerland), Verstrahlungstage in der Schweiz nach dem Unfall in Tschernobyl {in German, Contamination situation in Switzerland after the Accident at Chernobyl), Sept., Bern, Switzerland, 1986. Canadian Standards Association (AS): Guidelines for radiological monitoring of the environmentmA National Standard of Canada, Report CAN/CSA-N288.4-M90, November 1990, Ottawa, Canada. Canberra Industries Inc., Application software, Canberra Edition Seven, USA (1987). Cember, H., Introduction to Health Physics. Pergamon Press, New York, 1983, p. 407. Chao, J.-H. and Chung, C., Radionuclide monitoring in environmental water body using an in situ gamma probe. Appl. Radiat. Isot., 43 (1992) 475.
Monitoring Accidentally Released Radionuclides in the Environment
483
Clark, R.H., Principles for the establishment of intervention levels for the prediction of the public in the event of serious nuclear accidents, Emergency planning and preparedness for nuclear facilities (Proc. Symposium Rome, 1985), IAEA, Vienna, pp. 373-384, 1986. Commission of the European Communities (CEC), 1982. Radiological Protection: Results of Environmental Radioactivity Measurements in the Member States of the European Community for Air, Deposition, Water and Milk 1980, EUR 7639 DA/DE/EN/FR/IT/NL, Directorate "Health and Safety", Luxemburg. Cooper, E.L., Valkovic, V., Strachnov, V., Dekner, R. and Danesi, P.R., Results of the intercalibration study of laboratories involved in assessing the environmental consequences of the Chernobyl accident. Appl. Radiat. Isot., 43 (1992) 149. Cooper, E.L., Valkovic, V., Strachnov, V., Dekner, R. and Danesi, P.R., Intercalibration study of laboratories involved in assessing the environmental consequences of the Chernobyl accident. J. Environ. Radioactivity, 17 (1992) 129-145. Cooper, E.L., Valkovic, V., Strachnov, V., Dekner, R. and Danesi, P.R., Intercalibration study of laboratories involved in assessing the environmental consequences of the Chernobyl accident. J. Environ. Radioactivity, 17 (1992) 129. Currie, L.A., Limits for qualitative detection and quantitative determination. Anal. Chem., 40 (1968) 586-593. Currie, L.A., Lower limits of detection: Definition and elaboration of a proposed position. Anal. Chem., 40 (1968) 586-593. Dabrowska, M., Jaracz, P., Jastrzebski, J., Kaczanowski, J., Mirowski, S., Osuch, S., Piasecki, E., Pienkowski, L., Szeflinska, G., Sheflinski, Z., Tropilo, J. and Wilhelmi, Z., Proc. of Int. Workshop on Hot Particles from the Chernobyl Fallout, October 28-29, 1987, Teuern, Germany. David, H.A., Biometrika, 43, Parts 3 and 4 (1956) 449-451. Department of Energy, United States, Health and environmental consequences of the Chernobyl nuclear power plant accident. DOE/ER-0332 (1987). Devell, L., Tovedal, H., Bergstr6m, U., Appelgren, A., Chyssler, J. and Anderson, L., Nature, London, 321 (1986) 192. Devell, L., Tovedal, H., Bergstr6m, U. et al., Initial observations of fallout from the reactor accident at Chernobyl. Nature, 321 (1986) 192-193. Dixon, W.J. and Massey, J.F., Introduction to Statistical Analyses. McGraw-Hill, New York, 1951. Dux, J.P., Handbook of Quality Assurance for the Analytical Chemistry Laboratory. Van Nostrand Reinhold, 1986. EML Procedures Manual (Eds. N.A. Chieco, D.C. Bogen, E.O. Knutson), 27th Edition, Volume I, Report HASL-300, New York, Nov. 1990. Erdtmann, G. and Soyka, W., The Gamma Rays of the Radionuclides: Tables for Applied Gamma Ray Spectrometry. Verlag Chemie, New York, 1979. FAO (Food and Agricultural Organisation), Food Balance Sheets. FAO, Rome (1984). Fisenne, I.M. and Keller, H.W., Radium-226 in the diet of two U.S. cities, USAEC Report HASL-224, pp. 2-8, April (1970). Fisenne, I.M., Perry, P.M., Decker, K.M. and Keller, H.W., The daily intake of 234'235'238pu,228'23~ and 226'226Ra by New York city residents. Health Physics, 53 (1987) 357-364. Fry, F.A., Radiological criteria for the protection of the public from radionuclides in the environment, Report IAEA-SM-339/32, p. 18, 1996. Garfield, F.M., Quality Assurance Principles for Analytical Laboratories. Association of Official Analytical Chemists, 1984. Gilbert, R.O., Statistical Methods for Environmental Pollution Monitoring. Van Nostrand Reinhold, 1987. Gonzalez, A., Chernobyl--Ten years after. IAEA Bulletin, 3 (1996) 2. Gray, P.W. and Ahmad, A., Linear classes of Ge(Li) detector efficiency functions. Nucl. Instrum. Methods, A237 (1985) 577. Greenberg, A. et al. Standard Methods for the Examination of Water and Waste-Water, 15th Edition, Section 703 (1981). American Public Health Assn./Water Pollution Control Federation Washington, D.C.
484
Chapter 9
Harley, J.H. (ed.), EML procedures manual, HASL-300, Environmental Measurements Laboratory, U.S. Department of Energy, New York, 1972. Hartell, J.K., Detection Limits for Radioanalytical Counting Techniques, ARH-SA-215 Atlantic Richfield Handford Co., Richland, WA, 1975. Health Physics Society Committee Report, Upgrading environmental radiation data, HPSR-1, United States Environmental Protection Agency, Washington, DC, 1980. Helmer, R.G. and McCullagh, C.M., Gauss VII, a computer program for the analysis of gamma-ray spectra from Ge semiconductor spectrometers. Nucl. Instrum. Methods, 206 (1983) 477. Hensley, W.K., Lepel, E.A., Yuly, M.E. and Abel, K.H., Adaptation and implementation of the Raygun gamma-ray analysis code on the IBM PV. J. Radioanal. Nucl. Chem., 124 (1988) 481. IAEA, Intervention Criteria in a Nuclear or Radiation Emergency, Safety Series 109, 1996. ICRP, 1990 Recommendations of the International Commission on Radiological Protection. Annals of the ICRP, vol. 21, no. 1-3, 1991. ICRP, Protection against radon-222 at home and at work. Annals of the ICRP, vol. 23, no. 2, 1993. Imai, K. et al., SPEEDI: A computer code system for the real-time prediction of radiation dose to the public due to an accidental release. JAERI 1297, 1985. International Atomic Energy Agency, "Atmospheric Dispersion in Nuclear Power Plant Site", Safety series No. 50-SG-$3, 1980. International Atomic Energy Agency, "Manual on Environmental Monitoring in Normal Operation", Safety series No. 16, 1966. International Atomic Energy Agency, "Measurement of Radionuclides in Food and the Environment, A Guidebook", Technical Reports Series No. 295, 1989. International Atomic Energy Agency, "Objectives and Design of Environmental Monitoring Programmes for Radioactive Contamination", Safety series No. 41, 1975. International Atomic Energy Agency. Summary Report on the Post-Accident Review Meeting on the Chernobyl Accident. Safety Series No. 75-INSAG-1. IAEA, Vienna, 1986. International Atomic Energy Agency: Technical Reports Series no. 295 Measurement of radionuclides in food and the environment, a Guidebook, IAEA, Vienna, 1986. International Commission on Radiological Protection, "Principles of Monitoring for the Radiation Protection of the Population, ICRP publication 43, 1985. International Commission on Radiological Protection. Recommendations of the International Commission on Radiological Protection (adopted 17 Jan. 1977). ICRP Publication 26. Annals of the ICRP, Sowby, F.D., ed., Oxford, Pergamon Press, 1979. Iwashima, K. and Ohkubo, T., Food Sanitation Research, 37(7) (1987) 7. Izrael, Yu.A. and Petrov, V.N., Severov, D.A., Radioactive fallout simulation in the close-in area of the Chernobyl nuclear power plant. Soviet J. Meteorol. Hydrol., 7 (1987). Japan Nuclear Safety Commission, Guide for Environmental Radiation Monitoring, 1978. Japan Nuclear Safety Commission, Guide for Methods of Evaluation Compliance with Dose Objectives around the Site of LWR Plants", 1976. Jaworowski, Z. and Kownacka, L., Tropospheric and stratospheric distributions of radioactive iodine and caesium after the Chernobyl accident. J. Environ. Radioact., 6 (1988) 145-150. JEN, Vigilancia Radiol6gica Ambiental para Centrales Nucleares de Potencia. Junta de Energia Nuclear, Madrid, (1978). Joint FAO/WHO Food Standards Programme, Codex Alimentaries Commission, CAC/Vol. XVII, Ed. 1, Supplement 1, Rome (1989). Killian, E.W. and Hartwell, J.K., VAXGAP: A code for the routine analysis of gamma pulse-height spectra on a VAXY computer, lEE Trans. Nucl. Sci., 36 (1989) 615. Klusek, C.S., Strontium-90 in the U.S. Diet, 1982, USDOE Report EML-429, July (1984). Koskelo, M.J., Aarnio, P.A. and Routti, J.T., Sampo80: an accurate gamma spectrum analysis method for minicomputers. Nucl. Instrum. Methods, 190 (1981 ) 89. Kramer, L., Spencer, H. and Hardy, E.P., Dietary Strontium-90 intake in Chicago. Health Phys., 25 (1973) 445-448.
Monitoring AccidentalLy Released Radionuclides in the Environment
485
Kurttunen, J.O., Cesium-137 in various Chicago foods, USDOE Report EML-405, Part III, pp. 3-6, May (1982). Le Corre, A. and Bourcier, T., Environmental measurements around French nuclear power plants, IAEA-SM-339/9P, p. 115, 1996. Lepp~inen, A. and Niskala, P., Radionuclides in ground-level air, Finnish Centre for Radiation and Nuclear Safety (STUK), Report No. 40, January, 1994. Loshchilov, N.A., Kashparov, V.A., Yudin, E.B., Protsak, V.P., Zhurba, M.A., Ivanov, Y.A. and Parshakov, A.E., in Proceedings of the Radiobiological Impact of Hot Beta-Particles from the Chernobyl Fallout: Risk Assessment, IAEA Coordinated Research Programme, Kiev, Ukraine, 1991, IAEA Report No. J 1-RC-487, IAEA, Vienna, 1992, p. 5. Marr, J.W., Individual dietary surveys: Purposes and methods, World Rev. Nutr. Diet., 13 (1971) 105. Metson, P., Analyst, 94, (1969) 1122-1129 and references therein. Milles-Lacroix, J.-C., Bourlat, Y. and Masnier, R., Airborne radioactivity measurements in MontlhEry (France), Presented at Second International Meeting on Low-Level Air Radioactivity Monitoring, 14-18 February 1994, Madralin (near Warsaw), Poland. Morgan, K.J., Johnson, S.R., Rizek, R.L., Reese, R. and Stampley, G.L., Collection of food intake data: An evaluation of methods. J. Am. Dietetic Assoc., 87 (1987) 888-896. Najafi, S.I. and Fedoroff, M., Accurate gamma ray spectrum analysis. J. Radioanal. Nucl. Chem., 89 (1981) 143. National Council on Radiation Protection and Measurements, A Handbook of radioactivity measurement procedures, NCRP Report No. 58, Second Edition, 1985. NRPB, 1990 Recommendations of the International Commission on Radiological Protection, Recommendations for the practical application of the board' s statement. Documents of the NRPB, vol. 4, no. 1, 1993. NRPB, Emergency reference levels of dose for early countermeasures to protect the public. Documents of the NRPE, vol. 1, no 4, 1990. Osuch, S., Dabrowska, M., Jaracz, P., Jastrzebski, J., Kaczanowski, J., Le Van Khoi, Mirowski, S., Piasecki, E., Pienkowski, L., Szeflinska, G., Szeflinski, Z., Tropilo, J. and Wilhelmi, Z., Isotopic composition of high activity particles released in the Chernobyl fallout. Health Phys., 57 (5) (1989) 707. Petrow, H.S., Schiessle, W.J. and Clover, A.: Dietary intake of radium-228, USAEC Report NYO-3086-1, pp. 1-10 (1965). Piasecki, E., Jaracz, P., Mirowski, S., (Part I) and Jaracz, P., Piasecki E., Mirowski, S. and Wilhelmi, Z. (Part II). J. Radioanal. Nucl. Chem., 141 (1990) 221-259. Piasecki, P., Jaracz, S., Mirowski, S., Jaracz, P., Piasecki, E., Mirowski, S. and Wilhelmi, Z. (Part II), Analysis of It-radioactivity of"hot particles" released after the Chernobyl accident. J. Radioanal. Nucl. Chem. Articles, 141 (2) (1990) 221 & 243. Raunemaa, R., Saari, H., Luokkanen, S., Lehtinen, S., in: H. von Philipsborn and F. Steinh~iuser (eds.), Hot Particles from the Chernobyl Fallout. Bergbau- und Industriemuseum, Theuern, 1988, Vol. 16, p. 77. Report to the General Assembly: sources, effects and risks of ionizing radiation, United Nations Scientific Committee of the Effects of Atomic Radiation, United Nations, New York, 1988. Rivera, J., Cesium-137 in tri-city diets--Results for 1965, USAEC Report HASL-174, p. 29-36, January (1967). Rivera, J., Stable strontium in tri-city diets, USAEC Report HASL-131, p. 230, October (1962). Ruckdeschel, F.R., Basic Scientific Subroutines. McGraw-Hill, New York, 1981. Salbu, B., In: H. von Philipsborn and F. Steinh~iuser (eds.), Hot Particles from the Chernobyl Fallout. Bergbau- und Industrie-museum, Theuern, 1988, Vol. 16, p. 83. Salbu, B., Krekling, T., Oughton, D.H., Ostby, G., Kashparov, V.A., Brand, T.L. and Day, J.P., Hot particles in accidental releases from Chernobyl and Windscale nuclear installations. Analyst, 119 (1994) 125. Sanderson, C.S., Latner, N. and Larsen, R.J., Environmental gamma-ray spectroscopy at remote sites with satellite data transmission. Nucl. Instr. Meth. Phys. Res., A339 (1994) 271. Schtinhofer, F., Ecker, W., Hojesky, H., Junger, W., Kienzl, K., Nowak, H., Riss, A., Vychytil, P. and Zechner, J., Tschernobyl und die Folgen ftir Osterreoch [Chernobyl and its Impact on Austria], Ministry of Health and Environment Protection, Vienna, Austria, 1986.
486
Chapter 9
Schotzig, V. and Debertin, K., Photon emission probabilities per decay of 226Raand 232Thin equilibrium with their daughter products. Int. J. Appl. Radiat. Isot., 34 (1983) 533. Schubert, P. and Behrend, U., Radiochim. Acta, 41 (19876) 175. Shkvorets, O., Akinfiev, G., Garger, E. and Girij, V., Use of aerial surveying for the detail mapping of radiation contaminated area around Chernobyl NPP, Report IAEA-SM-339/27P, p. 141, 1996. Spencer, H., Kramer, L., Samachson, J., Hardy, E.P. and Rivera, J., 90SrmCalcium interrelationships in man. Health Phys., 24 (1973) 525-534. SSI (Swedish National Institute of Radiation Protection), Activities of the Swedish Authorities following the Fallout from the Soviet Chernobyl Reactor Accident, Report 1986-05-12, May, Stockholm, Sweden, 1986a. SSI, Chernobyl--Its Impact on Sweden, Report 86-12, August, Stockholm, Sweden, 1986b. Standards methods for examination of water and wastewater, 16th Edition, Published jointly by American Public Health Association, American Water Works Association, Water Pollution Control Federation, A.P.H.A., 1985. Statens Haverikommission, Beredskap after Tjernobyl (in Swedish, Preparedness after Chernobyl), Oct., Stockholm, Sweden, 1986. Stroube, W.B. Jr. and Jelinek, C.F., Survey of radionuclides in foods, 1978-1982. Health Phys., 49 (1985) 731-735. STUK (Finnish Centre for Radiation and Nuclear Safety), Interim Report on Fallout Situation in Finland from April 26 to May 4, Report STUK-B-VAL044, May, Helsinki, Finland, 1986a. STUK, Second Interim Report on Radiation Situation in Finland from 5 to 16 May, Report STUK-B-VAL045, May, Helsinki, Finland, 1986b. Suzuki, T., Inokoshi, Y., Chisaka, H. and Nakamura, T., Optimum geometry of large Marinelli-type vessels for in situ environmental sample measurements with Ge(Li) detectors. Appl. Radiat. Isot., 39 (1988) 253. Szekely, G., FGM-A Flexible Gamma-spectrum analyses program for small computer. Comput. Phys. Commun., 34 (1985) 313. Tagaki, S., Ohfou, Y., Nakaoka, A., Inove, T., Kaube, H., Fukushima, M., Sakagishi, K. and Koyama, M., Determination of environmental radioactivity (Part 5). Determination of Fallout Nuclides. Central Research Institute of Electric Power Industry, Tokyo, 1978. Tchrkezian, V., Shkine, V., Khitrov, L. and Kolesov, G., Experimental approach to Chernobyl hot particles. J. Environ. Radioactivity, 22 (1994) 127. Technical Report: The International Chernobyl Project, Report by an International Advisory Committee, Printed by IAEA, Vienna, 1991. Ter-Saakov, A.A., Glebov, M.V. and Gordeev, S.K., Working Materials to The 1st IAEA Co-ordinate Meeting on The Radiological Impact of Hot Beta Particles from the Chernobyl Fallout: Risk Assessment, August 26-30, 1991, Kiev, Ukraine, CIS. U.S. Department of Agriculture, Food consumption: Households in the United States, Spring 1977, Human nutrition information service NSCS 1977-78, Report H-1 September (1982) [regional tabulations are available for the Northeast (H-2), North Central (H-3), South (H-4), and West (H-5)]. U.S. Department of Agriculture, Food intakes: Individuals in 48 states, years, 1977-78, Human Nutrition Information Service NSCS 1977-78, Report I-1, august (1983). U.S. Environmental Protection Agency, California diet study. Radiation Data and Reports, 14 (Feb. 1973) 69-73. U.S. Environmental Protection Agency, Radionuclides in institutional total diet samples. Radiation Data and Reports, 15 (Feb. 1974) 126-128. United Nations Ionizing Radiation: Sources and Biological Effects. United Nations Scientific Committee on the Effects of Atomic Radiation 1982 Report to the General Assembly, with annexes. United Nations publication E.82.IX.8 New York, 1982. United Nations Scientific Committee on the Effects of Atomic Radiations, Sources, effects and risks of ionising radiation. United Nations, New York, 1988. Valkovic, V., Zeisler, R., Berasconi, G. and Danesi, P.R., Reference materials for micro-analytical nuclear techniques. Int. J. PIXE, 2 (1992) 651-664.
Monitoring Accidentally Released Radionuclides in the Environment
487
Vinjamuri, K., et al., A review of fission product behaviour during past accidents and destructive tests, EGG-TFBP-6026, 1982. Welford, G.A. and Baird, R., Uranium levels in human diet and biological materials. Health Phys., 13 (1967) 1321-1324. World Health Organisation, Guideline for the study of dietary intakes of chemical contaminants, Joint FAO/WHO Food Contamination Monitoring Programme Report WHO-EFP/83.53, WHO Offset Publication No. 87, World Health Organisation, Geneva (1986). Zagyguai, P., Parr, R.M. and Nagy, L.G., Additional results for the G- 1 IAEA intercomparison of methods for processing Ge(Li) gamma-ray spectra. J. Radioanal. Nucl. Chem., 89 (1985) 589.
489
CHAPTER 10
Bomb Test Sites
10.1 INTRODUCTION Since the explosion of the first atomic bombs in 1945, many countries have pursued the goal of becoming a nuclear power. Only a few have succeeded in this; and then it cost them large amounts of their GNP. The development of a nuclear arsenal required a comprehensive series of tests. The pattern of nuclear testing by countries has been to conduct atmospheric tests, some of which were of relatively high total yield, followed by a series of more numerous underground tests. As containment of debris is desired in an underground test, the yields of these tests have been generally much lower than atmospheric tests. The largest atmospheric test was 50 Mt in total conducted by the former Soviet Union in 1961, and there were 24 further atmospheric tests with yields from 4 to 25 Mt. There were a few underground tests of relatively high yield; the largest was in the range 1.5-10 Mt conducted by Russia in 1973, and an underground test of 5 Mt yield was conducted at Amchitka, Alaska in the United States in 1971. However, most of the underground tests have been of much lower yield. Tests of nuclear weapons in the atmosphere were conducted by five countries during the period 1945-1980. The most active test period was between 1952 and 1962, when many tests were conducted by the United States and the former Soviet Union and a limited testing programme was carried out by the United Kingdom. Atmospheric testing by France occurred from 1960 through 1974 and by China from 1964 through 1980. No further atmospheric tests have taken place since 1980. The number of nuclear tests conducted by all countries is shown in Fig. 10.1. The numbers of atmospheric tests each year are indicated by the scale above the horizontal axis and the number of underground test by the scale below the axis. Atmospheric tests were conducted primarily during the 1950s and in 1961 and 1962. In 1958, 1961 and 1962, there were 50 or more tests per year. Continued tests, but fewer in number, were conducted by China and France from 1960 until 1980. In all, there were 541 atmospheric tests. This compares with over 1800 underground tests~over three times as many as atmospheric tests. Underground testing began mainly after the limited
Chapter 10
490
150
100
50
_"a r ' ~
IST]
"1 50
100 1950
1960
1970
1980
1990
Fig. 10.1. Number of nuclear tests.
nuclear test ban treaty was signed by most countries in 1963. There was relatively constant underground testing, often of 50-100 tests per year, until a comprehensive test ban treaty was formulated in 1996. This treaty has not yet been ratified by all countries, and in fact, 11 further underground tests were conducted in 1998 by India and Pakistan. Table 10.1 summarises the information about nuclear tests performed by five nuclear powers (US, Russia, France, UK and China). The records of the annual yields from nuclear tests show a predominance of atmospheric tests. The total annual explosive yields of atmospheric tests were particularly high during 1961 and 1962. The total yield of all atmospheric tests was 440 Mt. The total yield of all underground tests was 90 Mt, just 20% of atmospheric total.
Table 10.1 Nuclear tests by five big nuclear powers
US
Number of tests
Date of first test
Date of last test
1030
16 July 1945
23 September 1992
USSR/Russia
715
29 August 1949
24 October 1990
France
210
13 February 1060
27 January 1996
UK
45
3 October 1952
26 November 1991
China
45
16 October 1964
29 July 1996
491
Bomb Test Sites
10.1.1 Atmospheric testing When a nuclear weapon is tested in the atmosphere, the large amount of radioactive debris produced in the explosion is freely released into the environment. This radioactive debris, consisting of gases and particulate radionuclides, disperses with atmospheric circulation and is transported and deposited throughout the world. People everywhere are then exposed to radiation from radionuclides in the air and on the ground and also from radionuclides that enter the body by inhalation of air and by ingestion of food and water. Since atmospheric testing of nuclear weapons has finished, since the radionuclides have dispersed and since the radiation doses that resulted will only continue to decrease as the longer-lived radionuclides decay, the legacy of atmospheric testing is of interest mainly from a retrospective standpoint. The present and future health implications of atmospheric nuclear testing can only be much less than they were at the time the tests were conducted. Still, it is of interest to review the exposures from this practice to add perspective to evaluation of exposures from underground testing or from other sources. Atmospheric testing has also resulted in widespread levels of longer-lived radionuclides in the environment, especially 137Cs,9~ and also 3H, a n d these form a baseline on which possible future releases of radionuclides will be added. Many measurements were made throughout the world during the period of atmospheric nuclear testing, and much is known of the release, dispersion, and deposition of radionuclides and the doses resulting from this practice. Exposures of the world population have been evaluated by the United Nations Scientific Committee on the Effects of Atomic Radiation. From this information the deposition and doses from individual tests or from a specific test series may be inferred. A summary of atmospheric testing by the countries that conducted the tests is given in Table 10.2. There were 541 atmospheric tests of total explosive yield 440 Mt. Depending on its explosive yield, a nuclear test may introduce radioactive materials to various heights in the atmosphere. The lowest level of the atmosphere is the troposphere, in which turbulent air movements occur. In addition to prevailing horizontal winds, there is also considerable vertical motion as evidenced by clouds, rain and Table 10.2 Atmospheric tests Country
Number
Yield (Mt)
USSR
219
247
USA
217
154
China
33
21
France
50
10
UK Total
22
8
541
440
492
Chapter 10
thunderstorms. Above the troposphere is the stratosphere, consisting of more stable air layers. The region separating the troposphere and stratosphere is the tropopause. The height of the tropopause is different in the polar and equatorial regions. Heights of 9 km in the polar regions and 17 km in the equatorial regions are typical. Gaps in the mid-latitude regions facilitate the downward transfer of air and particles from the stratosphere to the troposphere. A nuclear test results in a characteristic mushroom cloud formation that carries into the atmosphere the debris formed in the explosion, as well as soil or water particles from the earth's surface, if the explosion takes place near ground level. The height of the cloud formation and thus the partitioning of the debris injected at various altitudes are determined by the total explosive yield of the test and the height of the burst. The cloud formation height also depends on the latitude. For the same total explosive yield, the cloud rises to greater heights in the equatorial region than in the polar region. An example of the partitioning of radioactive debris in atmosphere regions following a nuclear test is shown in Fig. 10.2. for a 1 Mt explosion at an equatorial latitude, the cap of the mushroom cloud stabilises at altitudes between 12 and 20 km. Radioactive debris is located throughout the cloud but mainly in the central to lower central regions. With the tropopause taken to be at 17 km altitude, the partitioning of the radioactive debris is estimated to be 65% entering the troposphere and 35% entering the stratosphere. For the same 1 Mt explosion taking place at polar latitudes, the cloud cap stabilises between 8 and 16 km altitude (Fig. 10.3). With the tropopause at 9 km, the partitioning of radioactive debris is estimated to be 1% to the troposphere and 99% to the stratosphere. The partitioning of debris between atmospheric regions following tests in the equatorial and polar regions has been determined for the full range of yields of nuclear tests based on measurements made at the time of the tests. Low-yield tests introduce material primarily into the troposphere, and higher-yield tests introduce material in greater proportions into the stratosphere.
Fig. 10.2. Partitioning of radioactive debris for a 1 Mt explosion in equatorial region.
Bomb Test Sites
493
Fig. 10.3. Partitioning of radioactive debris for a 1 Mt explosion in polar region.
Following injection of radioactive debris into the atmosphere, the subsequent dispersion and eventual deposition of the material onto the earth's surface are determined by mass air circulation patterns in the atmosphere. These patterns have been established from meteorological studies and from measurements of the behaviour of fallout radionuclides. The injection of radionuclides into the atmosphere by atmospheric weapons testing has provided a unique tracer experiment, which has helped to provide understanding of the atmospheric processes. Debris injected into the troposphere circles the earth within one to two weeks, and particles are removed preferentially by rainfall but also by dry deposition onto the ground or water surface. The deposition occurs continuously and mostly within the latitude band of injection. Particles remain suspended in the troposphere for time periods of a few weeks. For high yield tests, with cloud formation extending into the stratosphere, debris may be suspended for time periods of one year or more. Atmospheric dispersion and interhemispheric exchange result in widespread global dispersal and deposition of the radionuclides produced in the test. The predominant air movements in the atmosphere are governed by pressure differentials in the troposphere, causing the well-known local and regional weather phenomena. There are also the west to east latitudinally flowing jet streams of air in the upper troposphere and lower stratosphere caused by the earth' s rotation. Radionuclides injected into the atmosphere attach to ambient aerosol particles, and they are then dispersed according to these airflow patterns. Particles injected into the thin air of the high atmosphere or upper stratosphere descend by gravitational settling. In the lower stratosphere random motion of air, called eddy diffusion, is important. In the troposphere and extending into the lowest parts of the stratosphere is the Hadley circulation. The Hadley cells are large circular airflows that shift in position and grow or decrease in size through the course of the year. There is little or no interhemispheric exchange in the upper troposphere during these times. During the northern hemispheric winter
494
Chapter 10
season, the equatorial Hadley cell expands greatly and spreads across the Equator. The predominant flow in the lower stratosphere in that season is toward the winter (northern) hemisphere, and interhemispheric exchange takes place. The predominant downward flow of air in the winter and spring seasons is in the mid-latitude region. The Hadley cells return to the balanced pattern in the spring and fall season, and then the opposite hemispheric cell develops into its dominant position during its winter season. The residence times of particles in the various atmospheric regions have been determined by numerous measurements of the tracer radionuclides in air and deposition. The intercompartment transfer halt-times are seasonally dependent, ranging from two years for particles to move from the high atmosphere to the upper stratosphere, six to nine months from the upper stratosphere to the lower stratosphere and for further movement to the troposphere and one month in the troposphere before deposition onto the earth's surface. The interhemispheric exchanges in the lower stratosphere and upper troposphere take place with residence half times ranging from one to two years, depending on the season. An empirical model has been formulated to describe these seasonally dependent transfers that take place in the atmosphere and that govern the movement and time dependence of suspended particles in air until deposition onto the ground or ocean surface. The partitioning of radioactive debris in the troposphere and stratospheric regions is determined by the total explosive yield and the height and latitude of the burst. The total yield is the sum of the fission and fusion yields of the device. The production of important fallout radionuclides is determined by the fission yield of the weapon. Smaller yield nuclear explosions are produced by fission reactions, while larger yield explosions result from boosted fission or thermonuclear fusion reactions. Of the total yield of all atmospheric tests of 440 Mt, an estimated 182 Mt, or about 40% of the total, was fission yield and the remainder was fusion yield. The contributions of countries conducting atmospheric tests to the total fission yield is shown in Table 10.3. Not all of the radioactive debris produced in nuclear tests is carried into the troposphere and stratosphere and dispersed as global fallout. For tests conducted on the ground or water surface, an estimated 50% of the debris remains in the local vicinity of the test site. Many tests conducted by the United States were surface explosions. Other Table 10.3 Fission yields of atmospheric tests Country
Fission yield (Mt)
USSR USA China France UK Total
88 72 11 6.6 4.4 182
(48%) (40%) (6%) (4%) (2%)
495
Bomb Test Sites
Table 10.4 Global injection Country
Fission yield (Mt)
USSR
88
(56%)
USA
49
(31%)
China
11
(7%)
France
6.3
(4%)
UK
4.3
(3%)
Total
158
Table 10.5 Hemispheric partitioning of global injection Region
Fission yield (Mt) North
South
Troposphere
14
4
Stratosphere
126
14
Total
140
18
tests and most tests of other countries were airbursts, with the devices carried aloft by balloons, aircraft or rockets. The total fission yields without the local fallout components are shown in Table 10.4. Airbursts greatly minimise local fallout production. The total fission yield of explosions representing injections of debris into the atmosphere (that is, the total, excluding local fallout) was 158 Mt. This is the total fission yield related to global fallout production. Most of the test sites for atmospheric nuclear testing were located in the northern hemisphere. The tests of France in the Pacific and of the United Kingdom in Australia were, with few exceptions, the only ones conducted in the southern hemisphere. Several tests of the United States were at or very near the equator, and from that location the injection of debris occurred to both hemispheres. The hemispheric partitioning of fission yields is shown in Table 10.5:140 Mt was injected into the atmosphere of the northern hemisphere, mostly into the stratosphere, and 18 Mt was injected into the atmosphere of the southern hemisphere. The relationship between the amounts injected into the atmosphere and the amounts ultimately deposited on the earth's surface depends on the location of injection and the air circulation patterns as described in the atmospheric model. Materials entering the troposphere circle the earth within about 10 days and are deposited within a time period of up to one month largely within the latitude band of injection. There is no interhemispheric exchange of note. For the injection of material from past nuclear tests, the
496
Chapter 10
atmospheric model specifies that, on average, 82% of the stratospheric material is deposited in the hemisphere of injection and 18% is deposited in the opposite hemisphere. There is a delay of one to two years before deposition of stratospherically injected materials. Environmental and human consequences of atmospheric nuclear tests have been recently summarised, see Shapiro (1998).
10.1.2 Underground testing The testing of a nuclear weapon in the atmosphere results in the immediate release of radioactive materials, allowing relative rapid and widespread dispersion in air of the materials produced. This contrasts with an underground nuclear test in which the radioactive debris is confined by design to the underground cavity. If the underground test has been done properly, there is no release or venting of gases or particles to the atmosphere, and there will be only very slow migration, if any, of radionuclides to the surrounding media. Following an underground test, immediate exposures are usually absent, and one only need consider the slow, long-term migration of radionuclides that could contribute to radiation exposure of the regional population. In terms of safety, the main advantage of underground testing is the containment capacity offered by the solid environment with respect to irradiation and contamination risks. The risk of irradiation is removed naturally by the thickness of the geological formations around the explosion. The risk of contamination, which can result only from gas or liquid leakage occurring immediately after the test or by long-term migration through the rock, is avoided if the ground and the stemming is sufficiently impermeable. Containment is thus guaranteed by the properties of the geological environment and of the materials around the source. To describe underground testing in some detail, we shall follow the report of Bouchez and Lecomte (1996) which describes the French nuclear testing in the Atolls of Mururoa and Fangataufa (French Polynesia). The measurements relating to the operation of the device were made in the immediate proximity of the explosion point. All the measuring instruments were housed in a container, the lower part of which contains the nuclear device. Before the shock destroys the measuring instruments, the radiation generated by the device is converted in the container into light signals which are carried by optical fibres to the analysis and recording devices located in the surface. For the under-lagoon tests, the cabins containing the recording devices are placed on a "Measurement Recording Barge" (B.E.M.) moored a few hundred metres from the ground-zero (point on the surface of the ground or the lagoon located vertically above the explosion). The optical fibres also carry the remote-control and telemetry information for the container and the device. The links between the container and the surface cabins also include the container power supply cables. The signals are digitised in the recording cabins. The digital data, like the remotecontrol and telemetry data, is transmitted by radio link to the personnel base, over a
Bomb Test Sites
497
distance that ranges from 10 to 20 km for a test on Mururoa up to 40 km for a test on Fangataufa. The purpose of the measurements made during a nuclear test is to characterise the source of gamma, X-ray, visible or neutron radiation. They also provide fundamental data in the field of thermonuclear fission and fusion plasma physics, useful for the evolution of weapon design. These measurements are difficult to carry out because of the inherent nature of the phenomena studied, the precision required and the experimental conditions. Sensors or detectors must be designed with wide dynamic range (the measurements must cover a range of 10 to 20 orders of magnitude), capable of analysing the radiation emitted by the device in less than a millisecond, with a resolution of the order of a nanosecond. Beyond a millisecond after the zero time, the temperature (one thousand million degrees) and the pressure (10 megabar) generated by the passage of the shock wave cause the total destruction of the container. These sensors must operate reliably, despite highly-disturbing background noise and interference phenomena due to the nuclear environment. Finally, they must occupy as little space as possible and be wellintegrated into the container, while withstanding the temperature and vibration stresses relating to transport and then lowering into the shaft. Given the low volume available inside the containers, there is little possibility of redundancy, so the detectors must be reliable and robust (and of small dimensions, of course). This reliability is also made necessary by the limited number of tests possible. The detectors must operate for as long as possible, and must consequently withstand the temperature, which rises very rapidly. Moreover, the response of the sensors as a function of temperature must be known very accurately. The equipment placed in the container is aligned very precisely, and this alignment must not be disturbed by transport. Selective protection is provided so that each sensor receives radiation only from the selected source. All the detectors used are usually developed specifically for the needs of nuclear testing. In the French test gamma radiation was analysed by Compton and Cerenkov detectors. X-ray spectrometry required photoelectric detectors and optical axes under vacuum. Neutrons were analysed by optical sensors and reverse proton telescopes. Observation of the geometry of the source and its deformation with time has led to the development of imaging measurements. The pin-diaphragm technique (pin-hole in an absorbent medium) has been used to convert photons, X-rays, gamma rays and neutrons into visible-light images. A digital camera records the image by means of a shutter with an exposure time of a few billionths of a second, and the data must be transmitted before the shock wave arrives. These measurements are particularly difficult. Apart from the instrumentation installed in the containers, advanced technology has also been developed and used for manufacturing the containers themselves, and in the field of drilling (large diameters, verticality and calibration, trajectory monitoring). The containers, cylindrical in shape are lowered to depths of between 500 and 1100 m is shafts filled with water. Their steel casing is therefore designed to remain watertight at pressures greater than a hundred bar. It must also withstand any handling
498
Chapter 10
shocks, vibrations during transport, scraping along the walls of the shaft, etc. The structure of a container must provide high rigidity, in order to maintain the perfect positioning of the various elements and the alignment of the optical axes from the time when they are adjusted at ground level to the time of the test. Depending on the instrumentation that they house, the lengths of the containers can vary from 10 to 25 m with a diameter of 1.3 m, and their weight can reach 70 tonnes. During the French nuclear testing the various elements to the container, excluding the nuclear device, were assembled in an assembly hall located on land on Mururuoa. The container was then transported by road on a trailer, in the horizontal position, to the test site if the test took place on land, or to the port area if the test took place under the Mururoa or Fangataufa lagoon. In the second case, the container was loaded on a barge which was towed to ground zero. Once the container was close to the test shaft, the measurement and remote-control cables were connected and the nuclear device was fitted to the base of the container. The container was then placed in the vertical position, using a gantry for land-based tests or the handling barge derrick in the lagoon. Finally, the container was lowered to the bottom of the shaft by means of a winch and a steel carrying cable, and the cables were connected to the measurement and recording cabins. Tests were carried out at various stages of these operations to check the correct operation of the equipment installed in the container and the quality of the connections (see Fig. 10.4). The large-diameter shafts (1.52 m) are drilled using a bit fitted with steel or tungsten-toothed wheels, driven from the surface by a drill pipe string. The rock is ground by the drill bit and the cuttings are removed by the "air fit" technique. Air under pressure is injected into the drill pipe string by low-diameter tubing lowered to a depth of about a hundred metres. The fluid column in the drill pipe string is lightened, and the difference in density between the air-water mixture and the sea water generates a violent circulation from the annular space to the inside of the drill pipe string; the debris is thus sucked to the surface. The cuttings from the borehole are analysed as drilling advances, in order to check whether the local geology is compatible with the scheduled nuclear test. If necessary, the planned depth for the explosion point can be increased until a favourable environment is encountered. The large-diameter boreholes must satisfy certain criteria to avoid jamming of the containers during lowering. In particular, the slant must always be less than one degree at any point of a shaft, and the walls must not have any break greater than 20 cm in the cemented zones. The shafts must be sufficiently stable to avoid any risk of infall that might damage or jam the container. The choice of the exact place (zero point) where the test device and measuring instrument container is positioned is an important step in the preparation of a test. The difficulties of access to certain parts of the rim (flat submerged at high tide) or of the lagoon (shallow bottom, coral patches) are natural limitations to the possible test areas. In the case of French testing on Mururoa the desire to keep the stresses induced by the explosions below acceptable levels resulted in the imposition of a minimum
499
Bomb Test Sites
- " l" - .- L. A G O . O N.' - - . ~ 4. 0 m .........
~~/.~... ~ _ -
'=.'~
-..'
_
Recording----"-"----"barge .
,
.J_. C A R B O N A T E S _ r _ L @ @ I i i i I
_ yx
I
I
"i
_
.I.
I
~ 200 to
I
300m
I o:
. . . . . . . . . . . . . . . .
•
~
,
,
.
.
.
,
I
_ i ,_ _i , , _ ,
I
~
?'~
;" Cement and aggregate
• •
> ~, 100 to 200 m ~
X X , • 2 1 5VOLCANISM X • X • X X • X X • • X • X X • •
"
• :d :d X >(
xX
'
Measurement and _r_L remote-control cables
• • ••
.
>( ~ 500 to 1100 m ~
.xx'R <xV-xV, xV,x' 42~
~ "Effective" part of ;'- the stemming (cement)
X •X
• X • • Future collapse: • zone • x • ~ Cement and aggregate •X • • • • • • • Basalt sand ,, )< " " X Container X
~
x
x x
~.Zero poini ;< ~
~:22s
X • X X X
xx
Fig. 10.4. Overall diagram of the operational zone of the time of the test (not to scale). At the bottom of the shaft, the container is surrounded by basalt sand. Above, the shaft is filled with cement and aggregates up to the top of the future collapse zone. The top of this zone corresponds to the base of the effective part of the stemming column. The container is linked by a cable harness to the recording barge, which transmits the measurements to the personnel base.
distance between the tests and the personnel base, and the reservation of the West of the atoll for the higher-yield tests. To define the position of a planned test within this area, additional criteria must be applied. The first one concerns the requirement for a given test to be separated from tests carried out previously in the neighbourhood. For a given explosion yield, a minimum distance between shafts must be complied with to avoid intersection of the
500
Chapter 10
damaged zones around the test points. The chosen location must also satisfy the rules imposed by stress limitation: for example, it is ensured that the shocks caused by the explosion do not endanger any drilling in progress, but above all that they do not threaten the stability of the flanks of the atoll. The application of these criteria reduces the range of possible areas, and a neardefinitive choice of zero point can be made. Within the limits defined in this way, the exact position of the zero point is first chosen as a function of the quality of the volcanic rock in the planned area. Based on knowledge of the environment, the geologist must assess the a priori properties of the various geological formations. He also takes into account both the feasibility of drilling, and the necessity to ensure containment. In general, the depth of the test is adjusted to guarantee that the zones damaged by the explosion remain mainly contained in the basalt. Of course the rock around the future zero point must not have any discontinuity, fault or geological dislocation whose nature would favour leakage. While meeting the required impermeability criteria, it must offer satisfactory mechanical properties in order to guarantee the stability of the shaft during drilling. The first requirement corresponds more to argillized, that is, altered materials, while the second corresponds more to massive materials. Another factor is also involved: the carbonate content of the volcanic rocks surrounding the future zero point. Its value must remain less than a few per cent in order to avoid the formation of excessive quantities of carbon dioxide at the time of the test. The presence of such gases under pressure in the cavity created by the explosion would represent a danger when the sampling drill-holes made after the test enter the fractured zone. The basaltic rocks of Mururoa and Fangataufa offer two advantages with respect to the containment of the explosion products. First, their low permeability opposes the vertical migration of radioactive products. Second, under the effect of the heat generated by the explosion, the basalts melt to form glasses, which trap the majority of these products (approximately 98% of the elements with half-lives longer than 30 years are trapped in this way). The choice of the zero point is an essential element in the safety file of a nuclear test. However, this file covers many other aspects. It specifies the conditions for performing the test in conformity with the defined principles, in order to avoid accidents and ensure the radiological protection of the personnel and the environment. It must satisfy both the severe criteria for defining the explosion point and the many rules governing operational implementation. After examination and approval, this file becomes the reference document for the test in terms of safety. Any subsequent change necessitates a new examination. The stemming of the shaft is started after the container has been lowered to the specified depth. The stemming plugs the shaft and contributes to the containment of the explosion. First, the space around the container and the shaft to a height of approximately 20 rn is filled with basalt sand, which provides thermal protection against the heat given offby the setting of the cement. Then a cement ring is made, from 10 to 20 m high, which constitutes the seating for the stemming. The shaft is then
Bomb Test Sites
501
gradually plugged with consecutive layers of basalt aggregates or cuttings and cement up to the top of the future collapse zone. From this depth the shaft is filled with a homogeneous cement plug, which remains completely attached to the rock when the explosion takes place. This plug, 100 to 200 m high, in fact constitutes the effective part of the stemming. The composition of the cement, which is in the form of a slurry to facilitate its application, is continuously monitored. This slurry must conform to a precise specification: defined theoretical density, strength 48 hours after setting greater than a specified value, and temperature increase less than 80~ to avoid damaging the measurement cables. The density and the mechanical properties are tested on samples taken during injection. Above the plug, layers of cement or aggregates can be added, but they do not contribute to containment. The general dimensions of the stemming are based on knowledge of the stresses that will be generated by the planned test and on the geology of the zone concerned. The behaviour of the stemming-rock assembly thus defined is then checked by numerical simulations which take into account the parameters of the explosion and the characteristics of the cement and of the rocks between the zero point and the surface of the ground. These characteristics are determined by compression and impact tests performed at stresses of up to several tens of kilobars, comparable with those reached close to the explosion. In several tests, sensors placed along the height of the cement column have measured the acceleration applied to the stemming during the passage of the shock wave. The signals recorded enabled the simulations to be validated. The sequence of events around the zero point, starting from the ignition of the nuclear device, is divided into several phases that can be differentiated by the energy transfer mode and the reaction of the geological environment (Bouchez and Lecomte, 1996). The nuclear phase corresponds to the expansion of the high-temperature plasma resulting from the nuclear reaction within the explosion chamber. It does not differ from that accompanying an atmospheric explosion as long as there are no significant interactions with the environment outside the casing of the device, that is, up to a few tens of nanoseconds. Beyond this time, the materials of the container and the surrounding basalt absorb the radiation and, in an analogous manner to what happens in the air during atmospheric tests, a fireball is generated, which gives off a shock wave. The hydrodynamic phase starts when the transfers become essentially mechanical. It corresponds to the period during which, under the effect of the shock, the environment behaves as a fluid: the pressure, more than several hundred kilobar, then governs the phenomena. This phase, which lasts for about ten milliseconds, concerns a region limited to a few metres around the explosion, where the rock is vaporised, melted and crushed. While the shock wave propagates ahead, the rock gases at high pressure and temperature push back the surrounding rock, causing the formation of a cavity whose expansion is countered by the resistance of the environment and the pre-existing stresses in the massif. The elasto-plastic phase constitutes the transition to the seismic propagation mode. During this phase, the shock wave is gradually replaced by a wave with a more
502
Chapter 10
extended profile which is less attenuated with distance. The rock no longer behaves as a fluid. The region concerned extends to a few hundred metres from the explosion, and can be subdivided into several zones in which the wave propagation and the environment reaction have different characteristics. During this phase a network of fractures develops, which becomes less dense moving away from the explosion point and whose extension is limited by the resistance of the surrounding rock and the stresses due to gravity. The final phase groups the events occurring at more or less long times after the passage of the shock wave. The cavity, which reaches its maximum radius when the thrust of the gases is no longer sufficient to compact the rocks, begins to cool. Most of the vaporised materials condense and drain to the bottom of the cavity with the melted material, forming a lava meniscus. As the cooling continues, the pressure drops in the rocks and in the cavity, and the roof of the latter collapses. This collapse, which is sometimes not completed for several hours, causes the formation of a debris cone called the chimney, whose height is rapidly limited by a natural stabilisation process. On the periphery of this cavity-chimney assembly, the water originally present in the saturated geological layers of the massif has been pushed back with the rocks. The gradual cooling of the materials and the condensation of part of the gases means that the cavity-chimney assembly is at reduced pressure compared with the surrounding rocks, which are decompressing. This disequilibrium favours the drainage of the water present in these rocks towards the cavity. This centripetal drainage can continue for weeks, even months, after the test, modifying the local hydrogeological system until equilibrium is re-established. The phenomena observed in the near-source region are closely dependent on the test conditions. Depending on the depth of the zero point and the nature of the rocks surrounding it, an explosion may or may not cause the formation of a subsidence or ejection crater at the ground surface, with the risk of release of radioactive products. When no crater is formed, the explosion is said to be contained. All the underground tests carried out on Mururoa and Fangataufa belong to this category. In the natural environment, the propagation mode of a perturbation and the transformations that it causes in the medium areclosely dependent on its amplitude with respect to the initial equilibrium state. This observation is fully applicable to the near-spherical waves that propagate in the ground immediately after the nuclear phase of the explosion, that is, from the time at which the energy transfers in the medium become essentially mechanical and thermal. Depending on the distance from the explosion, the levels of stress, temperature and deformation rate attained cause very different reactions within the various geological layers. The dimensions of the volume in which the nuclear explosive is placed play an important role in the phenomenology of an underground explosion. At the French Pacific Test Centre, the volume of the part of the container that receives the device is approximately one cubic metre. This part of the container constitutes the test chamber. After the nuclear reaction, the energy is released in a few tens of nanoseconds. This time is too short for the material to acquire significant movement. The energy is therefore transmitted in the form of radiation to the walls of the test chamber.
Bomb Test Sites
503
This radiation is absorbed in the surrounding materials over very short distances, of the order of a few to ten centimetres. This results in pressures of several megabar to several hundred megabar, depending on the energy of the explosion and the dimensions of the initial chamber. Inside this chamber the matter is ionised and dissociated, so the mean free path of the photons is long. The temperatures are therefore rapidly equilibrated. It can be assumed that the energy is uniformly distributed in the volume of the chamber, which resembles a fireball. The pressures generated at the walls of the chamber produce an intense shock wave which propagates in the medium, transforming it. It is during the hydrodynamic phase, under the effect of the shock, that the geological material undergoes the greatest transformations: for a 1-kiloton explosion, half the total energy released in a radius of approximately 3 m is used to modify the material. A large part of the energy carried by the shock wave is dissipated irreversibly in the form of heat transferred to the medium. The pressures reached are so high that the materials behave as fluids. The heat transfer causes a very large temperature increase. This produces physicochemical transformations: melting, vaporisation, dissociation, ionisation, etc. It is generally accepted that, in the dissociation and ionisation processes, the recombination and relaxation times are very short with respect to the times necessary for the change of state of the material; the corresponding energies are restored sufficiently fast to produce mechanical work immediately reusable for the propagation of the shock. In other words, in the vaporised zone, the energy lost irreversibly in the form of heat is limited to the heat of vaporising of the medium. Beyond the vaporisation radius, the transferred heat can no longer be reused to feed the shock. Knowledge of the quantity of heat transferred to the medium enables assessment of the size of the zones vaporised and melted when the shock passes. From the energy balance and the characteristics of the medium surrounding the explosion, the vaporisation and melting pressures can be calculated. For the standard material representative of the basaltic rocks at the French Pacific Test Centre, the vaporisation pressure is approximately 1.5 megabar and the melting pressure approximately 0.5 megabar. The weight of rock vaporised and melted per kiloton can then be deduced, approximately 100 tons and 210 tons, respectively. These weights are equivalent to a vaporised zone radius of 2.1 m and a melted zone radius of 3.1 m. The total weight of rock vaporised and melted instantaneously as the shock passes is thus approximately 310 tons per kiloton, and the initial cavity thus formed attains a radius of more than 3 metres. The weight estimated above is significantly lower than the total quantity of rock melted by the explosion, because it is necessary to take into account other melting processes, known as secondary processes. Beyond the solid matrix melting radius, over a distance of a few metres, the interstitial water is vaporised. As it expands, the high-pressure steam shatters the matrix, which is divided into particulate fines. When the cavity later expands under the effect of the thrust of the gases, part of this zone of powdered rock is recompacted and partially sintered. It forms a low-permeability shell around the cavity.
504
Chapter 10
In the vaporised, melted and shattered zones, the shock propagates supersonically and its amplitude decreases more or less as the reciprocal of the cube of the distance, that is, very rapidly: at this pressure level, the unloading catches up with the shock and contributes to its attenuation. At 1 m from a 1-kiloton explosion, the pressure is approximately 200 megabar, whereas at 2.5 m it barely exceeds 1.5 megabar. At the end of the hydrodynamic phase, the radius of the cavity is equal to the radius of the zone vaporised and melted by the shock wave. The high-pressure and hightemperature gases that fill this cavity then expand adiabatically, pushing back the surrounding rocks, while the shock propagates far ahead. The cavity continues to expand in this manner until the pressure of the gases is balanced by the reaction of the rocks. This reaction can result either from the lithostatic pressure of the rocks, which depends on the explosion depth, or from the cohesive forces of the medium, or from residual stresses of tectonic origin which are the trace of the forces exerted on the medium during the formation of the massif. Figure 10.5 shows the cavity formation process for the case of a 1-kiloton explosion in the standard material. The cavity stabilises at approximately a hundred milliseconds
Fig. 10.5. Cavity formation (a) before the explosion, (b) a few tens of microseconds after the explosion, (c) a few hundred microseconds after the explosion, (d) a few tens of milliseconds after the explosion.
Bomb Test Sites
505
after the detonation time. After stabilisation, the bottom of the cavity is filled with a meniscus of still-molten lava which traps most of the refractory radioactive isotopes generated by the nuclear reaction. When the equilibrium pressure is reached, certain secondary phenomena can still contribute to increase the final volume of the cavity, and above all to significantly increase the weight of rock initially melted by the shock. This mainly involves the melting to a certain thickness of geological material close to the walls under the effect of the temperature of residual hot gases, and the total or partial melting of blocks of rock which become detached from the fractured walls and fall into the hot lava meniscus when the pressure starts to drop. The collapse of the cavity causes stress modification (known as readjustments) in the medium around the chimney, and can generate secondary fracturing. This fracturing can take the form of either vertical fractures, roughly parallel to the walls of the chimney, or radial fractures caused by the subsidence of the geological layers near the chimney. The collapse of the vault and the readjustments are very well observed by seismographs placed up to a few kilometres from the test. After the nuclear test, a drill-hole is bored to take cores of lava melted by the explosion. Examination of the lava enables the operation of the nuclear device to be analysed. A schematic diagram of the mechanical effects in the rocky massif caused by an 8-kiloton explosion at depth of 600 m is shown in Fig. 10.6. The sampling drill-holes are of smaller diameter than that of the shafts accepting the containers, and are used to recover 5 to 10 cm in diameter. Their starting point on the surface is usually offset from the large-diameter shaft to avoid crossing the rubble zone, which constitutes a very heterogeneous medium that is difficult to drill. The upper part of the drill-hole, made using a "destructive" technique (that is, by crushing the rock under the action of the drilling bit) is vertical. On reaching the proximity of the cavity, the drill-hole is deviated to reach the coring zone, in the immediate neighbourhood of the zero point in the lava meniscus (see Fig. 10.7). During the drilling of these radiochemical sampling holes, special precautions must be taken with respect to the radioactivity encountered in the lower part of the drill-hole. Because of the tightness of the geological layers chosen for the zero point location, the drilling cannot encounter radioactive products before reaching the zone fractured by the explosion. In practice, no radioactivity is detected before the immediate surroundings (a few metres) of the cavity. Once the fractured zone is entered, the water of the drill-hole, carrying the rock debris ground by the bit, is taken up by the fractures in the rock and contributes to the filling of the cavity, which is at lower pressure than the surrounding rock and the drill-hole water. The radioactive materials therefore do not leave the fractured zone. To prevent the risk of a gas leak, the drill-hole head is equipped with a blow-out preventer. This device is similar to that used on oil wells to avoid this type of leak, which in their case constitutes a threat of explosion or fire. In fact the risk of leakage is limited in a radiochemical sampling drill-hole, as the gas formed at the moment of the
506
Chapter 10
Fig. 10.6. Schematic diagram of the mechanical effects in the rocky massif caused by an 8-kiloton explosion at a depth of 600 m. The shaft, shown by a vertical line, is plugged with cement and aggregates before the test.
test tends to migrate towards the top of the chimney, where it is trapped. The slanted trajectory of the lower part of the drill-hole means that in most cases the gas accumulation zone is avoided, although the possibility of a cavity totally empty of water and still containing gas in its lower part cannot be excluded. The blow-out preventer also enables any drilling fluid returns from the cavity to be controlled and the waste to be limited to negligible quantities without radiological consequences. Waste of this type is responsible for the traces of iodine- 131 detected in the Mururoa lagoon in the proximity of a sampling drill-hole by Mr. Cousteau in 1987. Before starting the coting operation, a sealed casing, the coting packing, is installed over the full depth of the drill-hole up to the drilling platform. This casing prevents, apart from any collapse of the walls, the contamination of the rock or the lagoon by the cores while they are being raised to the surface. Once in the free air, the cores
Bomb Test Sites
507
Fig. 10.7. Radiochemical sampling drill zone. The trajectory of the radioactive low sampling drillhole avoids the collapsed zones of the chimney as far as possible, as they are difficult to drill.
containing radiological activity are immediately packed in lead containers. When the sampling and the measurements in the drill-hole have been completed, the hole is plugged with cement. Figure 10.8 shows the difference between compaction caused by spalling and cratering by subsidence. At the Pacific Test Center, the depth of the explosion is chosen so that the vertical extension of the collapse of rocks above the cavity is contained within the volcanic layers. The subsidence observed at the surface is caused by the compaction of superficial rocks under the effect of the fallback of the spall layer. At the Nevada Test Site (United States), in the desert at Lop Nor (China), and at Semipalatinsk (Kazakhstan), the tests are carried out closer to the surface, and the full thickness of the rocks located above a test is affected by the collapse of the cavity roof. At the surface a cone-shaped depression with very steep sides is observed.
508
Chapter 10
Fig. 10.8. Top: Spooling- compaction; Bottom: Spooling- compaction- collapse.
10.2 M A R A L I N G A TEST SITE Between 1955 and 1963 the United Kingdom conducted a programme of nuclear weapons development trials at Maralinga in South Australia (Fig. 10.9). In all, seven major nuclear trials involving atomic explosions, and several hundred smaller scale experiments ("minor trials") which dispersed radioactive materials to the local environment were performed. It is 12 of these "minor trials", code-named the Vixen B trials, performed at the Taranaki site (Fig. 10.10) between 1960 and 1963 which are responsible for the plumes of plutonium contamination which extend from the former test range site and onto the Maralinga Tjarutja lands. In the Vixen B trials about 22 kg of plutonium, as well as enriched uranium and beryllium, were explosively dispersed in a sector extending from the west, through north to the north-east of the Taranaki site (Fig. 10.11).
Bomb Test Sites
509
..d 9
0
9
0
5 E 0
r~ 9
>
. ,..,
0
0 . ,..,
0
0
.=.
. ,..,
<
0 r~ 0
~5 . .~
510
Chapter 10
Fig. 10.10. Map of Maralinga area showing major tracks and Oak Valley. The plutonium contamination close-in at Taranaki is mainly in three forms, viz. as a fine dust, as small sub-millimetre particles, and as surface contamination on larger fragments (Bums et al., 1986). In the trials, the plutonium was dispersed in narrow plumes, the main ones extending to the west, north-west, north and north-east of Taranaki. The most extensive of these is the north-west plume, which can be detected crossing West Street between Fifth and Tenth Avenues (Fig. 10.10). In the central area
Bomb Test Sites
r
511
-6
~ ~.,~
r
, ,,N ;> ~ ,.,~
0
0
0
e~
"0
r13 ~ .,,~
0
E .t= eLO ~ ,.,~
0
.tr=
~ ..,.
M
~,~
0
~ ,.~
[...
0
...,
512
Chapter 10
at Taranaki, the surface soil was "ploughed" in the Operation Brumby clean-up in 1967. Beyond the ploughed area the plutonium contamination tends to be on the surface, and at distances beyond about 5 km from Taranaki only microscopic particles of contamination are found (Johnston et al., 1989). A wide-ranging ground survey of Taranaki area was conducted by the Australian Radiation Laboratory (ARL) over the period 1984 to 1986. Field measurements employing thin sodium-iodide detectors and single-channel analysers were used to provide a qualitative indicator of plutonium (Cooper et al., 1985), and quantitative data were obtained by gamma-ray analyses of soil samples (Cooper et al., 1985; Burns et al., 1988). With the exception of a few measurements to the north-west, these data were confined to the area of the former test range, viz. the Taranaki site and sampling points along the major north-south and east-west tracks on the range. Following concern by people living and working at the Aboriginal outstation at Oak Valley, soil sampling was performed by ARL in 1987 along the Oak Valley Road and along Western Avenue (Fig. 10.10) as well as at the Oak Valley settlement. Highresolution gamma-ray spectrometry was used in the laboratory to determine levels of plutonium in these soil samples. These showed the presence of trace amounts of plutonium at many of the sampling points in the general area to the north-west of Taranaki (Williams and Bums, 1987). However, the contamination was found to be particulate in character with a surface density of particles too low for soil sampling to be a suitable technique. A technique more suited to this particular situation is that of in-situ gamma-ray spectrometry using a portable germanium detector. This is the method which has been employed in the study by Johnston et al. (1989), conducted during 1987 and 1988, to detect plutonium contamination at levels well below 1 kBq m -2. The plutonium contamination at Maralinga contains a mixture of isotopes. While the predominant isotope is 239pu, there are significant quantities of 24~ and 241pu. 241pu decays to 241Am with a half-life of 14.4 years, and consequently most of the original 241pu has by now decayed to 241Am. 239pu and 24~ are not readily detected by gamma-ray spectrometry because of the low abundance of their gamma-ray emissions, whereas 24~Amhas an abundant gamma-ray (36%) of energy 59.5 keV. This serves as a good indicator of plutonium concentration once the relevant ratios of the plutonium isotopes to americium have been established. For much of the contamination dispersed in the 12 Vixen B trials at Taranaki, the ratio of total plutonium activity to americium activity was of the order of ten (Johnston et al., 1988). The activity ratios 239pu/241Am and 24~ in the individual plumes emanating from Taranaki have been measured (Burns et al., 1989) and the appropriate values are used in determining plutonium levels in the north and north-west plumes. In-situ gamma-ray spectrometry used by Johnston et al. (1989) was capable of detecting surface deposits of americium at surface densities as low as 0.05 kBq rn-2 (corresponding to 0.4 kBq -2 of plutonium in the north-west plume). The results indicate that the plumes extend well beyond the limits of the Maralinga range and traces of the north-west plume persist to beyond 80 km from Taranaki (see Fig. 10.11).
Bomb Test Sites
513
For practical considerations, one can use a conservative value of 7.4 kBq m -2 of plutonium proposed by the U.S. Environmental Protection Agency in 1985 as a recommended "screening level", below which land can be considered suitable for unrestricted use (US EPA, 1985). This corresponds to a concentration of approximately 0.5 Bq g-~ for the soils in the region, where the observed penetration of plutonium into the soil is about 1 cm, or to a surface density of approximately 1 kBq m -2 of 24~Amin the north-west plume for which the measured ratio of plutonium to americium is seven (Bums et al., 1989). This level is exceeded outside of the Maralinga Range immediately adjacent to the western boundary and extending for the first few kilometres in a north-westerly direction. At greater distance, the surface density falls off and the plume is observed to veer more to a westerly direction, consistent with meteorological data from the time of the particular test, and passes south of the Oak Valley area. Similarly, the north and north-east plumes are still detectable at the boundaries of the Range at Twenty-fifth Avenue and East Street respectively although the concentrations are somewhat lower. Cooper et al. (1994) have reported re-suspension studies on soils contaminated with plutonium during nuclear weapons tests by use of a mechanical dust-raising apparatus. Airborne dust was analysed in terms of mass and 241Am activities for particle sizes less than 7 ~tm. The AMAD was determined as 4.8-6 ~tm for re-suspended soil. Also, surface soil was characterised in the laboratory by means of sieving and microparticle classification, yielding mass and 24~Am activity distribution with respect to size. Data indicate the granularity of plutonium contamination at both major and minor trial sites. Depth profile analyses for undisturbed areas demonstrate that most (74% on average) of the americium and plutonium activity is found in the top 10 mm of soil. Plutonium and americium activities were found to be enhanced in the inhalable fraction over their values in the total soil, and the enhancement factors were similar in re-suspended dust and surface soil. Observed enhancement factors ranged from 3.7 to 32.5.
10.3 T E S T I N G SITE IN THE M A R S H A L L ISLANDS Bikini and Enewetak Atolls were used as sites for tests related to nuclear weapons by the USA between 1946 and 1958 (see map in Fig. 10.12 for the location of Marshall Islands and vicinity). A few of the nuclear weapon tests in the Pacific Ocean were conducted by the USA outside the Marshall Islands, near Johnston Atoll and Christmas Island (the latter Kiribati, formerly the Gilbert Islands); however, these tests were limited to high altitude explosions. Bikini Atoll was the site of 24 of the 66 tests conducted under water, at ground level and above ground in the Marshall Islands (see Fig. 10.13). The yields of the tests at Bikini Atoll amounted to about 72% of the total yield of 1.1 x 105 kilotonnes (kt) of TNT equivalent for both test sites (Simon and Graham, 1995).
514
o
. "~
,~_r
~ !
~
q
%,,
ti
eqbt9
" ~ . ' ~ " . ~ .,_.~".
/
.. !
ooo~ II 9
\
\ \
i
i
I
-
r~'~ ~
~
~'->-_
<
J/lfl
~
<
< ~
~ <
.~
~<
I z
lkC
~ - ~ \
~ -
9
~
--
t'r'') \ D ~
_
co
"~
Ir.#~
.
9
9
Z < ..1
< I
\
< < .
_o, r..p~ {.-
0 Z
z .2 Oz
r.i.1
<
r.c)
o z r~,_,
Chapter 10
.,..~ .,..~
c~
Bomb Test Sites
100,000
515
t i
,i
![
,
,
I
|
,
'
!
I
i
I
10,000
-
...................................
I
I
i ......................................................................................................................
I
1,000
I
. , ...............................................................................
[.Z
ua~ "0'
I
"6 . . . .
-10,000
I
(=1 MT)
>Z
100,000
o o
:
Z
I
1,000
i
O~ 100 ....................................i.........................................................................................................................9............z...................
X
- 100
Z < D 0
10
........................................................................................................................................................................................................................... 10
1
]
1946
|
1947 1948
I
I
1949
J
I
1950 1951
I
1952
I
1953
I
1954
J
1955
I
1956
1
1957
l
1
1 9 5 8 1959
DATE OF T E S T
F i g . 10.13. D a t e s o f 24 a t o m i c w e a p o n s tests at B i k i n i A t o l l a n d e x p l o s i v e y i e l d s (after S i m o n a n d G r a h a m , 1995).
The names of different islands (motu) on Bikini and Enewetak Atoll are shown in Table 10.6. Testing at Bikini Atoll started with "Operation Crossroads" in 1946. This experiment staged by the US Navy, which included the so-called "Able" and "Baker" shots, involved 242 ships, 156 aircraft and more than 42,000 military and civilian personnel, and used more than 5000 experimental animals. From July 1946, Bikini Atoll remained inactive as a test site and tests were conducted in Enewetak Atoll in 1948, 1951 and 1952. Then, in February 1954, Bikini Atoll was reactivated as a test site with the "Castle" series of tests. They continued in 1956 with the "Redwing" series and were terminated in 1958 with the "Hardtack I" series. The tests of highest yield were those in the "Castle" series, which included the "Bravo" shot, a thermonuclear device of 15 megatonnes (Mt) equivalent yield of TNT. Table 10.7 presents data for the trials at Bikini Atoll (Simon, Robinson, 1997; Bikini Atoll Rehabilitation Committee, 1983, 1984; Schell et al., 1980; USDOE, 1994; Carter and Moghissi, 1977). Figure 10.14 shows approximately where in Bikini Atoll the nuclear devices were detonated (Bikini Atoll Rehabilitation Committee, 1983, 1984). Prior to the Able test in 1946, the first nuclear test in Bikini Atoll, the 167 Bikinians then living on Bikini Island were evacuated to Rongerik Atoll, about 200 km to the east,
Chapter 10
516 Table 10.6 Islands (motu) of Bikini and Enewetak Atolls
Bikini
Enewetak
Marshallese Name
U.S. Code Name
Japanese Name
Nam Iroij Odrik Lomilik Aomen Bikini Bokantauk Lomelen Enealo Rojkere Eonjebi Eneu Aerokojlol Bikdrin Lele Eneman Enidrik Lukoj Jelete Adrikan Oroken Bokaetoktok Borkdrlul Bravo Crater Bikini Lagoon Bikini Ocean Boro Reef Bikini Reef Tewa Carter Zuni Crater
Charlie Dog Easy Fox George How Item Jig King Love Mike Nan Peter Roger Sugar Tare Uncle Victor William Yoke Zebra Alpha Bravo
Yurochi Yorikku Romurikku Aomeon Bikini Bokonfaaku Yomyaran Eniairo Rochikarai Ionchebi Enyu Airukiraru Bigiren Reere Eniman Enirik Rukoji Chieerete Arrikan Ourukaen Bokoaetokutoka Bokororyuru
Bokoluo Bokombako Kirunu Louj Bocinwotme N.E. of Bocinwotme
Alice Belle Clara Daisy Edna Flora
Namu
Ruchi Cochiti
seemingly to reside there until an unspecified future date when the testing would be completed. ( K n o w l e d g e at that time about the long-term consequences of radioactive fallout and the transfer of radionuclides through the food chain was limited.) The Bikinians r e m a i n e d on Rongerik Atoll for a period of two years. In 1948, they were m o v e d briefly to Kwajalein Atoll and later in the same year to Kili, a small (0.8 k_m2)
B o m b Test Site s
517
Table 10.7 Nuclear weapon tests conducted at Bikini Atoll Test series
Shot name
Date
Type
Yield (kt TNT equivalent)
Map reference (see Fig. 10.14)
Crossroads Crossroads Castle Castle Castle
Able Baker Bravo Romeo Koon
30 June 1946 24 July 1946 28 February 1954 26 March 1954 6 April 1954
Air drop Underwater Surface Barge Surface
23 23 15000 11000 110
A A B B C
Castle Castle Redwing
Union Yankee Cherokee
25 April 1954 4 May 1954 20 May 1956
Barge Barge Air drop
6900 13500 3800
D D E
Redwing Redwing Redwing Redwing
Zuni Flathead Dakota Navajo
27 May 1956
Surface
3500
C
11 June 1956 25 June 1956 10 July 1956
Barge Barge Barge
365 1100 4500
F F D
Redwing Hardtack I Hardtack I Hardtack I Hardtack I Hardtack I Hardtack I Hardtack I Hardtack I Hardtack I Hardtack I
Tewa Fir Nutmeg Sycamore Maple Aspen Redwood Hickory Cedar Poplar Juniper
20 July 1956 11 May 1958 21 May 1958 31 May 1958 10 June 1958 11 June 1958 27 June 1958 29 June 1958 2 July 1958 12 July 1958 12 July 1958
Barge Barge Barge Barge Barge Barge Barge Barge Barge Barge Air burst
5000 1360 25.1 92 213 319 412 14 220 9300 65
G B H B I B I H B J H
/ / /
Nam Odrik Bwikor ~ Lomilik / , ~ ~ \ Aomen ~ i 89 j B G r~ i ~ ~ . . ~ Nauticalmiles(NM)
//
Bikini Bokantauk ~j)13oKoclrolul . A --~,1 ~ Lomelan "~o_Bokaetoktok Sunken k~ Enealo Oroken "q'l_,_ ~hins ~ Rojkere Adrikan~-'~ C H . . . . . . . ]~'_.f~Bokonjebl Jelete " ~ ~ f - _ ~ e r o r , uj~o~ Jl) Eneu Lukoj I ~ ~ Bikdrin "" Enidrik Lele Eneman //' A Test site 'A' (,(
-
. . . . .
D
E
_
r -----
~
,--1
"~]
Fig. 10.14. Bikini Atoll: (A-J) locations of nuclear weapon test detonations (1 NM = 1.85 km).
518
Chapter 10
isolated island. Kili Island is fertile, with rich soil, but is less than half the size of Bikini Island. It has no lagoon, no protective reef and no fishing grounds. The small beach is frequently subject to high waves. The Bikinians saw the move to Kili as a temporary relocation and were reluctant to change from being fishermen to being farmers. Nuclear weapon testing in the Marshall Islands was terminated in July 1958. On 31 October 1958, the USSR, the United Kingdom and the USA suspended atmospheric nuclear weapon testing under an international moratorium. The Treaty Banning Nuclear Weapon Tests in the Atmosphere, in Outer Space and Under Water was signed in Moscow on 5 August 1963. In August 1968, following a number of radiological surveys (Robinson et al., 1977) that had been carried out since 1958 to assess the impact of the USA's programme of nuclear weapon testing, president Lyndon Johnson publicly announced that Bikini Atoll was safe for habitation and approved the resettlement of the B ikiniain people on the atoll. From February to October 1969, the atoll was cleared of debris. Fruit trees, including coconut, breadfruit, poandanus, papaya and banana, were replanted. A further radiological survey of Bikini Atoll was carried out in 1970. Initially, in 1970, three Bikinian families and about 50 Marshallese workers returned to the atoll. Eventually, 139 Bikinians would resettle there. However, the Bikinian people remained unconvinced of the safety of the atoll, and in 1975 they initiated a lawsuit against the government of the USA to terminate the resettlement effort until a satisfactory and comprehensive radiological survey had been carried out. In 1975, a further radiological assessment of Bikini Atoll was conducted (Robinson et al., 1977). However, at that time the trees planted in 1969 had not yet grown to maturity and few samples were available for reliable estimates to be made of radionuclide concentrations in food crops. In 1976, an external radiation survey programme for five northern atolls, which included some measurements at Bikini, was conducted. A continuing sampling and analytical programme began at Bikini Atoll in 1978 to gather additional data as a basis for more precise radiation dose estimates for the residents of Bikini and Eneu Islands. Radioanthropometry (whole body radiation measurements) for the purpose of estimating the intake of radioactive materials by Bikinian residents began in April 1977. In 1978, it was determined that for the inhabitants of Bikini Atoll a tenfold increase in the body content of the radionuclide 137Cshad occurred (Miltenberger et al., 1980). This increase was the result of a combination of the coconut trees starting to bear fruit and a drought that led to increased consumption of coconut fluid. Apart from assessments of the long-term impacts on the Bikinians, studies have been conducted on service personnel and Japanese fishermen exposed, in particular, as a consequence of the Castle Bravo test (Klenm et al., 1986; Kumatori et al., 1980; Eisenbud, 1987; Sharp and Chapman, 1957). In August and September 1978, in response to the high uptake of caesium in the population--then composed of the 139 Bikinians who had returned to Bikini Atoll-officials of the Trust Territory decided to relocate the B ikinians again from their atoll, back to Kili Island and to Ejit Island at Majuro Atoll.
Bomb Test Sites
519
At the time of the second relocation, a new radiological survey in 11 northern atolls of the Marshall Islands, sponsored by the USA (Department of Energy), was started. The survey used detectors mounted in helicopters which were flown in parallel flight lines in order to plot external gamma dose rate contours (Tipton and Meibaum, 1981). Also, samples of vegetation, marine foods, animals and soil were collected and analysed (Robinson et al., 198 la, 198 lb). Revised radiation dose evaluations were published in 1980 and 1982 which indicated that, should the Bikinians decide to resettle their island, the terrestrial food chain would be the most significant exposure pathway. This dose assessment was most recently updated in 1995 on the basis of a continued measurement programme at the atoll (Robinson et al., 1997; Kehl et al., 1995). Additional information on radiological surveys is reported in Robinson et al. (1997). The Marshall Islands Dose Assessment and Radioecology Project has been in existence at Lawrence Livermore National Laboratory (LLNL) since 1973. It was a program of the Health and Ecological Assessment Division (HEA), in the Environmental Programs Directorate at LLNL. The primary purpose of this program was to assess the radiological conditions in the Marshall Islands. The radiological dose via all exposure pathways is estimated for various living pattems at the atolls. LLNL project was also studying remedial measures for reducing 137Csuptake in vegetation, as part of the resettlement options at Bikini Atoll. Table 10.8 shows the total number of samples that have been collected during the twenty-two year history of the program. The samples include soil, edible food crops, other vegetation, fish, invertebrates, water and animals. The samples are prepared for gamma spectroscopy and/or wet chemistry and analysed. Databases were designed for all of the information associated with the samples. The Data Management Group (DMG) has the responsibility of managing this information (Stoker and Conrado, 1995). Their program includes relational database design, programming and maintenance; sample and information management; sample tracking; quality control; and data entry, evaluation and reduction. The usefulness of scientific databases involves careful planning in order to fulfil the requirements of any large research program. Compilation of scientific results requires consolidation of information from several databases, and incorporation of new information as it is generated. In the period from 1979 to 1989, approximately 25,000 Post Northern Marshall Islands Radiological Survey (PNMIRS) samples were collected, and over 71,400 radiochemical and gamma spectroscopy analyses were performed to establish the concentration of 9~ 137Cs,~41Am,and plutonium isotopes in soil, vegetation, fish and animals in the Northern Marshall Islands. While the Low Level Gamma Counting Facility in the Health and Ecological Assessment (HEA) Division of Lawrence Livermore National Laboratory accounted for over 80% of all gamma spectroscopy analyses, approximately 4889 radiochemical and 5437 gamma spectroscopy analyses were performed on 4784 samples of soil, vegetation, terrestrial animal, and marine organisms by outside laboratories. Four laboratories were used by Lawrence Livermore National Laboratory (LLNL) to perform the radiochemical analyses: Thermo Analytical Norcal, Richmond, California (TMA); Nuclear Energy Services, North
Chapter 10
520
Table 10.8 Total number of samples collected in the Marshall Islands from 1973 to 1994 (after Stoker and Conrado, 1995) Soil and Vegetation Samples Year taken
Bikini Atoll
Enewetak Atoll
Rongelap Atoll
Utirik Atoll
Other northern Total Marshall Atolls
1973
0
4474
0
0
0
1974
0
0
0
0
0
4474 0
1975
941
0
0
0
0
941
1976
0
991
0
0
0
991
1977
998
728
0
0
0
1726
1978
1556
124
728
463
2807
1979
1084
64
0
0
0
1980
823
75
0
0
0
898
1981
288
53
0
0
121
462
1982
314
246
0
0
0
560
1983
1008
180
0
0
166
1354
1984
489
398
0
0
0
887
1985
3136
138
31
0
0
3305
1986
3015
121
811
0
0
3947
1987
3270
598
45
0
24
3937
1988
3201
498
201
0
0
3900 4255
5678 a 1148
1989
1838
1102
1315
0
0
1990
2629
576
524
0
137
3866
1991
3527
556
635
0
0
4718
1992
2859
365
819
0
0
4043
1993
2449
1498
832
1230
0
6009
1994
3810
966
39
522
0
5337
91
9
0
0
91
37326
13751
5980
Others b Total
2215
3255
62527
Carolina State University (NCSU); Laboratory of Radiation Ecology, University of Washington (LRE); and Health and Ecological Assessment (HEA) division LLNL, Livermore, California. Additionally, LRE and NCSU were used to perform gamma spectroscopy analyses. The analytical precision and accuracy were monitored by including blind duplicates and natural matrix standards in each group of samples analysed. On the basis of reported analytical values for duplicates and standards, 88% of the gamma and 87% of the radiochemical analyses in this survey were accepted. By laboratory, 93% of the radiochemical analyses by TMA; 88% of the gamma-ray spectrometry and 100% of the radiochemistry analyses by LRE; and 90% of the radiochemistry analyses performed by HEA's radiochemistry department were accepted (Kehl et al., 1995).
Bomb Test Sites
521
The Congress of the USA created a "Resettlement Trust Fund for the People of Bikini Atoll" for the purpose of improving living conditions on Kili. It also set up the "Bikini Atoll Rehabilitation Committee" to study and report on the feasibility and cost of rehabilitating the atoll. In 1984, this Committee issued its first report, stating that Bikini could be resettled provided that no locally grown foodstuffs or ground water would be consumed. The Committee also considered other courses of action, including the removal of topsoil from the islands. In January 1986, a Compact of Free Association between the governments of the USA and the Marshall Islands was signed into law. This provided for the payment of compensation to the people of Bikini, Rongelap, Enewetak and Utirik Atolls. An additional trust fund was established for the cleanup and resettlement of Bikini Atoll. A separate radiological assessment--the Republic of the Marshall Islands Nationwide Radiological Study (NWRS)mwas commissioned by the Government of the Republic of the Marshall Islands. By this means, Bikini Atoll, as well as all other atolls in the Republic, was to be monitored for radioactive residues. Oversight was provided by a Scientific Advisory Panel of well known and respected scientists (McEwan et al., 1994). Laboratory quality control programmes were implemented to ensure that the NWRS surveys could provide accurate measurements. In general, the study confirmed the findings of earlier measurement programmes. The findings of the NWRS were published and a report on Bikini Atoll was released in February 1993 (Simon and Graham, 1997). The atomic weapons tests created many different radioactive elements, however, most of these existed for only a very short time. Some radioactive elements lasted only for a few seconds or minutes to a few days. The residual radioactivity that exists today is made up of radioactive elements with half-lives of more than a few years. The predominant part of the residual radioactivity still in existence today is 137Cs. Most radiation exposure or radiation dose comes from ~37Cs. The ~37Cs specific activity as a function of soil depth in 1987 is shown in Fig. 10.15 (after MARC 1984). In August 1995, six months after the NWRS issued its report on Bikini Atoll, the Nitejela of the Marshall Islands considered the NWRS findings but did not accept them. During this period several studies have been performed in this area. Let us only mention the report by Simon et al. (1995) in which plutonium-contaminated soil from the Republic of the Marshall Islands has been studied to determine the spatial and volume characteristics of contamination on two scales: (1) in macroscopic masses, i.e., gram sized samples, and (2) in microscopic masses, i.e., tens of ~tg to 1 mg. Three measures of volumetric homogeneity calculated from alpha track measurements on a plastic track detector (CR-39) were performed to quantitatively assess microspatial or microvolumetric variations. Data was reported for four different samples obtained from locations 40 to 90 m apart on Rongelap Island. The samples were of near equal concentration as determined by macrovolume measurements: about 122 Bq/kg 239'24~ and 73 Bq/kg 24~Am. The concentration of transuranic radioactivity (239'24~ plus 241Am) in the four samples generally increased with decreasing particle size in macro-size samples. The
Chapter I0
522
100
!
1 .....
I
I
I
80 50
A=80.5e -~176 ROOTING ZONE=28.6
BIKINI: 30
pCi/gm
(0-40 cm)
20
c~ [--,
10
8
> b-,
5
< C,.)
3
2~
2
ENEU: t-,--,
A=5.53e -0.0s24z R O O T I N G ZONE=2.31 (0-40 cm)
1
pCi/gm
0.8
!
~)
0.5
0.3
~I 0.1
0
I
I
I
I
I
10
20
30
40
50
60
DEPTH / cm Fig. 10.15. '37Csspecific activity as a function of soil depth in the year 1987 (after BARC 1984). variation of concentration among the four samples decreased with decreasing particle size indicating greater similarity in the size fraction <40 ~tm. Somewhat different results were found in studies from the British Atomic Weapons Test Site at Maralinga, South Australia, where specific activities were noted to be greater in the soil size fractions >90 ~tm (Ellis and Wall, 1982). Presumably there are numerous factors that might influence the relationship of plutonium activity with soil particle size including the nature of the contaminating event, the degree of weathering since the contamination event, the chemical nature of the soil, and the particle size distribution of the soil. The Republic of the Marshall Islands was accepted as the 122nd Member State of the IAEA on 26 January 1994. The Marshall Islands Government subsequently requested the IAEA to conduct an independent international review of the radiological conditions at Bikini Atoll, and to consider and recommend strategies for the resettlement of the atoll. The IAEA responded to this request by convening an Advisory Group, which met in Vienna on 11-15 December 1995 (IAEA-RAR Series, 1998).
Bomb Test Sites
523
On the basis of the amount and quality of the scientific information on the residual radionuclides from nuclear weapon testing at Bikini Atoll submitted for review, it is concluded that: 1. No further independent corroboration of the measurements and assessments of the radiological conditions at Bikini Atoll is necessary. This conclusion was based on: the excellent quality control of those measurements and assessments; the regular participation in intercomparison programmes by the various scientific groups that carried out those measurements and assessments; and the good agreement among the data submitted. Nevertheless, it is acknowledged that the Bikinian people have concerns about the actual radiological conditions in their homeland, and it is therefore considered that: 2. The Bikinians might be reassured about the actual radiological conditions at Bikini Atoll by a limited programme of monitoring of radiation levels, which should involve some participation by members of the community. In view of the information submitted and under the assumption that the B ikinian community decides to resettle Bikini Island (the main island of residence at Bikini Atoll) it is concluded that: 3. Permanent resettlement of Bikini Island under the present radiological conditions without remedial measures is not recommended in view of the radiation doses that could potentially be received by inhabitants with a diet of entirely locally produced foodstuffs. This conclusion was reached on the basis that a diet made up entirely of locally produced food which would contain some amount of residual radionuclides could lead the hypothetical resettling population to be exposed to radiation from residual radionuclides in the island, mainly from 137Cs, resulting in annual effective dose levels of about 15 mSv (if the dose due to natural background radiation were added, this would result in an annual effective dose of about 17.4 mSv). This level was judged to require intervention of some kind for radiation protection purposes. However, it is considered that: 4. In practice, doses caused by a diet of locally derived foodstuffs are unlikely to be actually incurred under the current conditions, as the present Marshallese diet contains--and would in the near future presumably continue to contain--a substantial proportion of imported food which is assumed to be free of residual radionuclides. Nevertheless, the hypothesis of a diet of solely locally produced food was adopted in the assessment for reasons of conservatism and simplicity, and also because the present level of imports of foodstuffs could decrease in the future. A number of straightforward environmental remediation strategies at Bikini Island have been considered, which, if properly implemented, would achieve very satisfactory results from the point of view of radiation protection. It is therefore concluded that: 5. Provided that certain remedial measures are taken, Bikini Island could be permanently reinhabited. Several possible remediation strategies were considered with the result that the following were selected as a basis for further assessment:
524
Chapter 10
9 the periodic application of potassium based fertiliser to all areas of Bikini Island where edible crops may be grown, supported by the removal of soil from around and beneath the dwelling areas and its replacement by crushed coral (known as the potassium fertiliser remediation strategy); 9 the complete removal of the topsoil from Bikini Island (called the soil scraping remediation strategy). 6. While no definite recommendations are given on which strategy to follow, it is considered that the strategy using potassium fertiliser is the preferred approach. In this connection, it was noted that the soils of Bikini Atoll are extremely deficient in potassium and extensive field trials have demonstrated that the application of potassium rapidly reduces the concentration of '37Cs in food crops since potassium is taken up by the plants in preference to caesium. The reduction of 137Csin the food crops is sustained for about four to five years, after which the values slowly begin to increase again. However, repeated application of fertiliser forms an effective strategy in reducing the estimated doses to the potential inhabitants of Bikini Island. Furthermore, the supporting strategy of removing soil from dwelling areas would eliminate most of the external and internal exposures from direct soil ingestion or inhalation. 7. The results expected from the potassium fertiliser remediation strategy are consistent with international guidance on interventions to avoid dose in chronic exposure situations and, therefore, this strategy would provide a radiologically safe environment permitting early resettlement. Depending on the assumptions made concerning diet, the annual calculated mean effective dose would be reduced as follows: from about 15 mSv (if the dose due to natural background radiation were added, this would result in an annual effective dose of about 17.4 mSv), for a high calorie diet of totally local foodstuffs, to about 1.2 mSv (if the dose due to natural background radiation were added, this would result in an annual effective dose of about 3.6 mSv); and from about 4 mSv (if the dose due to natural background radiation were added, this would result in an annual effective dose of about 6.4 mSv), for a high calorie diet of both local and imported foodstuffs, to about 0.4 mSv (if the dose due to natural background radiation were added, this would result in an annual effective dose of about 2.8 mSv). Even for the more conservative assumption of a high calorie diet of totally locally produced foodstuffs, the resulting doses will be below acceptable genetic action levels for intervention. The doses will be somewhat higher than those due to natural background radiation that were incurred by the inhabitants of Bikini Island before the evacuation and prior to when the nuclear weapon tests took place, and also somewhat higher than global average natural background doses, but lower than typical elevated levels of natural background doses around the world. 8. The alternative strategy, i.e. the soil scraping remediation strategy--stated to be the alternative preferred by the Bikinians--would be very effective in avoiding doses caused by the residual radionuclides, but it could entail serious adverse environ, mental and social consequences.
Bomb Test Sites
525
The consequences may be serious because the fertile topsoil supports the tree crops, which are the major local food resource. The replacement of the soil with topsoil from elsewhere would be an enormous undertaking which is likely to be prohibitively expensive. The content of natural radionuclides in any continental soil used as replacement soil would most probably exceed that of the present soil. 9. No remedial actions should be proposed at this stage for the islands of Bikini Atoll other than Bikini Island. The other islands have historically been non-residential and used only for occasional visits and for fishing. On the assumption that the proposed remediation strategy is undertaken, it is further recommended that: 10. Regular measurements of activity in local foodstuffs should be made to assess the effectiveness of the measures taken. A simple, local whole body monitor and training in its use should be provided as a further means of enabling potential inhabitants to satisfy themselves that there is no significant uptake of caesium into their bodies. (After IAEA-RAR series, 1998).
10.4 S E M I P A L A T I N S K N U C L E A R TEST SITE The test site, shaped like an irregular polygon and familiarly called the polygon, is a 19,000 km 2 zone in the northeast of the newly independent Republic of Kazakhstan, 800 km north of the Kazakh capital Alma-Ata. The zone lies southwest of the Irtysh River which flows into Kazakhstan from China and which for a short stretch, where it veers sharply northwards on its way to join the Ob River in Siberia, forms part of the polygon boundary (see Fig. 10.16). The USSR conducted 465 nuclear tests at three locations called "technical areas" within the polygon over a period of 40 years (1949-1989) for military and peaceful purposes. The earliest tests were above ground (atmospheric and surface) and were carried out in the northern technical area, otherwise called Ground Zero. There were 118 of these explosions in 1949-1962. Of particular concern were the 30 explosions carried out on the surface, especially five which were unsuccessful and resulted in the dispersion rather than the fissioning of the plutonium in the devices. The other 346 test explosions were underground, in the widely separate technical areas in the south (223 between 1961-1989) and east (123 from 1968 to 1989). 13 of these resulted in release of radioactive gases to the atmosphere. In addition, an explosion designed to build a dam across the small Tchagan River, close to the eastern technical area, was miscalculated and resulted in a lake, about 0.5 km in diameter and about 100 m deep with above-ground cliffs up to 100 m high, called Lake Balapan. The only habitations within the polygon during the 40 years were the custom built town of Kurchatov (code named Semipalatinsk-21) north of Ground Zero, dedicated to servicing the test site, and the small settlement of Akzhar on its northern edge. Neither was radiologically affected by the tests. Recently, semi-nomadic farmers and herders have formed small scattered "settlements" near two test areas, notably a farm about 12 km south of Ground Zero and another some 10 km east of Lake Balapan. There are
526
Chapter 10
Fig. 10.16. Semipalatinsk nuclear test site "Polygon", area = 18,000 km2. some 15 and about 100 people using the two areas respectively, but as yet not throughout the year. The principal settlements of concern to the Government are outside but close to the test site boundary, along its southern and eastern borders. They lie in the path of the radioactive plumes caused by the above-ground explosions.
Bomb Test Sites
527
Russian records show that the plumes travelled south from the northern technical areas, beyond the southern border, then veered sharply east and again sharply north before dispersing beyond the Irtysh. Recently, many countries of the former Soviet Union have expressed to the IAEA their deep concern about the radiological situation in areas that were once used for nuclear-related activities such as uranium mining/processing and weapons development, manufacture and testing. The independent Republic of Kazakhstan petitioned the IAEA for assistance in re-evaluating the radioactivity contamination in and around the former nuclear weapons testing site at Semipalatinsk in order to assess the health risks for the population. The IAEA established a special project through its Department of Nuclear Energy and Safety (Rosen, 1993). With the assistance of the Department of Technical Cooperation, two expert missions were sent to Semipalatinsk to provide a preliminary assessment of the environmental contamination in that area. The second mission consisted of experts from France, Kazakhstan, the Russian Federation, the United Kingdom, the United States of America and the IAEA (IAEA-1995). Their mission was to investigate the current radiological situation within and in the vicinity of the former USSR nuclear test site Semipalatinsk. Both missions were part of a project, launched at the request of the Government of Kazakhstan, to make a preliminary assessment of the radiological situation in terms of its potential effects on the health of people living in the area today, rather than to reconstruct doses and health effects of the nuclear experiments to people in the past. It is estimated that 30,000 to 40,000 people now live along the plume path, though only some of the settlements were visited to assess the current hazards from living in the area. The assessments were made by measuring external dose rates, making measurements of radionuclide concentrations in many materials and assessing the usage that people make of the environment. From data of this type the doses to people can be assessed and the corresponding risk estimated. The precise objectives were to corroborate the levels of environmental contamination arrived at by recent Russian and Kazakh studies and to make a preliminary radiological assessment of the situation in the settlements and in the area. Sampling protocols which had been outlined in detail before the start of the second mission along IAEA recommended guidelines (IAEA, 1989) were followed carefully with only minor deviations. Fresh milk was collected from individual farms in new fluoro-plastic 1-1itre bottles deep-frozen in liquid nitrogen just before air shipment in an insulated box. Grass-vegetation samples, typically about 300 g, were cut from 1 m 2 pastoral areas with a light-weight grass trimmer, flesh weight was determined on site using an electronic hanging balance, and the specimens were air-dried. Soil cores of 5 cm diameter and about 20 cm depth were removed from selected sites with a steel pipe coring tool. In most cases, the intact soil cores were carefully extruded, wrapped in aluminium foil and packaged tightly into plastic shipping tubes in the field. All of the milk, grass/vegetation and soil core samples were transported in secured containers, shipped by air freight to Vienna, and retrieved by the Agency's Laboratories. Reasonable precautions were taken to preserve sample integrity throughout the
528
Chapter 10
mission. Several duplicate samples were placed in the custody of the Kazakhstan team members. The corroboration of environmental contamination levels obtained by independent equipment and measurements of the team was generally good, the best being with recent Russian and Kazakh data using gamma dose rate measurements. Acceptable corroboration was observed for gamma-emitting radionuclides in food and environmental samples. The preliminary results on plutonium levels in soil samples from contaminated sites in the polygon showed values comparable with the data reported by Russian scientists (see also Stegnar and Wrixon, 1998). In addition soil samples were collected to estimate the inventory and to determine the depth distribution of 137Cs, 152Eu, 155Eu, 6~ 24)Am and 154Eu. Soil samples were collected at locations where gamma-ray exposure data indicated reasonable local uniformity. Almost all soil samples were collected in flat, undisturbed areas used for grazing cattle, sheep and/or horses, and thus had short-cropped vegetation. Samples were collected using 8.9 cm diameter soil cutters. A 5 cm deep cut was removed, followed by a 10 cm corer inserted into the same hole to obtain a 5-10 cm cut, and finally, a 15 cm corer was used to obtain a 10-15 cm cut. In some instances, a core down to 30 cm was obtained using an auger. This sampling procedure is described in the EML Procedures Manual (Chieco et al., 1990). Due to time, weight and other logistical considerations, all sites were sampled using 3 cores. The samples were collected at approximately equidistant locations and 3 m from the gamma spectrometer. The surface area collected using this technique (186 cm 2) does not represent the site as precisely as the normal 10-core sample as per ASTM procedures (1983). However, experience in soil analyses indicates that the total error in the sampling, preparation and the gamma analysis will be about 15% for the 3-core samples as opposed to an estimated 8% error when using the 10-core method. The respective cuts of the soil from the three cores were composited, broken up by hand, and homogenised as well as possible. The sample was then spread out on a plastic tarp and quartered, with stones and vegetation evenly distributed. Two of the quarters were kept, resulting in an approximate split of the sample so as to reduce the sample size. In the laboratory, the soil samples were air dried for 3-10 days in plastic trays. The samples were not sieved but large stones were removed before the samples were sealed in 90 ml aluminium cans. They were then allowed to stand for several weeks so that the radon progeny could build into equilibrium. An HPGe spectrometer system comprising a reversed bias 35% efficiency (relative to a 7.62x7.62 cm NaI crystal at 1332 keV) was used to analyse the samples. The energy region examined was 20-3000 keV, and counting times ranged from one to several days depending on the activity of the sample and desired accuracy of the results (Shebell and Hutter, 1995). The major sites selected for field work by the team were the settlement around the polygon of Kainar (population of about 10,000) in the south, Sharzhal (2000) and Karaul (5000) in the east, and Dolon (2000) just north of the Irtysh. Akzhar, within the polygon just south of the river, was used as a reference site. Inside the polygon the efforts were concentrated in the Lake Balapan area including the semi-permanent farm
529
Bomb Test Sites
Table 10.9 Contamination levels assumed in dose assessment (IAEA Yearbook, 1995) Nuclide
Soil activity concentration (Bq/kg) Settlement
Lake Balapan
Ground Zero
9~ 137Cs 238pu 239pu
40 40
25 000 35 000
20 000 30 000
0.5
6 000
5 000
1.0
14 000
10 000
240pu
1.0
14 000
10 000
241Am
0.2
2 300
2 000
around Ground Zero, and a selection of sampling sites along the plume paths of atmospheric and above-ground explosions. The operations carried out in the field included: gamma dose rate measurements; in situ gamma spectrometry; and the collection of samples of grass, meat, milk, offal, vegetables and soil, as well as biological indicators such as animal bones, mushrooms and moss. The levels of contamination in the soil at the locations specified are shown in Table 10.9. The contamination by 9~ in milk, drinking water and the lake water was also measured, together with results for 137Cs in meat. The external gamma dose rates in settlements and in the polygon, excluding the Lake Balapan and Ground Zero areas, were around 0.1 ~tGy per hour, against rates of up to 40 ~tGy per hour around Lake Balapan and Ground Zero. The dose assessment included consideration of all relevant pathways, of which the most important were external gamma exposure from material on or in the ground, inhalation of material resuspended from the ground and consumption of contaminated foods. These pathways were taken into account in assessing the doses to people in the identified settlements and the other areas. For the assessment of current doses it was assumed that people lived all year round in the settlements, that those in the vicinity of Lake Balapan spent one hour per day close to the lake and that those in the vicinity of Ground Zero spent one hour per day at the location. It was also assumed that meat, offal and milk are consumed from animals that take 10% of their total feed from the lake area or Ground Zero area. The results of this assessment of annual doses are shown in Table 10.10. The preliminary conclusions are: 9 no more detailed assessment of the radiological situation is required, because the doses today to local populations in the settlements are very low; 9 access to land with high dose rates within the polygon, namely Lake Balapan and Ground Zero, should be restricted in order to prevent reoccupation; and 9 further specific and systematic studies are needed on the plutonium levels in the soil around Lake Balapan and Ground Zero, and on the levels of radionuclides in drinking water sources of the settlements outside the polygon.
Chapter 10
530
Table 10.10 Estimated annual doses to local population (IAEA Yearbook, 1995). Exposure pathway
External gamma
Estimated annual adult dose (mSv) Settlement
Lake Balapan
Ground Zero
0.009
10.95
10.95
Inhalation
0.001
0.79
0.59
Ingestion
0.043
2.19
1.84
Total dose
0.053
13.9
13.4
The second of these studies has recommenced because, although the expert team was told that the explosion that created Lake Balapan was set off on hard rockwwhich would make it unlikely that radioactive nuclides would leach into the ground water sources of the settlementsmthis drinking water issue is not yet resolved. Similarly, the measurements and data gathered in the project were inadequate to establish the risks of plutonium from the five failed above-ground tests being resuspended and inhaled. Actinides were also released by nuclear tests conducted below the ground, and the directions and deposition patterns of their plume paths have not been identified. Plutonium doses to individuals today depend on their habits and particularly on the time they spend in contaminated areas. So the levels of radiation and radioactivity concentrations which correspond to an intervention level of 5 mSv in a year using a 100% occupancy factor are of interest. While the project obtained a reasonable understanding of the gamma dose rate situation, only limited information on actinides is available. An intervention level of 5 mSv a year corresponds to a dose rate of 0.5 ~tSv per hour, assuming 100% occupancy. Restriction to land with higher dose rates should be relatively straightforward, and could apply to Ground Zero and Lake Balapan. An appropriate criterion for limiting exposure to plutonium might be to restrict access to land contaminated above a few becquerels per gram. More restrictive standards are likely to result in unnecessary expenditure and possibly cause unnecessary anxiety among the local populations. The recommendation is that a systematic study be made of plutonium (and other actinides) in the soil outside a 1 km radius of the site of the above-ground tests. For additional references on this subject see LaRosa et al., (1996); Algazin et al. (1996) and Shoihet et al. (1996). The long-term consequences of a number of tests in Semipalatinsk/Altai Region have recently been summarised by Shapiro et al. (1998).
10.5 F R E N C H T E S T I N G SITES As the decision took shape to build an atomic bomb in the 1950s, the French began to look for a suitable test site. Possible locations included the Kerguelen Islands in the Indian Ocean, Clipperton Island and the Tuamotu Archipelago in the Pacific Ocean,
Bomb Test Sites
531
and French Algeria. Clipperton and Tuamotu were ruled out for lack of an airfield. The Kerguelen Islands were too far away and had poor weather. This left French Algeria (Aillert, 1968). In July 1957 the Reggane site was chosen, and as discussed above, in April 1958 the French government set a goal to conduct its first nuclear test in the first quarter of 1960. To help prepare for this test, several French delegations visited the U.S. Nevada Test Site (NTS)in 1957 and 1958 to witness and participate in U.S. nuclear tests. These visits provided an orientation in nuclear test effects, culminating in the French participation in the U.S. atmospheric test Smoky on 31 August 1957, at which the French tested a selection of their underground personnel shelters, equipment, and test instrumentation. The first French nuclear test, code-named Gerboise B leue, occurred on 13 February 1960 from a 344-ft (105-m) tower south-west of Reggane, in the Tanezrouft desert of Algeria. While the base Headquarters (0:17 East, 26:42 North) was near Reggane, the detonation sites were some 48 km to the south-west, closer to Hammoudia. At 60-70 kt, the yield of this plutonium device was three times that of the first devices tested by the U.S. or Britain. Three additional but less powerful atmospheric tests were conducted at the Reggane site in 1960 and 1961. All were plutonium fission devices, detonated from towers, and studied for their weapons effects. Following each of these tests, neighbouring African countries protested, some even going so far as to temporarily break off diplomatic relations with France (Gouldschmidt, 1968). These first French tests, moreover, were held during a U.S.-Soviet British testing moratorium that began in the fall of 1958 and lasted until September 1961. Following the first four atmospheric tests, the French moved their testing programme underground. Thirteen tests were carried out from 1961 to 1966 in the Taourirt Tan Afella granite intrusive (also called the Hoggar Massif) at In Ecker. The In Ecker Proving Grounds were located about 560 kilometers south-east of Reggane, in the southern part of Algeria (5:03 East, 24:03 North). Each nuclear device was placed at the end of a spiral-shaped tunnel dug into the rock. Safety doors were installed at various intervals to reduce the venting of gases during the explosions. The yield of these tests varied greatly, ranging between 3.6 kt and 127 kt. The military appropriations covering the 1960-1965 period stated that the goal of the nuclear development program was "the creation of a first system of operational nuclear weapons consisting of Mirage IV bombers carrying a fission bomb (the AN 11) with a power equivalent to 50 kt." The underground tests purportedly involved the miniaturisation of the AN 11 bomb (a prototype was successfully tested on 1 May 1962) (Gouldschmidt, 1968), as well as investigating the potential peaceful applications of nuclear explosives. After Algeria gained independence in July 1962, France had little choice but to move its nuclear test program (following completion of the underground test series). Later that year, the Pacific Test Site (Centre d'Exp6rimentations du Pacifique, DEP) was officially established. France chose the uninhabited atolls of Mururoa (originally the island was called Moruroa, the local traditional name, which in the Maohi language
Chapter 10
532
of Polynesia means Place of the Great Secret. However, it was changed to Mururoa by the French military in the 1960s) and Fangataufa in the Tuamotu Archipelago in the Pacific Ocean. The Tuamotu Archipelago is one of five archipelagos making up French Polynesia and is comprised of about 80 Tuamotuan atolls. Located in the extreme south-east comer of the Tuamotu Archipelago are the small uninhabited atolls of Mururoa and Fangataufa. These atolls are located about 1200 km from Tahiti. Mururoa and Fangataufa atolls are situated at 21~ 138~ and 22~ 138~ respectively, in French Polynesia in the South Pacific Ocean, about halfway between Australia and South America (Fig. 10.17). The two atolls belong to the Pitcaim-Gambier island chain at the south-eastern extremity of the Tuamotu Archipelago which consists of 76 atolls. Polynesia ranges from New Zealand, 4800 km from Mururoa, to Hawaii. The two atolls lie towards the eastern boundary of French Polynesia which comprises five archipelagos of about 130 atolls and reef islands in all. It seems likely that some atolls of the Tuamotu Archipelago may have been settled for more than 1000 years. However, habitation of the atolls of Polynesian archipelagos has been discontinuous and populations are generally sparse. Around 8000 people at present live within 1000 km of Mururoa Atoll. Tureia Atoll, 130 km distant, is the closest inhabited land, with a population of around 120 living a semi-subsistence lifestyle; that is, depending of fish and seafood gathered from the ocean and the atoll's lagoon, produce harvested from a small area of cultivated land and some imported food. Oo
J-
l
7" / / .*.
pAol 1A
/ /'
. "-I
/
i
'
.~'. "
* *
t
i *'** ] * ~'
~~
\ F"'J"
Society Islands
:! i
",~,
i Tropic of / ~ ~ \'~, x.'~,~ i Cancer ! "-' "~ \ z : : Z : ~ -
i0Hawaiian . ;o
~'~-,,....~% \ Galapagos Ga!apagos
Islands
i
lslanos
/) A
0~
o "N~,r..x'~,~--) :, ~l-J. Marqu01sas ---', I~;lands ' (.]\ -..s ~ ~,.Solomon xr~. 9 -- *'~ .~ 9 i. .. [/ "x-'3 /(~ , I * r"-'--" ,&.~ ! FP~ENCH I SO6~H (....~l) ~,* *" . . . ~ l , , ~..p.~+,"oj~ ~OINNESIA AMI~RI~ -~ X'~\ o~,~** .~ometyisaanos f-o~ , ~ // / x __ * . 9 p ~~R ~ o \ ~*~:a 'I " . !"I.I, Tah,t,'"/ ~ . ," ~i: ./ T;~rPicsf/Tro ic o f ? 9 ' , Gam ,er Islandss ' capncom AUS A ! Austral ~[~lands "** SOUTH PAC FI( t ~ 3 ~~
~ ~ ~ E ! Z E*~A L A N I ) \
140~
180 ~
Fig.
10.17.
~/~
140 ~
3 / /
AN
/
/ ~ ~ /
100 ~
Location of French Polynesia, South Pacific Ocean.
"
60~
Bomb Test Sites
533
These sites, originally chosen because of their isolation, were thought to be especially suitable for atmospheric tests. However, both atolls are surrounded to the west, north, and east by inhabited islands. In May 1966 the CEP promised to detonate bombs only when the winds were blowing to the south where there are no islands. In 1964, Mururoa and Fangataufa atolls were ceded by French Polynesia to the French Government for the Centre d'Experimentations du Pacifique (CEP). The legislation to establish the CEP was enacted by the French Parliament on 29 March 1963, and development of the Mururoa site began almost immediately. The first civilian workers, from Anaa Atoll in the Tuamotu Archipelago, were engaged on 7 September 1964. Altogether, some 57,750 people, including many French Polynesians, worked on Mururoa during the French test programme between 1964 and 1996. All infrastructures associated with the weapon test site~living quarters, buildings, laboratories, harbour facilities and an airstrip~were on Mururoa Atoll. The residential zone was at Anemone, at the eastern end of the atoll. As the rim of Mururoa is only a metre or so above sea level at high tide, a sea wall 4 m high was built in the 1980s as a storm barrier. Fangataufa Atoll, which had originally been envisaged as an observation post, was used mainly for larger tests. The Hao atoll initially served as a rear base where the nuclear test devices were assembled. The components were flown from France on planes that were refuelled in Martinique to avoid altogether the densely populated Tahiti Hao, a bigger atoll than Mururoa, which is located 450 km north-west of Mururoa and 900 km east of Tahiti. Built by the military, Hao has one of the longest runways in the South Pacific (3600 m) plus a large number of storehouses and workshops. Following the construction of a runway on Mururoa, the nuclear device assembly facility (Centre Technique CEA/DAM) at Hao was deactivated and transferred to Mururoa. A number of nearby atolls continue to provide logistical support (including security) to the CEP. Along with Mururoa and Fangataufa, these peripheral stations (the atolls of Tureia Tematangi, and Reao) are collectively known as the Base InterarmEes des Sites (BIA). A maximum of 3600 people (military personnel, scientists, and engineers) are present at BIA during the testing period, with a minimum of 3000 otherwise (of which about 1500 are military personnel). An additional 1000 military personnel of the three services are based in Tahiti at Papeete, Faaa, Aru~, and Mahina. Tahiti also served as a rear base for rest and recreation. France had conducted atmospheric nuclear weapon tests at the CEP site between July 1966 and September 1974 and underground nuclear tests from June 1975 up to July 1991. In June 1995, the French Government announced that it would carry out a final series of eight underground tests at the CEP site before acceding to a comprehensive nuclear test ban treaty, then under negotiation at the UN Conference on Disarmament in Geneva. After five tests of the series had been carried out at Mururoa and one at Fangataufa, the complete cessation of French nuclear weapon testing was announced on 29 January 1996. The Mururoa base has now largely been dismantled. The only structures that will remain are the sea wall, the harbour, the airstrip and three concrete blockhouses (and
Chapter 10
534
two more on Fangataufa), too massive to remove, built to protect observers of the atmospheric tests. French programmes to monitor the levels of certain radionuclides in the environment of Mururoa and Fangataufa atolls and in the carbonate rock and to monitor the stability of the coral rim of the atolls are being maintained, and automated equipment is being kept on site for use in geological and radiological surveys. These continuing measures will be supplemented by annual programmes of sampling from the physical environment (air, soil, water, sediment) and the biological domain (plants, fish, plankton, shellfish). In all, 193 expgriences nuclgaires (nuclear tests and safety trials) were conducted at the French nuclear weapon test site at Mururoa and Fangataufa atolls. Of these, 178 were "nuclear tests", in which a nuclear device was exploded with a large release of fission and, in some cases, fusion energy; and 15 were "safety trials" in which more or less fully developed nuclear devices were subjected to simulated accident conditions and the nuclear weapon cores were destroyed by means of conventional explosives, with no o r ~ o n a few occasions~very small releases of fission energy. A total of 41 atmospheric nuclear tests were conducted in the open air between July 1966 and September 1974, 37 at Mururoa Atoll and four at Fangataufa Atoll, and 137 underground nuclear tests took place deep below the surface of the atolls between June 1975 and January 1996, 127 at Mururoa Atoll and ten at Fangataufa Atoll. The energy released from a nuclear explosion is measured in units of kilotonnes (kt) of trinitrotoluene (TNT) equivalent, defined to be 1012 cal, i.e. 4.184• joules of explosive energy. The explosive yield of all tests at the CEP site (atmospheric and underground) was equivalent to about 13 000 kt; about 10 000 kt from atmospheric tests and about 3000 kt from underground tests. A summary of the atmospheric testing programme at CEP is given in Table 10.1 I. The first nuclear test at the CEP, on 2 July 1966, was of a 28 kt device mounted on a barge in the lagoon of Mururoa. Four barge-mounted tests were carried out in all, three Table 10.11 French atmospheric nuclear tests and safety trials at Mururoa and Fangataufa Atolls Date
Name
Type
Height (m)
Yield (kt) Fission
2 July 1966
Ald6baran
Barge
19 July 1966
Tamour6
Air drop
21 July 1966
Ganymbde
Safety trial
11 September 1966
Betelgeuse
Balloon
24 September 1966
Rigel a
Barge
Total
0
28
28
1000
50
50
12
0
0
470
110
110
3
125
125 205
4 October
Sirius
Barge
10
205
5 June 1967
Altair
Balloon
295
15
15
27 June 1967
Antares
Balloon
340
120
120
535
Bomb Test Sites
Date
Name
Type
Height (m)
Yield (kt) Fission
2 July 1967
Arcturus
Barge
7 July 1968
Capella
Balloon
Total
0
22
22
463
115
115 450
15 July 1968
Castor
Balloon
650
450
3 August 1968
Pollux
Balloon
490
150
24 August 1968
Canopus a
Balloon
520
-
2600
8 September 1968
Procyon
Balloon
700
-
1280
15 May 1970
Androm~de
Balloon
220
13
22 May 1970
Cassiop6e
Balloon
500
-
224
30 May 1970
Dragon a
Balloon
500
-
945
24 June 1970
Eridan
Balloon
220
12
3 July 1970
Licorne
Balloon
500
-
27 July 1970
P6gase
Balloon
220
2 August 1970
Orion a
Balloon
400
-
72
6 August 1970
Toucan
Balloon
500
-
594
5 June 1971
Dione
Balloon
275
34 -
150
13
12 914
0.05
0.05
34
12 June 1971
Encelade
Balloon
450
4 July 1971
Japet
Balloon
230
9
440 9
8 August 1971
Phoebe
Balloon
230
4
4
14 August 1971
Rh6a
Balloon
480
25 June 1972
Umbriel
Balloon
230
0.5
955 0.5
30 June 1972
Titiana
Balloon
220
4
4
27 July 1972
ObEron
Balloon
220
6
6
31 July 1972
Ariel
Safety trial
21 July 1973
Euterpe
Balloon
220
28 July 1973
Melpom~ne
Balloon
270
0.05
0.05
18 August 1973
Pallas
Balloon
270
4
4
24 August 1973
Parthenope
Balloon
220
0.2
0.2
29 August 1973
Tamara
Air drop
-
6
6
13 September
Vesta
Safety trial
4
0
0
16 June 1974
Capricorne
Balloon
220
4
4
1 July 1974
B61ier
Safety trial
7 July 1974
G6maux
Balloon
10
0.001 11
5.6 312
0 -
0.001 11
0 150
17 July 1974
Centaure
Balloon
270
4
25 July 1974
Maquis
Air drop
250
8
8
28 July 1974
Pers6e
Safety trial
0.001
0.001
14 August 1974
Scorpion
Balloon
24 August 1974
Taureau
Balloon
270
14
14 September 1974
Verseau
Balloon
433
-
5.6 312
-
4
96 14 332
aAll tests were performed at Mururoa Atoll except for these four tests which were carried out at Fangataufa Atoll.
536
Chapter 10
at Mururoa and one in the lagoon of Fangataufa. These barge-mounted nuclear tests produced most of the residual radioactive material at present in the accessible environment of the atolls. Most of the atmospheric nuclear tests were carried out with the device suspended from a balloon some hundreds of metres above the surface of the lagoons. Over the next eight years, 34 such devices were exploded in the atmosphere (31 at Mururoa and three at Fangataufa). A further three tests were explosions of devices dropped from aircraft. In all cases the detonation altitude was sufficient for the fireball not to reach sea level thereby minimising the production of local fallout. The largest test was of a 2600 kt thermonuclear device detonated 520 m above Fangataufa lagoon in August 1968. In the last atmospheric test, on 14 September 1974, a 300 kt device was exploded 433 m above Mururoa. Of the 137 underground nuclear tests, 127 were conducted at Mururoa and ten at Fangataufa, between 1975 and 27 January 1996 (Figs. 10.18 and 10.19). The tests were carried out in the volcanic rock at the bottom of sealed vertical shafts drilled 500 m to 1100 m deep beneath the rims of the lagoons. Earlier tests were carried out in shafts drilled vertically from the rims of the atolls. From 1981, some t e s t s ~ a n d all tests from 1987 o n w a r d s ~ w e r e carried out under the lagoons. No underground test had a yield exceeding 150 kt and the total energy release associated with all underground testing
Fig. 10.18. Locations and yields of underground tests at Mururoa Atoll.
Bomb Test Sites
537
Area 2 8 tests Wmax. <150 kt W .... <750 kt
Area 1
2 tests <10 kt W .... <20 kt Wmax.
Fig. 10.19. L o c a t i o n s and yields o f u n d e r g r o u n d test at F a n g a t a u f a Atoll.
was about 3000 kt. The final test explosion was detonated under Fangataufa lagoon on 27 January 1996, after which all nuclear testing at the CEP ceased. Table 10.12 lists all French underground tests and safety trials at Mururoa and Fangataufa atolls (after IAEA, 1998). Safety trials were conducted to investigate the behaviour of the core of a nuclear device under simulated faulty detonation conditions. The core is destroyed by the conventional explosive detonation of such a device, with the production of finely divided plutonium and plutonium oxide which are widely dispersed if the test is not confined. Usually no fission takes place, though there was a very small fission energy release in three of the French underground safety trials. (Since there was some explosive yield, these three trials are sometimes counted as "nuclear tests" which would put the total number of underground nuclear tests at Mururoa and Fangataufa atolls at 140 rather than 137.) All of the 15 safety trials were carried out at Mururoa. Of the 15 safety trials, five were carried out in the atmosphere and ten underground. The five atmospheric safety trials were conducted between 1966 and 1974 on the surface at the northern tip of the atoll on the three motus of Colette, Ariel and Vesta. The ten underground safety trials were performed in the north-eastern part of Mururoa Atoll, three in vertical drilled shafts that penetrated from the rim into the volcanic rock.
Chapter 10
538
Table 10.12 French underground nuclear tests and safety trials at Mururoa and Fangataufa Atolls (after IAEA, 1998, and references therein) Year Fangataufa rim 1975
Date
Name
Category
Yield (kt)
5 Jun 26 Nov
Achille Hector
A B
23 17
Total Fangataufa rim
Fangataufa lagoon
2 tests
1988
30 Nov
Cycnos
C
103
1989
10 Jun 27 Nov
Cyuocps Lycos
C C
74 87
1990
26 Jun 14 Nov
Cypselos Hyrtacos
C C
100 118
1991
29 May
Periclymenos
C
106
1995
1 Oct
Ploutos
C
97
1996
27 Jan
Xouthos
C
46
Total Fangataufa lagoon Muroroa rim
40 kt
8 tests
731 kt
1976
3 Apr 11 Jul 22 Jul 30 Oct 5 Dec
Patrocle Menelas Calypso Ulysse A Astyanax
A B A A A
1 12 0a 1 1
1977
19 Feb 19 Mar 2 Apr 28 Jun 6 Jul 12 Jul 12 Nov 24 Nov 17 Dec
Ulysse B Nestor Oedipe Andromaque Ajax Clytemnestre Oreste Enee Laocoon
B C A A B A A C A
5 47 1 0 28 0a 16 50 12
1978
27 Feb 22 Mar 25 Mar 1 Jul 19 Jul 26 Jul 2 Nov
Polypheme Pylade Hecube Aanthos Ares Idomenee Schedios
A A A A B A A
1 12 1 1 2 4 3
14 Nov 30 Nov 17 Dec 19 Dec
Aphrodite Priam Eteocle Eumee
A C A B
0a
64 14 12
539
Bomb Test Sites
Year
Date
Name
1979
1 Mar 9 Mar 24 Mar 4 Apr 18 Jun 29 Jun 25 Jul 28 Mar 19 Nov 22 Nov
Penthesilee Philoctete Agapenor Polydore Pyrrhos Egisthe Tydee Palamede Chrysotemis Atre
8 14 8 6 4 28 112 14 1 4
1980
23 Feb 3 Mar 23 Mar 1 Apr 4 Apr 16 Jun 21 Jun 6 Jul 9 Jul 19 Jul 25 Nov 3 Dec
Thyeste Adraste Thesee Boros Pelops Euryple Ilus Chryses Leda Asios Laerte Diomede
1 11 78 18 2 26 9 5 0a 78 2 51
1981
27 Feb 6 Mar 28 Mar 8 Jul 11 Jul 18 Jul 3 Aug 6 Nov 11 Nov 5 Dec
Broteas Tyro Iphicles Lyncee Eryx Theras Agenor Leto Procles Cilix
A A B B A A C A B B
8 2 5 22 8 2 16 1 3 5
1982
20 Feb 24 Feb 23 Mar 31 Mar 27 Jun 1 Jul 21 Jul 27 Nov
Aerope Deiphobe Evenos Aeson Laodice Antilokos Pitane Procris
A A A A A C A A
3 1 1
25 Apr 18 Jun 20 Jul 3 Dec
Automedon Burisis Battos Linos
A A B A
1983
Category
Yield (kt)
0~
2 20 2 1 1 3 10 4
continued
Chapter 10
540
Table 10.12 continuation Year
Date
Name
Category
Yield (kt)
1984
8 May 12 Jun 27 Oct 1 Dec
Demophon Aristee Machaon Miletos
A B B A
22 2 3 1
1985
30 Apr 3 Jun 24 Oct 24 Nov
Cercyon Talaos Hero Zetes
B B A B
13 11 2 5
1986
26 Apr 6 May 27 May 10 Nov 6 Dec
Hyllos Ceto Sthenelos Hesione Peneleos
B A B A A
5 5 4 6 9
1989
25 Nov
Daunus
A
0a
1981
10 Apr 8 Dec
Clymene Cadmos
B B
8 15
1982
20 Mar 25 Jul
Rhesos Laios
B C
17 56
1983
19 Apr 25 May 28 Jun 4 Aug 7 Dec
Eurytos Cinyras Oxylos Carnabon Gyges
C C B C B
40 42 33 8 15
1984
12 May 16 Jun 2 Nov 6 Dec
Midas Echemos Acaste Memnon
C C C C
56 34 34 53
1985
8 May 7 Jun 26 Oct 26 Nov
Nisos Erginos Codros Megaree
C B C C
90 5 18 54
1986
30 May 12 Nov 10 Dec
Galatee Nauplios Circe
C B C
30 24 32
1987
5 May 20 May 6 Jun 21 Jun 23 Oct 5 Nov 19 Nov 29 Nov
Jocaste Lycomede Dirce Iphitos Helenos Pasiphae Pelee Danae
B C B C C B C B
5 30 3 15 51 18 62 3
83 tests
Total Mururoa rim Muroroa lagoon
976 kt
541
Bomb Test Sites
Year
Date
Name
Category
Yield (kt)
1988
11 May 25 May 16 Jun 23 Jun 25 Oct 5 Nov 23 Nov
Nelee Niobe Antigone Dejanire Acrisios Thrasymedes Pheres
C C A B A C C
20 82 5 30 2 47 45
1989
11 May 20 May 3 Jun 24 Oct 31 Oct 20 Nov
Epeios Tecmessa Nyctee Hypsipyle Erigone Tros
B A C C B B
16 2 20 24 20 28
1990
2 Jun 7 Jun 4 Jul 21 Nov
Telephe Megapenthes Anticlee Thoas
B B B C
30 4 18 36
1991
7 May 18 May 14 Jun 5 Jul 15 Jul
Melanippe Alcinoos Patthee Coronis Lycurgue
A C C A C
1 16 28 1 34
1995
5 Sep 27 Oct 21 Nov 27 Dec
Thetys Aeoytos Phegee Themisto
B C C C
8 39 17 21
Total Mururoa Lagoon
54 tests
1445 kt
aSafety trial without nuclear yield. Note: The name and yield category (A < 5 kt; B 5-20 kt; C 20-150 kt) are as done by the French authorities. The estimate of the yield of each test in kt of TNT equivalent is from (IAEA- 1998).
It was three of the safety trials that took place in the carbonate rock that had small releases of fission energy associated with them. Each safety trial reportedly involved 10 TBq of 239+24~ If the material used in these trials were "weapons grade" plutonium--say, 7% 24~ about 3.4 kg 239pu and 0.3 kg 24~ would have been dispersed into the atmosphere or, for the underground tests, retained in the cavity at the end of each trial. Virtually the same mass would remain in those trials in which there was some fissioning of the plutonium; however, even the relatively small release of fission energy would have melted the surrounding volcanic basalt backfill to produce some lava in which some of the residual plutonium would be trapped, thereby reducing its potential for migration.
542
Chapter 10
There are thus three main sources of the man-made residual radioactive material that is at present found at very low levels in the terrestrial and aquatic environments of Mururoa and Fangataufa atolls: 9 The a t m o s p h e r i c tests: The radioactive material resulting from the atmospheric tests is essentially all residual material from the four barge tests; there was very little local fallout from the explosion of any device suspended from a balloon, since the fireball did not touch the surface and radioactive material was drawn up into the atmosphere. Most of the residual material is plutonium found in the lagoon sediments immediately underneath where the barge tests were conducted, remains of the nuclear cores of the test devices. There is also present in the environment of the atolls global fallout from atmospheric nuclear tests conducted by other states, mainly in the Northern Hemisphere. 9 The u n d e r g r o u n d tests: Tritium originating from underground sources is measurable in the lagoon water of both atolls and there is some evidence to suggest that present 9~ levels may be slightly enhanced by material that has migrated from underground. 9 The s a f e t y trials: The safety trials conducted at the northern tip of Mururoa Atoll on the motus of Colette, Ariel and Vesta have left some particulate plutonium on the surface in that area, despite extensive cleanup operations carried out in 1982-1987 and in an adjacent sand bank in the lagoon. Next we shall describe in some detail the organisation of the environmental radiological surveillance in French Polynesia. The Office de Protection contre les Rayonnements Ionisants (OPRI, formerly SCPRI, the Service Central pour la Protection contre les Rayonnements Ionisants) is responsible to the Ministry of Health and to the Ministry of Labour and has an institutional role in protecting workers and the population against ionising radiation. OPRI operates a statutory programme of surveillance of the levels of radioactivity in the environment, both in France and in some overseas territories. It also has its headquarters at Le VEsinet, with public access to the data through Minitel. OPRI's programme is conceived to ensure that levels of environmental radioactivity are such that the radiation exposure of members of the public is well below the Community limits. The tasks of IPSN (Institut de Protection et de Sfiret6 Nucl6aire, D6partment de Protection de la Sant6 de l'Homme et de Dosim6trie) are rather in the field of research and in providing models and data in support of radiation protection policies. IPSN runs a world-wide programme of environmental radioactivity monitoring (Observatoire Mondial de la Radioactivit6) part of which is an important programme of foodstuff sampling and aerosol sampling in French Polynesia. In addition to the samples collected by IPSN in different Polynesian archipelagos, IPSN also analyses a similar number of samples provided by DIRCEN. Analysis is carried out in such a way that actual levels of radioactivity are measured and correspondingly the limits of detection pursued are far below levels which would be significant from the point of view of health. The overall programme aims at an accurate assessment of population doses.
Bomb Test Sites
543
DIRCEN (Direction des Centres d'Expdrimentations NuclEaires) in conjunction with CEA (Commissariat ?al'Energie Atomique) conducts the military operations and collects environmental samples both on the test sites (fixed equipment and local sampling) and in the environment at large. DIRCEN collects biological samples on archipelagos for which IPSN has no agent of its own. Such samples together with pelagic fish, plankton and seawater are collected by the DIRCEN vessel "Marara" from various archipelagos and sent to the IPSN laboratory at Tahiti. Only the DIRCEN/CEA laboratories at Mururoa and at Monthl~ry analyse samples taken at the test sites and within the limits of territorial waters around the test sites. The relationship between SMSRB (Service Mixte de Surveillance Radiologique et Biologique de l'homme et de l'environnement), which is in charge of the CIRCEN/CEA environmental monitoring programme, and the IPSN laboratories is historically very close. Each organisation's responsibilities and the relationship between them seem now to be well defined. The overall environmental radioactivity monitoring programme in French Polynesia is considered to be the conjunction of the programmes run by OPRI, IPSN and CIRCEN/CEA. Scientists from these French institutions have published many of their results in open literature. It would consume a great deal of space to review all of this literature. We shall mention only some of the published reports. Bodie et al. (1987) have reported on 9~ levels in teeth. In order to measure 9~ levels in Polynesians' teeth, samples were taken from individuals of different geographical origins and ages. The analyses demonstrated especially that 9~ activity was correlated with fallout activity due to the nuclear weapon tests carried out between 1952 and 1963, whatever the geographical origin of the populations. The two age groups showing the lowest activities are: (1) the group of individuals born before 1945 and (2) the group of children born between 1976 and 1986 (i.e. after the largest fallouts). The highest mean l e v e l ~ 3 0 mBq/g Ca, corresponding to the individuals born between 1956 and 1965~is about half the amounts detected in Europe for the corresponding age groups. Kobis de Saint-Chamas et al. (1991) have described the monitoring programme which included monitoring of 137Cs contents in coconuts from the whole French Polynesian territory. The methodology is described and the results of the 2589 samples collected for 1967 to 1988 are presented. The maximum content found since 1967 is 52 Bq kg -l for coconut water and 289 Bq kg -~ for coconut copra. The decrease of 137Cs content is constant without discontinuity, whatever the distance from the explosion sites. The committed dose equivalent from 137Csdelivered by coconut water and copra consumption represents only a few microsieverts a year. Bourlat (1991) reported some results of radioactivity measurements of 9~ 137Cs and 239+24~ in Polynesian oceanic waters (at the surface and at different depths) and in Mururoa and Fangataufa lagoon waters. These results come from annual measurements carried out by the SMSR during systematic monitoring of the ocean and lagoon waters, and also during special studies such as vertical profile determination in northern Mururoa in 1990.
544
Chapter 10
The water activity concentrations of 9~ and 137C9have approximately the same values inside and outside Mururoa and Fangatuafa lagoons: from 1.3 to 1.9 Bq/m 3 for 9~ and from 2.4 to 2.8 Bq/m 3for 137Cs. However, activity levels in 239+24~ are lower in oceanic waters (from 2 to 4 mBq/m 3) than in lagoon waters (0.4-0.5 Bq/m 3for Mururoa and 1 Bq/m 3 for Fangataufa). This is due to the slow solubilization of plutonium deposited in lagoon sediments by the atmospheric nuclear tests carried out in 1966. The differentiation in suspended particulate matter and soluble fraction (0.45 ~tm filtration) is only carried out in lagoon waters (total plutonium for oceanic waters). In the lagoons, the levels measured for the soluble fraction have been unchanged for several years. Activity concentrations in the suspended particulate matter are lower and do not go beyond an average of 10% (exceptionally 35%)of activity concentrations in the soluble fraction. The vertical profiles of the activity concentration of 9~ and 137Cs, in the polynesian oceanic waters, show a layer of homogeneous mixture (first 100-200 m), then a fast radioactivity decrease in the deep waters (less than 0.1 Bq/m 3 of 137Cs below 800 m). The water temperature profile is similar. The activity ratio of 137Cs/9~ lies between 1.6 and 1.9 for oceanic waters up to 300 m. This ratio is not determined for deeper waters, because measurement errors in 9~ determination become too high (see Fig. 10.20). The surface activity concentration of 239+24~ is stable (2-4 mBq/m3). Then it increases quickly to a maximum value at 600 m (13 mBq/m3), before decreasing quite quickly up to 1000 m (5 mBq/m3), and more slowly thereafter. This typical plutonium profile is usually observed in world oceanic waters (maximum activity between 500-1000 m). In the three samples (corresponding to the greater values of 239+24~ activity), in which 238pu was superior to the detection limit, the 239ptl/239+24~ activity ratio lies between 0.15 and 0.17 (usual values for these latitudes). This leads to the conclusion that the origin of the determined plutonium in the water samples is due to global fallout (atmospheric tests and SNAP). Under pressure from public opinion against nuclear testing, France has opened the Pacific test site. Since 1983, a number of scientific delegations have visited the Mururoa site, either by invitation of the French Government or at the request of SMSRB. 9 October 1983: Scientists from Australia, New Zealand and Papua New Guinea, headed by Prof. Atkinson of NRL Christchurch, (NZ); 9 June 1987: Cousteau Foundation representatives, assisted by the Marine Biogeochemistry Institute of Ecole Normal Superieure, Montrouge (France); 9 March 1991: Scientists from the International Atomic Energy Agency (MEL/ Monaco) and the Lawrence Livermore National Laboratory (California, USA); 9 October 1994: Scientists from IAEA/Monaco, Lund University (Sweden) and LLNL/USA), whose mission was to collect biological samples representative of the Mururoa site, to be used in an international comparison exercise. During these visits, a number of samples were taken by visiting scientists, part of which was analysed by SMSRB:
Bomb Test Sites
545
STRONTIUM 90
500
~
~
-
CESIUM 137
~l
:
'!,
500
i
.....................
i
. . . . . .
1000
...................................
1500
i
'~176176 i .i i i ...i
-
1500
2000
.........
2000 i
i
,
,
]
,
,
,
i
0.5
i
1 ACTIVITY
i
I
....
L. . . . - . . . . . . . . . . .
- .... - ......
I
I
I
I
,,
',
,,
,,
, ,
I. . . . . , . . . .
!
2
0
0.5
1
/ Bg/m 3
,. . . . . . 3
o
|
i-,
i
1500
.......
0.004
t 0.008
0.012
1000 't. . . . . . . . . . . . . . .
2000 t
0.0
. . . . . .
. . . . . . . . . ~- . . . . . .
137Cs,
239+24~
! ~ . . . . . . . . . . . .
. . . . . . . .
0
ACTIVITY / Bg/m 3 F i g . 10.20. 9~
Y
.J_
I
0.000
3.5
/ Bg/m 3
i
1500 t-- . . . . . .
2000 t __ .i....d---~ __
2.5
TEMPERATURE
500
..... i',
,..... , .... 2
ACTIVITY
50O
'~176176 ...... , : ~
,. . . . . .5
i
1.5
PLUTONIUM 239, 240 0
I ,
vertical profiles and temperature
5
I
10
. . . .
I .......
] .......
',
',
i a . . . . . . . i
15
. . . .
i .z . . . . . . . . I
20
. . . .
25
TEMPERATURE / celsius degrees p r o f i l e in t e r r i t o r i a l w a t e r s , n o r t h e r n
M u r u r o a in y e a r 1 9 9 0 ( a f t e r B o u r l a t et al., 1 9 9 1 ) .
9 soil, plants, crustaceans and shellfish (from the emerged parts of the atoll); 9 water, sediments, fish and shellfish (from the lagoon); 9 seawater and plankton (from the inshore oceanic environment). Burlat et al. (1995) have presented an overview of these four missions. It was concluded that artificial activity levels are rather low and have negligible impact on man. For
546
Chapter 10
further details see Atkinson (1984), Bourlat et al (1995), Foundation Consteen (1988), Martin et al. (1990), Bourlat and Martin (1992), Ballestra and Noshkin (1991 ), Bourlat et al. (1994), Ballestra et al. (1995). Bourlat et al. (1995) have presented results on the plutonium radioactivity levels in Mururoa lagoon water during the 1985-1991 period. The low radioactivity levels recorded, from 0.01 to 1.5 Bq/m 3 are due to the slow solubilization of plutonium deposited in lagoon sediments following atmospheric experiments which took place from 1966 to 1974. The average concentrations of the lagoon water decrease from one year to the next. Since the Mururoa lagoon is open to the ocean, plutonium radioactivity traces are also detectable in the immediate vicinity of the atoll. The atmospheric nuclear tests which took place at the Mururoa Lagoon between 1966 and 1974 did not produce significant fallout on the emerged part of the atoll. The tests (three carried out on a barge) only affected the sediments at the bottom of the lagoon, especially in the area directly under the barge. In contrast, a few safety air tests that took place between 1966 and 1974 in the northern part of the atoll (between the Colette Motu and the Denise area) resulted in localised plutonium deposition on the coral bedrock. Immediately after each of these safety tests, the deposited plutonium was either fixed by asphalt surfacing or removed after each individual test. The asphalt surface gradually deteriorated with time, especially after the heavy tropical depressions of 1981. There was a risk that it would become detached from its support. For this reason, it was decided to proceed with a general radioactivity cleanup of the area involved. From 1981 to 1986, action was taken to eliminate all movable debris and to break up the concrete slabs. Later, from January to July 1986, scraping of the coral bedrock with bulldozers and removal of the debris by suction was undertaken to reduce surface activity, which had exceeded 2.108 Bq/m 2 in places, to a value below 10 6 Bq/m 2, a threshold comparable to those accepted for Palomares (Iranzo et al., 1987) and Enewetak (Friesen et al., 1982). The debris was buffed in the basalt section of a well, bored for that purpose, and sealed off by cement plugs. During this operation, an approximate 1000 m 3 of debris was buried, corresponding to the cleaning of a 50,000 2 m heterogeneous area and to a total activity of 2x10 ~2 Bq. Between August and October of the same year (1987), coral aggregates of low-level activity, initially dumped on an open site in the Denise area, were sorted and eliminated. The above cleanup operations have led to the elimination of the main sources of plutonium activity on the ground. The extremely high sensitivity for the detection of atmospheric plutonium (of the order of 10.8 Bq/m 3) has made it possible to measure the re-suspension of 239'24~ on the Mururoa Atoll, as a result of the cleanup work. This was addressed by Bourlat et al. (1994); they indicated that the yearly 239'24~ air concentrations increased by a factor of 6 due to re-suspension during cleanup operations at Mururoa. Figure 10.21 shows the 239'24~ radioactivity fluctuations at the two measurement stations: Kathie and Martine. Plutonium measurements made on monthly composites have shown that the average yearly level of 239'24~ increased by a factor of 6 from 1986 to 1987 (4• 10-8 Bq/m 3, then 2.4x 10-7 Bq/m3). The concentrations fluctuated
547
Bomb Test Sites
10
I
!
!
I
I
KATHIE 1
m
Z 9 [..-,
Z
0.1
r,..)
m
Z 9 0.01 el
m m m
m m
i
0.001 1986
10
I 1987
I
B
I 1988
I
I 1989
I
I 1990
I
1991
I
_
MARTINE o-' m.
1 m
Z 9 k-,
Z
0.1
L)
Z 9
L) 0.01 z o, eq
I
0.001
I 1987
Fig. 10.21.
239+24~
I 1988
I 1989
i 1990
1991
radioactivity fluctuations at Martine Station (after Bourlat et al., 1991).
548
Chapter 10
monthly between 10-s Bq/m 3 and 10-6 Bq/m 3, with a yearly decrease from 1987. In 1991,239'24~ concentrations were close to 7x 10-8 Bq/m 3. The above activity levels measured at Mururoa are comparable to the levels observed world-wide in the early 1980s (with the exception of a few specific sites). These concentrations result from earlier safety tests made on the motus in the northern zone of the atoll from 1966 to 1974, and from the cleanup work undertaken in the years 1981-1987. This work resulted in a great reduction of radioactivity on the ground, but produced a slight temporary increase of plutonium in the air. The French nuclear testing programme has attracted international attention. France is a founding member of European Union, and party to the European Treaty. Article 35 of the Euratom Treaty requires that each Member State shall establish the facilities necessary to carry out continuous monitoring of the level of radioactivity in air, water and soil and to ensure compliance with the basic standards. Article 35 also gives the Commission the right of access to such facilities in order that it may verify their operation and efficiency. For the Commission, the Directorate-General Environment, Nuclear Safety and Civil Protection (DG XI) is responsible for undertaking the verifications. The verification visit to French Polynesia was made as a result of concerns about the resumption of French nuclear testing. It took place at very short notice from 18-29 September 1995, in immediate response to the letter of M. de Charette to President Santer and M. Barnier to Commissioner Bjerregaard dated 13 September 1995, confirming the French authorities were ready to welcome the experts of the Commission for a verification visit to French Polynesia. The plan of verification visits was put together and here is part of the report (EU Report 5037): "It is the Commission's view that all facilities pertaining to the environmental radioactivity monitoring in French Polynesia are within the scope of verification, including such facilities which are necessary for the assessment of the environmental impact of sources of radioactive contamination at the sites where nuclear testing is carried out. As described in the overview of the surveillance system put in place in French Polynesia (document: "Synth~se de la surveillance effecture par I'OPRI en Polynrsie Franqaise depuis 1980") a distinction is made between two complementary types of surveillance, with a view on the one hand to ensuring adequate protection for the whole of French Polynesia and on the other hand to concentrating specific means of investigation on what can be qualified as the "source term" of possible radioactive pollution of the area concerned, i.e. the atolls on which the tests are carried out. Hence it was the Commission' s view that a complete verification mission required access for verification purposes to the atolls of Mururoa and Fangataufa. Access was granted to Mururoa but the French authorities held the view that this invitation was solely for the purposes of information, not of verification. Access was denied to Fangataufa for reasons of "defence security". Access was also denied to military premises at Faaa (Tahiti Airport) in which facilities for aerosol sampling and gamma monitoring are installed. While the last-mentioned facilities are of less importance because they duplicate similar facilities nearby, the denial of verification access at
Bomb Test Sites
549
Mururoa and Fangataufa does not allow a comprehensive assessment of the functioning and adequacy of facilities relevant to the monitoring of sources of radioactive contamination in French Polynesia. Moreover, the Commission has subsequently been denied verification access to the DIRCEN/CEA laboratory at Monthl6ry (France)". This was September 1995. The French testing programme was not yet finished. Mururoa has been the subject of three independent scientific missions during the 1980s. These studies were not allowed complete freedom to take samples and the information available to them was limited. The general conclusions were that there were low concentrations of radioisotopes in the environment consistent with the levels expected from global fallout from atmospheric tests in both the northern and southern hemispheres (which ceased in 1980). There was insufficient evidence available for the missions to form any firm conclusions about the likelihood and rates of leakage of radioisotopes from the underground tests. Nevertheless, the missions found that by the mid-1980s there had been some visible damage to Mururoa atoll, in the form of fissures, subsidence and submarine landslides off the side of the atoll. Moreover, the tests, which are conducted between 600 and 1200 m underground in volcanic rock, by their very nature result in localised fracturing of the rock around each test site. A substantial number of tests have been conducted since the most recent of the missions, which will have added to the structural changes observed in the atolls. The scientific concerns arising from underground testing at Mururoa relate to the structural integrity of the atoll, and the release of radioactive materials into the surrounding ocean and biosphere, through leakage of radioisotopes from the atoll structure and through venting of gaseous and volatile radioisotopes (Pitman, 1995). The meeting of South Pacific Environment Ministers (1995) resulted in the requirements that: "More information is required to define and understand the structural integrity of the atolls and to assess the timing and scale of any leakage of radioactivity. The French authorities have a considerable database of geoscientific, environmental and other relevant information, built up over two decades of monitoring in French Polynesia. France should release this (currently confidential) information to the international scientific community for independent consideration and analysis, as a matter of priority. Moreover, it should allow international scientists to have unfettered access to the atolls before, during and after the proposed program of eight tests, to obtain independent samples and conduct experiments. In the longer term, there is a need to continue to monitor the atolls, and international scientists should be allowed to participate in, and publish outcomes of, such long-term monitoring activities." In August 1995, following the announcement in June 1995 of a final series of eight underground nuclear tests at the South Pacific site, the French Government submitted a written request to the IAEA to assess independently the radiological consequences of the nuclear tests at Mururoa and Fangataufa atolls, and undertook to provide information needed for the assessment.
550
Chapter 10
The IAEA undertook to organise and manage a study of the radiological conditions at Mururoa and Fangataufa atolls as a. consequence of all the experiences nucl~aires, upon the conclusion of nuclear testing. An International Advisory Committee (IAC) of experts was convened by the IAEA to provide scientific guidance and direction to the IAEA in the conduct of the Study and to report on the Study' s findings, conclusions and recommendations. The IAC comprised scientists from ten IAEA Member States and ex officio experts nominated by the European Commission, the South Pacific Forum, the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) and the World Health Organisation (WHO). For a detailed scientific account of the findings and the conclusions and recommendation of the Study, the reader is referred to the IAC's Main Report and its supporting Technical Report prepared for the Study of the Radiological Situation at the Atolls of Muroroa and Fangataufa. The reports resulting from the Study, which have been issued by the IAEA in its Radiological Assessment Reports Series, are: 9 the Main Report, which incorporates an Executive Summary that is also available separately: the Main Report is the primary publication of the Study; 9 a Technical Report in six volumes: Radionuclide Concentrations Measured in the Terrestrial Environment of the Atolls; Radionuclide Concentrations Measured in the Aquatic Environment of the Atolls; Inventory of Radionuclides Underground at the Atolls; Releases to the Biosphere of Radionuclides from Underground Clear Weapon Tests at the Atolls; Transport of Radioactive Material within the Marine Environment; and Doses due to Radioactive Materials Present in the Environment or Released from the Atolls; 9 the Summary Report: the Summary Report presents a summary of the Main Report together with its findings, conclusions and recommendations for the benefit of a wider audience. The IAEA has a statutory responsibility to establish safety standards for protection against exposure to ionising radiation, and its statute also authorises its Secretariat to provide for the application of these standards upon the request of a State. At the request of Member States, the IAEA carries out reviews and assessments of the radiological conditions of areas affected by previous nuclear activities and events. The IAEA's safety standards are established for the peaceful uses of nuclear energy and its byproducts. Nevertheless, their basic protection criteria are also applicable to exposure to ionising radiation due to existing radioactive residues, such as in the particular radiological conditions at Mururoa and Fangataufa atolls. The Study therefore used as its principal international authority on radiation protection matters the interagency International Basic Safety Standards for Protection against Ionising Radiation and for the Safety of Radiation Sources. The Basic Safety Standards were mainly used to establish the criteria for use in assessing whether the radiological conditions at Mururoa and Fangataufa atolls represent any hazard to people presently in the region or who may reside on the atolls, and in making recommendations on any monitoring or remedial action that might be required.
Bomb Test Sites
551
The information provided by the French Government for the Study was extensive and assessed to be of high quality. Most of it was newly available and had not been subject to the scrutiny of the scientific community, so it was checked and independently evaluated where this was possible. In particular, the IAC decided: to conduct an independent environmental sampling and surveillance campaign to measure present levels of radioactive material in the environment of the atolls, so as to provide a basis for the evaluation of the French data from environmental monitoring; and to carry out its own sampling and analysis of the water in the cavity-chimney resulting from two nuclear tests, and in the carbonate rock of Mururoa and Fangataufa atolls. A major part of the total investigative effort in the Study was associated with assessing the possible migration of radioactive material from its underground repositories and passage into the human food chain. Predictive models were developed to estimate the rate of migration through the geosphere, mixing in the lagoons and dispersion in the ocean. Information from a number of sources, including the open scientific literature, was consulted in addition to that provided by the French authorities. Reports on three independent scientific missions to the atolls that are available in the public domain were consulted~the Scientific Mission of French Polynesia (Tazieff mission) of June 1982, the New Zealand, Australia and Papua New Guinea Scientific Mission of Mururoa Atoll (Atkinson mission) of October 1983, and the Scientific Mission of the "Calypso" to the Site of the Nuclear Experiments at Mururoa Atoll (Cousteau mission) of June 1987. Much information of relevance for the environment of the two atolls on the distributions and amounts of residual radionuclides is available in the results of the French monitoring programmes. For the reasons given above, the IAC decided to conduct an independent environmental sampling and surveillance campaign at the two atolls. The survey campaign was conducted at the atolls in July 1996. The campaign comprised a terrestrial part co-ordinated by the IAEA Seibersdorf Laboratory and an aquatic part co-ordinated by the IEA Marine environment Laboratory in Monaco. The terrestrial sampling and surveillance campaign was focused on Mururoa and Fangataufa atolls but at the time of the atmospheric nuclear testing, there was also some deposition of radionuclides on neighbouring islands, most notably at Tureia Atoll. Levels of activity were therefore measured at Tureia Atoll in order to calculate present dose rates for the inhabitants. In the terrestrial sampling and surveillance campaign, some 300 samples were collected (of vegetation, coconuts, sand, top soil, corals, cores of coral bedrock and aerosols) and analysed in ten laboratories in nine countries and at the IAEA Seibersdorf Laboratory. Over 1000 radioanalytical determinations were made on the samples. In addition, a large number of in situ gamma spectrometric measurements were carried out, particularly on the motu of Colette at the northern tip of Mururoa, near where the five atmospheric safety trials had been conducted (i.e. on the motus of Colette, Ariel and Vesta). Sand, coral and coral bedrock from this area were also examined for residual plutonium-containing particles.
552
Chapter 10
The aquatic sampling and surveillance campaign was conducted over four weeks using five naval vessels. Gamma spectrometric surveys were made of the sea bed in order to optimise sampling. Over 300 samples were collected (from lagoon water, ocean water, sediment pore water, sediment, corals and biota). Some 13 000 litres of water and 1 tonne of solid samples were collected, processed, packaged and transported to Monaco for distribution to analytical laboratories. A separate underground water sampling campaign was carried out at Mururoa and Fangataufa atolls in May-June 1997. Samples were taken from two test cavities on Mururoa and from nine monitoring wells in the carbonate rock. This is discussed later in the section on the "solution source term". The IAEA campaign corroborated the extensive data already available and provided additional scientific information. The activity concentrations of radionuclides in the terrestrial and aquatic environments are generally low and comparable with reported concentrations of the same radionuclides at similar atolls where no nuclear weapon testing took place. The general level of fixed plutonium contamination from the safety trials remaining in the motus of Colette, Ariel and Vesta following extensive cleaning up by the French authorities was estimated to be perhaps three times the level of the French cleanup criterion of 106 Bq/m 2. The significance of this as an inhalation hazard is discussed later. The Study survey did, however, find a number of small particles containing plutonium and americium in the area, and in a lagoon sandbank adjacent to Colette. The radiological risk associated with these residual particles is also discussed later. From the information provided by the French Liaison Office and the Study results taken together, there is evidence of a time trend in the concentrations in lagoon water of isotopes of hydrogen, strontium, caesium and plutonium: tritium (3H), 9~ 137Cs and 239+24~ Confinement was not equally effective for all 137 underground nuclear tests, and it is apparent from the elevated tritium concentrations that have persisted for some years that some leakage of tritium has been occurring into both lagoons. This was confirmed by information from the French Liaison Office on levels of tritium and its distribution in the carbonate rock of both atolls. Study measurements made in 1996 indicate that tritium levels in the lagoons may be beginning to fall (this decline may, however, be reversed for the lagoon at Mururoa as more tritium from underground nuclear tests migrates to the lagoon water: see next Section). The 9~ levels at present may be showing a slight increase. The levels of ~37Cs and 239+24~ have been falling for some years, faster than would be expected through radioactive decay alone. In summary, the study found that the terrestrial and aquatic environments of Mururoa and Fangataufa atolls that are accessible to people contain residual radioactive material attributable to the experiences nucl~aires, but at generally very low concentrations which the Study concluded were of no radiological significance. There are, however, some features of note whose radiological implications are examined at the end of the next chapter.
Bomb Test Sites
553
a. Several kilograms of plutonium resulting from the atmospheric nuclear tests carried out at the atolls remain in sediments under the lagoon of each atoll. Some of the plutonium in the sediments of the Mururoa Atoll lagoon came from the atmospheric safety trials. b. The concentration of tritium in each lagoon was found to be higher than in the open ocean, as the result of leakages from a number of the cavity-chimneys created by underground nuclear tests. c. Particles containing plutonium and small amounts of americium resulting from atmospheric safety trials remain in the area of the trial sites--the motus of Colette, Ariel and Vesta on Mururoa Atoll. The Study analysed these types of particles, found in samples of sand and coral collected from the surface of the motu of Colette and in sand taken from a sandbank adjacent to it. d. Elevated levels of 137Cs were found over small areas totalling several hectares on the Kilo-Empereur rim of Fangataufa Atoll. The conclusions and recommendation of this Study were as follows: 1. The Study concluded that there will be no radiation health effects which could be either medically diagnosed in an individual or epidemiologically discerned in a group of people and which would be attributable to the estimated radiation doses that are now being received or that would be received in the future by people as a result of the residual radioactive material at Mururoa and Fangataufa atolls. 2. Nevertheless, the Study noted that the reported cancer incidence in populations in the South Pacific region and throughout the world is changing for a number of reasons, including: the improved diagnosis and registration of cancer cases; modifications in environmental exposure to cancer-causing agents and in personal habits (such as dietary and smoking habits); population migrations that alter baseline cancer incidence rates; and changes in the incidence of other diseases. The Study emphasised, however, that at the very low levels of dose estimated in the Study there will be no changes in cancer incidence rates in the region attributable to radiation exposure caused by the residual radioactive material at Mururoa and Fangataufa atolls. 3. The Study assessed the dose rates to native biota resulting fromthe residual radioactive material at Mururoa and Fangataufa atolls and, in the great majority of cases, found them to be similar to or lower than dose rates due to natural radiation sources. An exception is the potentially high dose rates that could be experienced by individual members of some species owing to plutonium contained in particulates--for example, from the sediment of the sandbank adjacent to the Colette motu in the northern part of Mururoa Atoll. Overall, the Study concluded that the expected radiation dose rates and modes of exposure are such that no effects on biota population groups could arise, although occasionally individual members of species might be harmed, but not to the extent of endangering the whole species or creating imbalances between species.
554
Chapter 10
4. Given the measured and predicted radionuclide activity levels, and the low dose levels estimated for the present and for the future, and with account taken of international guidance, the Study concluded that no remedial action at Mururoa and Fangataufa atolls is needed on radiological protection grounds, either now or in the future. 5. Similarly, the Study concluded that no further environmental monitoring at Mururoa and Fangatuafa atolls is needed for purposes of radiological protection. 6. Although many assumptions were made in the modelling of systems, the findings are robust: i.e. the Study concluded that the expected extent of changes in the conclusions due to uncertainties in the parameters used in the modelling is slight. Furthermore, the predicted doses are so low that large errors (even of an order of magnitude) would not affect the conclusions. 7. The study noted that a scientific programme of monitoring of the radionuclide concentrations in the carbonate formations and in the nuclear test cavitychimneys is under way at Mururoa and Fangataufa atolls. Should this programme continue, the Study recommends that emphasis be placed on monitoring the migration behaviour of long-lived and relatively mobile radionuclides and radiocolloids because of its particular scientific interest. The scientific programme, supplemented by some monitoring of radionuclide levels in the biosphere, may also be useful in assuring the public about the continuing radiological safety of the atolls.
REFERENCES Aki, K. and Richards, P.G., Quantitative SeismologymTheory and Methods. W.H. Freeman and Company, San Francisco, 1980. Algazin, A.I., Kalyado, V.B., Shoikhet, Ya.N., Demin, V.F., Zakharchenko, I.E., Ilyin, L.A., Kiselyov, V.I. and Zaitsev, E.V., Assessment of radiological consequences of the nuclear weapon tests on the Semipalatinsk test site for population of Altai region. IAEA-SM-339/46P, page 171, Vienna, Austria, 1996. Atkinson, H.R., Davis, J.P., Davy, D.R., Hill, L. and McEwan, A.C., Report of New Zealand, Australian and Papua New Guinea Scientific Mission to Mururoa Atoll, 1983. Atkinson, H.R., Report of a New Zealand, Australian and Papua New Guinea-Scientific Mission to Mururoa Atoll, October/November 1983. Published by the Ministry of Foreign Affairs, New Zealand (1984). Badie, C., Arnould, C., Sarbach, J., Arnaud, M., Lemercier, R., Bernard, C., Howell, P. and Texier, T., Teneur en 9~ de deuts humaines collect6es en Polyn6sie Franqaise. Radioprotection, 22 (1987) 325-332. Ballestra, S., Baxter, M.S., Gaustaud, J., Pettersson, H., Provinec, P.P. and Vas, D., Intercomparison report on radionuclide measurements on biological samples collected at Mururoa atoll. IAEA/AL/085, IAEA-MEL Report no. 60, August 1995. Ballestra, S. and Noshkin, V., Summary of radionuclide intercomparison results in seawater and plankton collected outside Mururoa atoll. IAEA/AL/044, IAEA-ILMR Report no. 48, July 1991. Baurlat, Y., Millies-Lacroix, J.-C., Dunoyer, B., Plutonium-239,240 atmospheric radioactivity measurements at Mururoa from 1986 to 1991. J. Radioanal. Nucl. Chem., 177 (1994) 107-120.
Bomb Test Sites
555
Bennett, B.G., "Retrospective assessment of the French atmospheric tests and their contribution to exposures", report on Int. Conf. on the Radiological situation at the Atolls of Mururoa and Fangataufa, Vienna, Austria, 30 June-3 July, 1998. Berthet, C., Propagation atmosphErique lointaine de signaux acoustiques puissants--Th~se d'Etat, Paris, 1969. Bikini Atoll Rehabilitation Committee, BARC Resettlement of Bikini Atoll: Feasibility and Estimated Cost of Meeting the Federal Radiation Protection Standards, Rep. No. 1, submitted to the US Congress, House and Senate Committees on Interior Appropriations, pursuant to Public Law 97-257 (15 November 1984). Bikini Atoll Rehabilitation Committee, Resettlement of Bikini Atoll: Feasibility and Estimated Cost of Meeting the Federal Radiation Protection Standards, Interim Report, submitted to the US Congress, House and Senate Committees on Interior Appropriations, pursuant to Public Law 97-257 (23 November 1983). Blanc, E. and Jacobson, A.R., Observations de perturbations produites par des explosions chimiques puissantes dans l'ionosphEre. AGARD, SP 485, 1990. Blanc, E. and Rickel, D., Nonlinear wave fronts and ionospheric irregularities observed by HF sounding over a powerful acoustic source. Radio Sci., 24 (3) (1989) 279-288. Bolt, B.A., Earthquakes. W.H. Freeman, New York, 1988. Bolt, B.A., Nuclear Explosions and Earthquakes. The Parted Veil. W.H. Freeman, New York, 1976. Bouchez, J., and Lecomte, R., The Atolls of Mururoa and Fangataufa (French Polynesia)--Nuclear Testing. CEA/DAM, 1996. Bourlat, Y. and Martin, G., Precise determination of the concentration of radiocesium in the water of Mururoa lagoon. J. Environ. Radioactivity, 17 (1992) 13-29. Bourlat, Y., Millies-Lacroix, J.-C., Abt, D., Measurement of low-level radioactivity in the Modane underground laboratory. Nucl. Instrum. Meth. Phys. Res., A339 (1994) 309-317. Bourlat, Y., Millies-Lacroix, J.-C. and Martin, G., Results of measurements made by SMSRB on samples collected in Mururoa by international scientific delegations, ICRM Conference on Low Level Measurement Techniques Seville, Spain, 2-6 October 1995. Bourlat, Y., Millies-Lacroix, J.-C. and Nazard, R., Determination of plutonium radioactivity in Mururoa lagoon water. J. Radioanal. Nucl. Chem. Articles, 197 (2) (1995) 393-414. Bourlat, Y., Millies-Lacroix, J.-C. and Rancher, J., 239+24~ 137Cs,9~ contents in seawater in the vicinity of Mururoa and Fangataufa atolls and in French Polynesia, Int. Symp. on Radionuclides in the Study of Marine Processes, 9-14 September 1991, Norwich, U.K. Broche, P., Propagation des ondes acoustico-gravitationnelles excitEes par des explosions. Ann. GEophys., 33 (3) (1977) 281-288. Brown, J., Coiling, A., Park, D., Phillips, J., Rothery, D. and Wright, J., Ocean Circulation. The Open University, Milton Keynes, 1989. Buiques, D. Gachon, A. and Guille, G., L' atoll de Mururoa (PolynEsie franqaise): I. Structure et Evolution gEologique. Bull. Soc. GEol. France, 163 (5) (1992) 645-657. Bullen, K.E., An Introduction to the Theory of Seismology. Cambridge University Press, New York, 1980. Burns, P.A., Cooper, M.B., Duggleby, J.C., Mika, J.F. and Williams, G.A., Plutonium-contaminated fragments at the Taranaki site at Maralinga, ARL/TRO75, July 1986. Burns, P.A., Cooper, M.B., Williams, G.A. and Lokan, K.H., Radioactive contamination at Maralinga, in: Radiation Protection Practice (IRPA 7, April 1988), Vol. 3, pp. 1374-1377. Pergamon Press, Sydney, 1988. Carter, M.W. and Moghissi, A.A., Three decades of nuclear testing. Health Phys., 33 (1977) 55-71. Chaliand, G. and Jan, J., Atlas du NuclEaire Civil et Militaire des Origines ~i la ProlifEration. Polyot & Rivages, Paris, 1993. Chieco, N.A., Bogen, D.C. and Knutson, E.O. (eds.), EML Procedures Manual, USDOE Report HASL-300, 27th Edition, Vol. 1, 1990. Cooper, M.B., Burns, P.A., Tracy, B.L., Wilks, M.J. and Williams, G.A., Characterisation of plutonium contamination at the former nuclear weapons testing range at Maralinga in South Australia. J. Radioanal. Nucl. Chem., 177 (1994) 161.
556
Chapter 10
Cooper, M.B., Burns, P.A., Williams, G.A., Lokan, K.H. and Duggleby, J.C., Plutonium contamination at Maralinga, in Residual Radioactive Contamination at Maralinga and Emu, 1985, ed. K.H. Lokan, ARL/TRO70, April 1985. Crusem, R. and Caristan, Y., Moment tensor inversion, estimation, and seismic coupling variability at the French Centre d'Exp6rimentations du Pacifique. Bull Seism. Soc. Am., 82 (3) (1992) 1253-1274. Dalhman, O. and Israelson, H., Monitoring Underground Nuclear Explosions, Elsevier, Amsterdam, 1977. Delclos, C., Blanc, E., Broche, P., Glangeaud, F. and Lacoume, J.L., Processing and interpretation of microbarograph signals generated by the explosion of Mount St. Helens. J. Geophys. Res., 95 (D5) (1990) 5485-5494. Donn, W.L. and Balachandran, N.K., Mount St. Helens eruption of 18 MaY 1980--Air waves and explosive yield. Science, 213 (1981) 539-541. Eisenbud, M., Environmental Radioactivity from Natural, Industrial and Military Sources. Harcourt Brace Jovanovich, Orlando, FL, 1987. Ellis, E.W. and Wall, T., Use of particle track analysis to measure fissile particle size distributions in contaminated soils. Nucl. Instrum. Meth., 200 (1982) 411. Encyclopedie de la Polynesie 1, Les iles Oc6aniques. Christian Gleizal/Multipress, 1986. ESEPA, Interim recommendations on doses to persons exposed to transuranium elements in the general environment, U.S. Environmental Protection Agency, Report, April 1985. European Commission, DGXI, Verifications tender the terms of Article 35 of the Euroatom Treaty, French Polynesia. Technical Report, Doc. Nr. 5037/45. Fages, J., Petit Atlas de la Polyn6sie Franqaise. Soci6t6 des Oc6anistes, 1974. Fondation Cousteau, Mission Scientifique de la Calypso sur le site d'exp6rimentations nucl6aires de Mururoa. Published by the Cousteau Fondation, 25 avenue Wagram, 75017-Paris, November 1988. Francis, S.H., Global propagation of atmospheric gravity waves: a review. J. Atmos. Terr. Phys., 37 (1975) 1011-1054. Friesen, B., Holmes & Narver, Inc., Enewetak Radiological Support Project--Final Report, September 1982. United States Department of energy, Nevada Operations Office, Las Vegas, Nevada, NVO-213. General Charles Ailleret, L'Aventure Atomique Franqais. Grasset, Paris, 1968, p. 228. Georges, T.M., HF Doppler studies of travelling disturbances. J. Atmos. Terr. Phys., 30 (1968) 735-746. Glasstone, S. and Dolan, P.J., The Effects of Nuclear Weapons. U.S. Government Printing Office, Washington, DC, 1977. Goldschmidt, The Atomic Adventure, p. 122. Guille, G., Goutiere, G. and Sornein, J.F., Les atolls de Mururoa et Fangataufa (Polyn6sie franqaise), Tome 1, Editions Masson, 1993. Guille, G., Maury, R.C., Buigues, D., Bellon, H., Gachon, A. and Caroff, M., L'atoll de Mururoa (Polyn6sie franqaise): III. Conclusions g6n6rales, Bull. Soc. G6ol. France, 163 (5) (1992) 681-685. Guy, CH., Scott, J., Destrigneville, C. and Chiappini, R., Low temperature alteration of Mururoa basalts by interstitial seawater. Geochim. Cosmochim. Acta, 56 (1992) 4169-4189. Houard, S., Plantet, J.L., Massot, J.P. and Nataf, H.C., Amplitude of core waves near the PKP Caustic, from Nuclear Explosions in the South Pacific Recorded at the LDG network, in France. Bull. Seism. Soc. Am., 83 (1993) 1835-1854. IAEA, Assessment of the Kazakhstan nuclear test site. IAEA Yearbook 1995, p. D64. IAEA Technical Report Series No. 295. IAEA, Vienna, 1989. IAEA, Radiological Conditions at Bikini Atoll: Prospects for Resettlement, Radiological Assessment Reports Series, IAEA, Vienna, 1998. IAEA, Report by an International Advisory Committee: The Radiological Situation at the Atolls of Mururoa and Fangataufa, 6 volumes, IAEA, Vienna, 1998. Iranzo, E., Salvador, S. and Iranzo, C.E., Health Phys., 52 (4) (1987) 453-461. Jaeger, J.C. and Cook, N.G.W., Fundamentals of Rock Mechanisms, 3rd edn. Chapman and Hall, London, 1979. Johnston, P.N., Lokan. K.H., Richardson, C.K. and Williams, G.A., Plutonium contamination in the Maralinga Tjarfje Lands Report ARL/TRO85, ISSN 0157-1400, August 1989. Kehl, S.R., Mount, M.E. and Robinson, W.L., The Northern Marshall Islands Radiological Survey: A
Bomb Test Sites
557
Quality Control Program for Radiochemical and Gamma Spectroscopy Analysis, Rep. UCRL-ID-120429, Lawrence Livermore National Laboratory, Livermore, CA (1995). Kehl, S.R., Mount, M.E. and Robinson, W.L., The Northern Marshall Islands Radiological Survey: A Quality Control Program for Radiochemical and Gamma Spectroscopy Analyses, Report UCRL-ID-120429, September, 1995. Kennett, B.L.N., Seismic Wave Propagation in Stratified Media. Cambridge University Press, New York, 1983. Kinney, G.F. and Graham, K.J., Explosive Shocks in Air, 2nd edn. Springer Verlag, New York, 1985. Klemm, J. Goetz, J., Phillips, J. and Thomas, C., Analysis of Radiation Exposure to Service Personnel on Rongerik Atoll, Operation Castle, Shot Bravo, SAIC-86/1608, Science Applications International Corp., La Jolla, CA (1986). Kobis de Saint-Chamas, L., Boblet, J.P., Arnould, C. and Ducousso, R., Evolution de la teneur en cesium 137 depnis 1967 dans l'eau de coc et le coprale en Polyn6sie Franqaise. Radioprotection, 26 (1991) 75-88. Kumatori, T., et al., Follow-up studies over a 25-year period on the Japanese fisherman exposed to radioactive fallout in 1954. The Medical Basis for Radiation Accident Preparedness. Elsevier-North Holland, New York, 1980, pp. 33-54. La Rosa, J., Danesi, P.R., Fajgely, A., Makarewicz, M., Vajda, N., Valkovic, V. and Zeisler, R., Analytical approach to the measurement of radionuclides in environmental contamination of a former nuclear weapons testing area. IAEA-SM-339/144P, p. 305, Vienna, Austria, 1996. LaRosa, J., Danesi, P.R., Fajgelj, A., Makarewicz, M., Vajda, N., Valkovic, V., Zeisler, R., Stegnar, P., Analytical approach to the measurement of radionuclides in environmental contamination of a former nuclear weapon testing area, presented at International Symposium on Environmental Impact of Radioactive Releases, International Atomic Energy Agency, Vienna, 8-12 May 1995. Latter, A.L., Martinelli, E.A. and Teller, E., Seismic scaling law for underground explosions. Phys. Fluids, 2 (1959) 280. Martin, J.M., Thomas, A.J., Charrier, B., Cousteau, J.Y. and Sarano, F., Assessment of artificial radionuclides from French nuclear bomb testing at Mururoa (French Polynesia). Environ. Technol. Lett., 11 (3) (1990) 197-204. Martin, J.-M., Thomas, A.J., Charrier, B., Cousteau, J.-Y. and Sarano, F., Assessment of artificial radionuclides issued from French nuclear bomb testing at Mururoa (French Polynesia). Environ. Technol., 11 (3)(1990) 197-204. Massot, J.P. and Rocard, Y., Variation of amplitude of PKP from underground explosions in the Southcentral Pacific. Geophys. Res. Lett., 9 (1982) 1211-1214. Maury, R.C., Caroff, M., Achard, S., Guille, G., Joron, J.L., Gachon, A., Rocaboy, A. and Leterrier, J., L'atoll de Mururoa (Polyn6sie franqaise): II) La s6rie magmatique. Bull. Soc. G6ol. France, 163 (5) (1993) 659-680. McEwan, A.C., Baverstock, K.F., Paretzke, H.G., Sankaranarayanan and K., Trott, K.R., The Scientific Advisory Panel to the Nationwide Radiological Study, Findings of the Nationwide Radiological Study, Summary Report prepared for the Cabinet of the Government of the Republic of the Marshall Islands, Majuro (December 1994). Melsoh, H.J., Impact Cratering: A Geologic Process. Oxford University Press, New York, 1989. Miltenberger, R.P., Greenhouse, N.A. and Lessard, E.T., Whole body counting results from 1974 to 1979 for Bikini Island residents. Health Phys., 39 (1980) 395-507. Muller, G. and Murphy, J.R., Seismic characteristic of underground nuclear detonations, Part I: Seismic spectrum scaling. Bull. Seism. Soc. Am., 61 (6) (1971) 1675-1692. Patterson, M.S., Experimental Rock Deformation~The Brittle Field. Springer Verlag, New York, 1978. Pitman, M. and collaborators, The impact of nuclear testing at Mururoa and Fangataufa. A paper proposed for the South Pacific Environment Ministers Meeting, Brisbane, August 1995. Price, G.H., The Electromagnetic Pulse from Nuclear Detonations. Rev. Geophys. Space Phys., 12 (3) (1974) 389-400. Regnault, J.M., La Bombe Franqaise dans le Pacifique~L'implantation 1957-1964. Edition Scoop, 1993. Robinson, W.L., Bogen, K.T. and Conrado, C.L., An updated assessment for resettlement options at Bikini
558
Chapter I0
Atoll--A U.S. nuclear test site. Health Phys., 73 (1) (1997) 100-114. Robinson, W.L., Conrado, C.L., Eagle, R.J. and Stuart, M.L., The Northern Marshall Islands Radiological Survey: Sampling and Analysis, Summary, Rep. UCRL-52853, Part 1, Lawrence Livermore National Laboratory, Livermore, CA ( 1981). Robinson, W.L. et al., An Updated Assessment of Bikini and Eneu Islands at Bikini Atoll, Rep. UCRL-53225, Lawrence Livermore National Laboratory, Livermore, CA (1980). Robinson, W.L. et al., The Northern Marshall Islands Radiological Survey: Terrestrial Food Chain and Total Doses, Rep. UCRL-52853, Part 4, Lawrence Livermore National Laboratory, Livermore, CA (1982). Robinson, W.L. et al., The Northern Marshall Islands Radiological Survey: Data and Dose Assessments. Health Phys., 73 (1) (1997) 37-48. Robinson, W.L., Noshkin, V.E., Phillips, W.A. and Eagle, R.J., The Northern Marshall Islands Radiological Survey: Radionuclide Concentrations in Fish and Clams and Estimated Doses via the Marine Pathway, Rep. UCRL-52853, Part 3, Lawrence Livermore National Laboratory, Livermore, CA (1981). Robinson, W.L., Phillips, W.A. and Colsher, C.S., Dose Assessment at Bikini Atoll, Rep. UCRL-51879, Part 5, Lawrence Livermore National Laboratory, Livermore, CA (1977). Rodean, H.C., Nuclear Explosion Seismology. U.S.A.E.C. Critical Review Series TID-25572, 1971. Rosen, M., International Atomic Energy Agency Bulletin, 35 (4) (1993) 34-38. Schell, W.R., Lowman, F.G. and Marshall, R.P., "Geochemistry of Transuranic Elements at Bikini Atoll", Transuranic Elements in the Environment (Hanson, W.C., Ed.), Rep. DOEF['IC-22800, Technical Information Centre, US Department of Energy, Oak Ridge, TN (1980). Sedov, L.I., Similarity and Dimensional Methods in Mechanics. Academic Press, New York, 1959. Shapiro, C.S. (ed.), Atmospheric Nuclear Tests: Environmental and Human Consequences. NATO ASI Series 2, Environment 35. Proc. Workshop, Vienna, Austria, Jan. 1994. Springer-Verlag, New York, 1998, 280 pp. Shapiro, C.S., Kiselev, V.I. and Zaitsev, E.V. (eds.), Nuclear Tests: Long-term Consequences in the Semipalatinsk/Altai Region. NATO ASI Series 2, Environment 36. Proc. Workshop, Barnaul, Russia, Sep. 1994. Springer-Verlag, New York, 1998, 193 pp. Sharp, R. and Chapman, W.H., Exposure of Marshall Islanders and american military personnel to fallout, Rep. WT-938, Atomic Energy Commission, Washington, DC (1957). Shell, P. and Hutter, A.R., Environmental radiation measurements at the former Soviet Union's Semipalatinsk nuclear test site and surrounding villages. Environmental Measurements Laboratory, New York, Report 1995. Shoihet, Ya.N. et al., Radiation impact of nuclear tests at the Semipalatinsk test site on the population of the Altai region in the Russian Federation. IAEA-SM-339/82, p. 39, Vienna, Austria, 1996. Simon, S.L. and Graham, J.C., Findings of the first comprehensive radiological monitoring program of the Republic of the Marshall Islands. Health Phys., 73 (1) (1997) 66-85. Simon, S.L. and Graham, J.C., Marshall Islands radiological survey of Bikini Atoll, Marshall Islands Nationwide Radiological Study, Republic of the Marshall Islands, Majuro (February, 1995). Simon, S.L., Jenner, T., Graham, J.C. and Borchert, A., A comparison of macro- and microscopic measurements of plutonium in contaminated soil from the Republic of the Marshall Islands. J. Radioanal. Nucl. Chem., 194 (1995) 197. Simon, S.L. and Robinson, W.L., A compilation of atomic weapons test detonation data for US Pacific Ocean tests Health Phys., 73 (1) (1997) 258-264. Sornein, J.F. and Guy, CH., Hydro-g~ochimie et circulation naturelle dans un atoll, Chocs/CEA, 7, pp. 37-47, 1993. Stegnar, P. and Wrixon, T., Semipalatinsk revisited. IAEA Bulletin, 40 (1998) 12. Stoker, A.C. and Conrado, C.L., The Marshall Islands Data Management Program, Report UCRL-ID- 120430, Sept. 1995. Taylor, S.T., Patton, H.J. and Richards, P.G., Explosion Source Phenomenology. Geophysical Monographs Series. American Geophysical Union, Washington, 1991. Tazieff, H., Rapport d'Haroun Tazieff sur l'ensemble de la mission scientifique en Polyn~sie franqaise, 1982.
Bomb Test Sites
559
Tipton, W.J. and Meibaum, R.A., An aerial radiological and photographic survey of eleven atolls and two islands within the Northern Marshall Islands, Rep. EGG- 1183-1758, Eg&G, Las Vegas, NV (1981). United States Department of Energy, United States Nuclear Tests, July 1945 through September 1992, Rep. DOE/NV-209, Rev. 14, USDOE Nevada Operations Office, Las Vegas, NV (1994). Williams, G.A. and Burns, P.A., Report on measurements of radioactivity levels in soil samples from the road to Oak Valley, April 1987; and Second Report on measurements of radioactivity levels in soil samples from the Oak Valley area, June 1987; unpublished ARL reports, ARL File No. 240-4-1 part 3. Yeh, K.C. and Liu, C.H., Acoustic-gravity waves in the upper atmosphere. Rev. Geophys. Space Phys., 12 (2) (1974) 193-216. Zel'dovich, Ya.B. and Raiser, Yu.P., Physics of Shock Waves and High-temperature Hydrodynamic Phenomena. Academic Press, New York, 1967.
561
CHAPTER 11
International Safeguards
The Treaty on the Non-Proliferation of Nuclear Weapons (NPT) is the single most important component of the non-proliferation regime. Under it, the 182 non-nuclearweapon State Parties have committed themselves not to manufacture or otherwise acquire any nuclear explosive device and to accept IAEA safeguards on all source or special fissionable material to verify that commitment. Under it also the five nuclear weapon State Parties have committed themselves to embark on effective measures relating to nuclear disarmament. The Treaty embodies and promotes the idea of the non-legitimacy of the spread of nuclear weapons. Of course, the first line of defence against proliferation is the governmental decision that it is not in the political or security interest of the State to acquire nuclear weapons. However, there can be no doubt about the relevance of the NPT to that policy conclusion. Such decisions are not made in a vacuum: they are made against the background of a number of considerations including not only domestic politics, bilateral, and regional relationships, but also the broader international environment of which the NPT and the non-proliferation regime are a part. The NPT does not dictate a decision one way or the other, but it is a part of the framework of the decision process of most, if not all, governments. It is difficult for a country to move to nuclear weapons status when other States, many of whom are important to it for political, economic, cultural, and other reasons, have rejected such action for themselves, seek a world ultimately devoid of such weapons, and have joined a treaty repudiating nuclear weapons. The non-parties who may have acquired the capability to make nuclear weapons, but have not openly acknowledged either the fact or that they may actually have produced them, have clearly been influenced by the NPT and by the substantial support the Treaty enjoys. The normative constraint of the Treaty should not be underestimated. Parenthetically, the decisions of the French and Chinese Governments to adhere to the Treaty brought all five nuclear-weapons States under formal NPT obligation, even if it does not lead to all remaining States joining the NPT, can only serve to strengthen the norm and to raise further the barriers to proliferation (Scheinman, 1992).
562
Chapter 11
Verification is a central ingredient of modern-day arms control agreements. It may serve either as a deterrent or nuclear arena where the same technologies, materials, and facilities can serve peaceful or military ends, verification of use is particularly important and the NPT provides for verification by the IAEA of compliance with the undertaking not to divert nuclear materials. Prior to the NPT, verification safeguards were limited to situations in which a supplier State conditioned transfer of nuclear plant or materials on the recipient accepting safeguards on whatever was transferred, but without requiting that any other nuclear activities in the State also be placed under safeguards. Under NPT, non-nuclear parties agree to submit all of their peaceful nuclear activities to safeguards whether imported or developed indigenously. The idea is to have a comprehensive picture of the location and status of all nuclear materials in the State.
11.1 T H E T R E A T Y ON THE N O N - P R O L I F E R A T I O N OF N U C L E A R WEAPONS In accordance with Article X of the Treaty, which entered into force in 1970, the States party to it needed to decide, in 1995, on its extension. Article X provides that "twenty-five years after the entry into force of the Treaty, a conference shall be convened to decide whether the Treaty shall continue in force indefinitely, or shall be extended for an additional fixed period or periods. This decision shall be taken by a majority of the parties to the Treaty." It is interest to quote here the decisions made at a UN conference held in New York 17 April-12 May 1995: 9 Universal adherence to the Treaty on the Non-Proliferation of Nuclear Weapons is an urgent priority. All States not yet party to the Treaty are called upon to accede to the Treaty at the earliest date, particularly those States that operate unsafeguarded nuclear facilities. Every effort should be made by all State Parties to achieve this objective. 9 The proliferation of nuclear weapons would seriously increase the danger of nuclear war. Every effort should be made to implement the Treaty in all its aspects to prevent the proliferation of nuclear weapons and other nuclear explosive devices, without hampering the peaceful uses of nuclear energy by States parties to the Treaty. 9 Nuclear disarmament is substantially facilitated by the easing of international tension and the strengthening of trust between States which have prevailed following the end of the cold war. The undertakings with regard to nuclear disarmament as set out in the Treaty on the Non-Proliferation of Nuclear Weapons should thus be fulfilled with determination. In this regard, the nuclear-weapon States reaffirm their commitment, as stated in article VI, to pursue in good faith negotiations on effective measures relating to nuclear disarmament.
International Safeguards
563
9 The achievement of the following measures is important in the full realisation and effective implementation of article VI, including the programme of action as reflected below: the completion by the conference on Disarmament of the negotiations on a universal and internationally and effectively verifiable comprehensive Nuclear-Test-Ban Treaty no later than 1996. Pending the entry into force of a Comprehensive Test-Ban Treaty, the nuclear-weapon States should exercise utmost restraint; - the immediate commencement and early conclusion of negotiations on a non-discriminatory and universally applicable convention banning the production of fissile material for nuclear weapons or other nuclear explosive devices, in accordance with the statement of the Special Coordinator of the Conference on disarmament and the mandate contained therein; the determined pursuit by the nuclear-weapon States of systematic and progressive efforts to reduce nuclear weapons globally, with the ultimate goals of eliminating those weapons, and by all States of general and complete disarmament under strict and effective international control. 9 The International Atomic Energy Agency (IAEA) is the competent authority responsible to verify and assure, in accordance with the statute of the IAEA and the Agency's safeguards system, compliance with its safeguards agreements with State Parties undertaken in fulfilment of their obligations under article III (1) of the Treaty, with a view to preventing diversion of nuclear energy from peaceful uses to nuclear weapons or other nuclear explosive devices. Nothing should be done to undermine the authority of the IAEA in this regard. State Parties that have concerns regarding non-compliance with the safeguards agreements of the Treaty by the State Parties should direct such concerns, along with supporting evidence and information, to the IAEA to consider, investigate, draw conclusions and decide on necessary actions in accordance with its mandate. 9 All States required by article III of the Treaty to sign and bring into force comprehensive safeguards agreements and which have not yet done so should do so without delay. 9 IAEA safeguards should be regularly assessed and evaluated. Decisions adopted by its Board of Governors aimed at further strengthening the effectiveness of IAEA safeguards should be supported and implemented and the IAEA's capability to detect undeclared nuclear activities should be increased. Also States not party to the Treaty on the Non-Proliferation of Nuclear Weapons should be urged to enter into comprehensive safeguards agreements with the IAEA. 9 New supply arrangements for the transfer of source or special fissible material or equipment or material especially designed or prepared for the processing, use or production of special fissionable material to non-nuclear weapon States should require, as a necessary precondition, acceptance of IAEA full-scope safeguards and internationally legally binding commitments not to acquire nuclear weapons or other nuclear explosive devices. -
-
Chapter 11
564
9 Nuclear fissile material transferred from military use to peaceful nuclear activities should, as soon as practicable, be placed under IAEA safeguards in the framework of the voluntary safeguards agreements in place with the nuclearweapon States. Safeguards should be universally applied once the complete elimination of nuclear weapons has been achieved. 9 Particular importance should be attached to ensuring the exercise of the inalienable fight of all the parties to the Treaty to develop research, production and use of nuclear energy for peaceful purposes without discrimination and in conformity with articles I, II as well as III of the Treaty. 9 Undertakings to facilitate precipitation in the fullest possible exchange of equipment, materials and scientific and technological information for the peaceful uses of nuclear energy should be fully implemented. Table 11.1 IAEA safeguards inspections in the year 1993 Facility Type
Facilities
Inspection type
Inspections
PDI
PDI per facility
LWR
144
PIV Interim
127 693
261 1143
2 8
Onload Reactor
16
PIV Interim
14 80
521 1332
33 83
Other Reactor
7
PIV Interim
6 66
15 163
2 23
Research Reactor & Criteria Fac.
122
PIV Interim
117 247
241 432
2 4
DNLEU, Th Conversion & Fuel Fabr.
37
PIV Interim
36 89
350 256
9 7
HEU, Pu Conversion & Fuel Fabr.
12
PIV Interim
12 96
200 1048
17 87
Reprocessing
7
PIV Interim
6 49
66 579
9 83
Enrichment
8
PIV Interim
8 120
129 442
16 55
Storage
39
PIV Interim
37 151
150 464
4 12
Other Facilities
41
PIV Interim
38 71
87 163
2 4
Locations Outside Facilities
32
PIV Interim
31 2
65 2
2 0
Non-Nuclear Installations
2
PIV Interim
1 6
5 60
3 30
Total Facility Types
467
2103
8174
18
PDI = person-days of inspection.
International Safeguards
565
9 In all activities designed to promote the peaceful uses of nuclear energy, preferential treatment should be given to the non-nuclear-weapon States party to the Treaty, taking the needs of developing countries particularly into account. The summary of IAEA safeguards experiences are regularly presented (see Agu and Iwamoto, 1983; Agu et al., 1987; Schrefer et al., 1994). For example, Table 11.1 gives an overview of activities performed during 1993.
11.2 IAEA ANALYTICAL CAPABILITIES FOR SAFEGUARDS One of the International Atomic energy Agency's (headquarters in Vienna, Austria) most important responsibilities is to verify the fulfilment of safeguards obligations assumed by States, under agreements with the IAEA concerning the peaceful use of nuclear materials or equipment. The IAEA safeguards system comprises technical measures for performing these verifications within the framework of international non-proliferation policy entrusted to the IAEA in its Statute and by the Treaty on the Non-Proliferation of Nuclear Weapons (the NPT). This function may be accomplished in a number of ways: 1. by accountability verification, which consists of auditing the accounting records of States and/or facilities; 2. by containment, which involves physically restricting or controlling the movement of, or access to, nuclear materials or equipment. Integrity of containment is maintained by the use of suitable tamper-indicating seals; 3. by surveillance, which is a process of observation, both by inspectors and by cameras and electronic aids, designed to detect undeclared movements of, or tampering with, nuclear materials; 4. by item identification verification, whereby the Agency's inspectors verify during their inspection that declared nuclear materials or equipment are actually present in the facility, whether it be in the form of fuel elements, fuel assemblies, storage drums or tanks etc.; 5. by material accountancy verification, a major instrument of the Agency's safeguards system by which the amount of nuclear materials present at any given facility is verified by independent measurements. This requires independent measurement of the weights of materials in solid form, or of the volumes of materials in solution. A vital part of the accountancy verification process is the independent analysis of the composition of nuclear materials present at the facility. Some analyses are carried out during Agency inspections without physically affecting the items being examinedma process known as Non-Destructive Assay (NDA)--but it is also necessary to measure some samples by destructive techniques which chemically modify or destroy the original material. This necessitates sending a sample of the material to a specialised safeguards analytical laboratory (SAL) (after Deron et al., 1994).
Chapter 11
566
Light-water reactors, LWRs, use low-enriched uranium, LEU, categorised as "indirect-use" material from the standpoint of its potential use in the manufacture of nuclear weapons. After these nuclear materials have been fuelled in the reactor core, the spent fuels are categorised as "direct-use" material. Plutonium contained in spent fuel, as well as fresh mixed uranium-plutonium oxide, MOX fuels, represent a strategic material from a safeguards standpoint. This is one of the determining factors that affects the safeguards approach and the inspection goal for a facility (see Harms and Rodriguez, 1996; Schriefer, 1996). The application of IAEA safeguards on LEU is based on a number of criteria, specifying inspection goals whereby the significant quantity is an amount of uranium containing 75 kg of uranium-235, and the timeliness goal is one year. This means that the Agency, when implementing its safeguards system, shall be able to detect a diversion of at least 75 kg of uranium-235 contained in LEU during a time period of one year (Nilsson, 1996). In general, the safeguards approach is based on an analysis of all technically possible diversion paths at a facility and on the requirements of the particular safeguards agreement. The approach is also designed to counter the possible undeclared production of direct-use material. It refers to the system of nuclear materials accountancy, containment, surveillance, and other measures chosen for implementation of safeguards. The following are also taken into consideration: (i) measurement methods and techniques available to the Agency; (ii) the design features of the facility; (iii) the form and accessibility of the nuclear material; (iv) the possible existence of unsafeguarded nuclear activities; and (v) inspection experience. The inspection goal for a facility consists of a quantity component and a timeliness component (see Table 11.2). The quantity component relates to the scope of the inspection activities necessary in order to provide assurance that there was no diversion of a significant quantity (SQ) of nuclear material over a material balance period (MBP). The timeliness component on the other hand relates to the periodic inspection activities Table 11.2 IAEA Safeguards inspection goals Category
Type
Significant quantities
Timeliness goals
Direct-use material
Plutonium* High-enriched uranium
8 kg plutonium 25 kg uranium-235
Plutonium in spent fuel Uranium-233
8 kg plutonium 8 kg uranium-233
1 month 1 month (fresh) 3 months (spent) 3 months 1 month
Indirect-use material
Low-enriched uranium** 75 kg uranium-235 Thorium 20 t thorium
*For plutonium containing less than 80% plutonium-238. **Less than 20% uranium-235; includes natural and depleted uranium.
12 months 12 months
International Safeguards
567
necessary to provide assurance that no abrupt diversion has taken place. The inspection goal for each facility is regarded as attained if all the criteria relevant to the material types and categories present at the facility have been satisfied. In its implementation of safeguards, the IAEA strives for full attainment of both components of the inspection goal. Approximately 180 research reactors and critical assemblies currently are under IAEA safeguards. The vast majority of the research reactors operate at relatively low power levels (10 megawatts-thermal or lower) and the critical assemblies at virtually zero power. Research reactors are widely used for scientific investigations and various applications. Neutrons produced by research reactors provide a powerful tool for studying matter on nuclear, atomic, and molecular levels. Neutrons, often are used as probes by nuclear and solid state physicists, chemists, and biologists. Neutron experiments can also be performed outside the biological shield by means of installed beam tubes. Additionally, specimens can be positioned in or near research reactor cores for neutron irradiation, e.g. to produce radioactive isotopes for medical or research use. For research reactors, the following diversion scenarios are considered by IAEA: 1. Diversion of fresh or slightly irradiated fuel for clandestine chemical extraction of fissile material: this scenario is given particular safeguards attention at facilities where the fresh fuel contains HEU of plutonium, for which no further transmutation or enrichment would be needed for use in a nuclear explosive device. About 20 research reactors under IAEA safeguards are currently using such direct-use fissile material in amounts equal to more than one significant quantity (SQ) (Zuccaro-Labellarte and Fogerholm, 1996). 2. Diversion of spent or extensively irradiated fuel for clandestine chemical extraction of fissile material in a reprocessing facility: this scenario is technically more demanding and time-consuming than the one mentioned above because of the high level of radioactivity from the fuel which is involved. However, it is of particular concern at about 15 research reactors under IAEA safeguards due to large accumulated quantities of spent fuel, and it is of importance at more than 100 others. 3. Clandestine production: the possibility exists for clandestine production of plutonium or uranium-233 through irradiation of undeclared fertile material. As techniques for using neutrons have developed, there has been an accompanying need for higher levels of neutron flux in order to carry out more complex and time-consuming experiments in a shorter time. Large research reactors have been constructed to provide these flux levels. At such reactors, production of substantial quantities of plutonium or uranium-233 through irradiation of undeclared fertile material would be technically feasible. This could be achieved, for example, by placing target materials in irradiation positions in or near the core, or by replacing reflector elements by fertile material targets. However, studies have shown that it is not possible to produce one SQ of plutonium in one
568
Chapter 11
year at a research reactor that operates below about 25 megawatts-thermal. The actual production capability depends on the individual reactor design and operating parameters. The Agency' s current safeguards system requires that all research reactors operating at power levels above 25 megawatts-thermal are evaluated with respect to their capability to produce at least one SQ of plutonium (or uranium-233) per year. At present, there are about 30 thermal research reactors with power levels of 10 megawatts-thermal or higher, which are subject to IAEA safeguards. About 10 of these operate at power levels exceeding 25 megawatts-thermal and are subject to additional safeguards measures with respect to the clandestine production scenarios (ZuccaroLabellarte and Fagerholm, 1996). To ensure the independence of the destructive analyses required by the safeguards system, the IAEA set up a Safeguards Analytical Laboratory as a part of its other Laboratories, adjacent to the Austrian Research Centre near the village of Seibersdorf some 30 kilometres south of Vienna. Here, a team of 20 scientists and technicians determine the concentration and isotopic composition of safeguarded nuclear materials by the analysis of samples taken at all stages of the nuclear fuel cycle, at some 56 nuclear facilities all over the world. The construction of SAL at Seibersdorf followed an agreement between the IAEA and the Austrian Research Centre in July 1973. Completed in 1975, the Laboratory went into operation in January 1976 with a limited licence. A full licence was granted in 1979, allowing SAL to analyse every kind of fissile material. Since then, the Laboratory has been working at almost full capacity, handling about 1200 samples a year. This total comprises about 700 non-irradiated uranium samples, 300 non-irradiated plutonium samples and 200 spent fuel samples, giving rise altogether to as many as 8000 data measurements. SAL's budget in 1990 was about US $2.4 million, less than 4.5% of the total safeguards budget. The sampling of nuclear materials and the packaging of samples are very critical steps in the measurement process. Differing conditions at nuclear plants dictate the need for a variety of sample-taking and handling procedures before samples are sent for analysis. SAL has developed a set of procedures that take account of local conditions, and has been instrumental in improving the overall quality of the verification of uranium and plutonium samples. Tracer techniques, for example, are used to obtain very small but representative and measurable samples of highly radioactive spent fuel solutions. One millilitre of the solution is then spiked with a known amount of uranium and plutonium tracer isotopes. A few microlitres of the spiked solution are dried and shipped to SAL. One to fifty nanograms of uranium or plutonium extracted from this tiny sample are sufficient for a complete analysis representing the composition of half a tonne of irradiated fuel with an accuracy of 0.3 to 0.5 %. Well characterised reference materials similar in composition, shape and packaging to actual nuclear materials are necessary to calibrate non-destructive techniques or to verify their accuracy. SAL performs analyses requested for the verification of such
569
International Safeguards
Harwell ]
Geel
Mol
Petten
Berlin
Dresden
\
CEA ~ Fontenay I - ' - ~ ~
-.,. ~ ,
~
~
, _
%>
_-._ I St. Petersburg
~~.~
AEC Chalk River AN~
~
/
~ N N ~ r~ N ~
~
NMC~
~
ORNL
7
l ........
|
OeFZS ] Seiberdorf I
EUREX Saluggia
ITREC Bari
NRI Prague
CRPI Budapest
BHABHA Bombay
Fig. 11.1. The IAEA network of laboratories for the analysis of fissile materials. reference materials. Some 50 materials have now been certified, meeting a wide variety of plant and material specific criteria. At nuclear fuel chemical processing plants of large throughputs, it is desirable to perform verification analyses directly at or near the plant with an accuracy of 1% or better. This is done using specialised analytical instrumentation, such as robotized wet chemical analytical stations, compact mass spectrometers, X-ray absorptiometers or fluorescence analysers, installed at the facility or at a local verification laboratory. SAL provides in-house support to these approaches in the form of qualification and calibration of methods and instruments, evaluation of the performance of on-site analyses, resolution of technical problems, and training of inspectors. Although all safeguards samples are addressed to SAL, they are not all analysed there. Some are re-dispatched to a network of analytical laboratories (NWAL) designated by national authorities to provide contractual services to the Agency (see Fig. 11.1). The laboratories of the Network as well as SAL, provide services in support to NDT measurements and on-site analytical procedures. To date, the Agency has invited fourteen Member States and the European Communities to co-operate in this network. Advisory Group meetings are held once every two years, attended by representatives of SAL and NWAL together with other Agency and national experts to review the quality of analytical measurements for safeguards. SAL is equipped to implement the four major types of analytical measurements required in material accountancy. 1. Wet chemical analyses are performed to determine the concentration of thorium, uranium and plutonium in non-irradiated nuclear materials. This involves most frequently potentiometric titration backed up by ignition gravimetry with spectrographic analysis.
570
Chapter 11
2. Mass spectroscopy is the basic technique used for isotopic analysis. It is also used to determine the concentrations for fissile elements in samples that have been diluted and thus contain too little uranium and plutonium for accurate assay by standard wet chemical techniques. A tracer technique, known as isotope dilution analysis, is used in such cases, particularly with irradiated spent fuel or resin bead samples. 3. Radiometric techniques, essentially alpha and gamma spectrometry, allow the determination of a number of fissile and fission product isotopes. 4. Emission spectrography is used to determine the content of impurities in samples of uranium and plutonium oxide powders and pellets. Here is how SAL works: Samples are received in a reception and storage room, then routed to the appropriate wet chemical analysis laboratory. There, they are analysed for uranium, thorium or plutonium content, and purified aliquots (portions of the sample) are prepared for the isotopic analysis of three elements. Isotopic analyses are performed routinely by mass spectrometry, and radiometric techniques are used for back-up. Emission spectrography serves to detect the presence of impurities which could interfere with the measurements and thus distort the results of the chemical and isotopic analysis of uranium, thorium and plutonium. Complex calculations and quality checks are performed on minicomputers, which are connected in a network to a central laboratory mini-computer. A central laboratory data system stores and provides analytical reports and enables the quality of the analyses and the status of the flow of samples through the laboratory at any time to be monitored.
11.2.1 Analytical requirements for safeguards Two features characterise the analytical requirement for the support of safeguards verifications: 1. analytical accuracies of 0.1% or better are desirable and achievable for the accountability measurements at large bulk handling facilities coming under IAEA Safeguards; 2. significant improvements in analytical technology have occurred in particular in mass spectrometry and X-ray absorptiometry, so that accuracies of 0.1% can be now a standard of practice for all feed and products at industrial chemical processing plants, when necessary. Sufficiently accurately certified materials are needed to ensure that the uncertainties of the certificates of the reference materials used for calibration remain small compared to other sources of uncertainties. As the accuracy of nuclear material accountability and safeguards measurements improves, it is appropriate to examine whether the reference materials used for the calibration of such measurements and the physical constants used in their evaluation provide with sufficient accuracy the expected link between field measurements and the SI system.
International Safeguards
571
An often-asked question is: which measures may be needed and taken to improve the traceability of nuclear material measurements by the use of certified reference materials. The following means may be considered: a. The certificate of CRMs should give the certified values in SI units. b. The certificates should identify the links which make the certified values traceable to the SI system. c. In particular, the average atomic weight or the isotopic composition of the analyte should be provided as a certified value with adequate accuracy. d. The organisation producing a CRM should render public the technical report on the basis of which the RM was certified. e. A copy of this report should be delivered to the user along with the certificate. f. Independent assessment of this report by a suitable international technical panel would enhance the evidence of the quality of the CRM and its international recognition. Traditional safeguards measurements involve primarily the elemental and isotope analysis of nuclear materials encountered in declared plants and processes. New safeguard procedures, designed to verify that no illegitimate activity takes place clandestinely, lead to the more stringent performance of other analytical measurements. The following are examples of such additional measurements: a. high accuracy measurements of minor isotope abundances or impurity analysis of nuclear materials in order to identify or confirm the origin of the test material; b. determinations of trace and ultratrace levels of selected signatures in samples taken in the process or in the environment. Reference materials and measurement evaluation programmes are needed to verify that also such measurements are done accurately. The two groups of reference materials used by IAEA Laboratories are shown in Tables 11.3 and 11.4.
11.2.2 Equipment needs for special and non-routine inspections The equipment used should provide the IAEA with an immediate capability to carry out also non-routine inspection activities. It is assumed that in the course of such inspections, it may be necessary to locate and identify nuclear-related materials, equipment, and activities. It would also be necessary to provide a clearly documented audit trail of the inspection times, locations, activities, and results. The equipment selected to provide this capability should meet the highest current standards for high measurement sensitivity, flexibility in use, portability, and quality. The equipment should provide, but not be limited to, the following functions: 1. General radiation monitoring 2. Non-Destructive Analysis (NDA) a. Gamma-ray detection b. Neutron detection c. Detector signal acquisition and analysis
Chapter 11
572
Table 11.3 List of Reference Materials commercially available for IAEA Safeguards Destructive Analyses Use
Material
Source
Redox Titr.
Potassium dichromate
NIST 136e
U Assay
Nat-U Metal
NBL 112a EC-101 CETAMA MU2
Analyte
Accuracy
(%) 0.005
U U U
0.017 0.005 0.005
Depl-U Metal
NBL 115
U
0.005
Depl-UO2 Pellets
EC- 110 UK-UI/80350
U U
0.015 0.007
Nat-UO2 Pellets
CETAMA OU 1
U
0.102
LE-UO2-Pellets
NBL- 125
U
Natural U308
(prev. NBS 950) NBL 129
U
0.02
Nat-UO3
NBL 18
Nat-UF4
NBL 17b
LE-UF6
NBL 113 NBL 118
U 235U Th
0.02 0.0014 0.012
HEU+Th Carbide TRISO
NBL 119
U 235U Th
0.01 0.0014 0.009
Th Assay
Th-metal
WRM RI 022-92
Th
0.2
Pu-Assay
239-Pu metal
NBL 126 CETAMA MP2
Pu Pu
0.018 0.04
239-PuGa Alloy
UK-Pu 1/80990
Pu
0.017
NBL 122
Pu
0.039
CETAMAVENDEMIAIRE
U Pu
0.053 0.23
LWR (U,Pu)O2 pellets
CETAMA-CETAMOX
U Pu
0.03 0.05
LWR (U,Pu)O2 powder
WRM RI 025-92
U Pu
0.36 0.79
U/Pu/N d Assay
U/Pu/Nd-solutions
UK-UPuNd 1/90085 (LWR) UK-UPuNd2/90088 (FBR)
Np Assay
237-Np Solution (10 mg Np/g) (53 ~tg Np/g)
RI WRM 007-1-90 RI WRM007-2-90 AEA LMRI
Np Np Np
0.22 0.22 0.9
U/Th Assay HEU+Th Carbide BISO
239-PUO2 U/Pu Assay FBR (U,Pu)O2 pellets
1
International Safeguards
573
Accuracy (%)
Use
Material
Source
Analyte
Am Assay
241-Am Solution
NBL-132 LMRI
Am
Am
0.8
233/235/238-U solutions (various isotope ratios)
CBNM-072
235/238-U 233/238-U
0.03 0.03
233/235/238-U solutions (1:1:1)
NBL-117 EC-199 UK-U2/96059 UK-U2/96058
3/5, 3/8, 3/5, 3/5,
0.06, 0.03, 0.04, 0.04,
U Isotopic
Pu Isotopic
8/5-U 5/8-U 8/5-U 8/5-U
.09 .02 .05 .05
233/236-U nitrate
CETAMA MIRF 02
3/6-U
0.1
0.02-97% 235-U oxides
NBL (prev. NBS U-xxxx) BNFL LMRI ORGDP Technabexport
5/8-U 5/8-U 5/8.U
0.05 0.05 0.05
0.02-97% 235-UF6
CBNM-21/-27 BNFL LMRI ORGDP
5/8-U 5/8-U 5/8-U
0.05 0.05 0.O5
239/240-Pu solutions (var. ratios)
UK-Pu4/92136 UK-Pu4/92028
0/9-Pu 0/9-Pu
0.11 0.08
239/240/242-Pu solutions (1:1:1)
UK-Pu3/92134 UK-Pu3/92133
0/9, 2/9-Pu 0/9, 2/9-Pu
0.06, 0.1 0.06, 0.1
239/240/242/244-Pu solution (3:3:3:1)
UK-Pu5/92138
0/9, 2/9-Pu 4/9-Pu
0.11,0.19 0.24
239/242-Pu nitrate (various 239/242-Pu ratios)
IRMM-290
9/2-Pu
0.01
239/242-Pu nitrate (1:1)
NBL-128 CETAMA MIRF 01
9/2-Pu 9/2-Pu
0.026 0.05
239/244-Pu
CBNM-047a
9/4-Pu
0.2
Pu-sulfates (various isotope ratios)
NBL- 136 (prev. NBS-946) NBL- 137 (prev. NBS-947) NBL-138 (prev. NBS-948)
0/9-Pu 0/9-Pu 0/9-Pu
0.12 0.12 0.12
Pu (var. isotope ratios)
N B L - 1 4 0 - NBL- 143
Am Isotopic 241/243-Am nitrate solution
NBL-134
Nd Isotopic
AEA RM W 10930
Nd seven isotope mix. (2 ~tg Nd/g) Nd isotope mix.
0.04-0.11
IRMM
continued
Chapter 11
574
Table 11.3 (continuation) Use U Tracer
Analyte
Accuracy
233-U solution (0.5 mg 233-U/g) NBL-111A (0.5 mg 233-U/g) (NBL-111, prev. NBS 995) (1 mg 233-U/g) CBNM-040a (1 mg 233-U/g) (CBNM-040-1)
233-U
0.025 0.05 0.28 0.15
235-U metal (93 %) 235-U solution (7 mg 335-U/g)
U, 235-U
0.007, 0.005 0.18, 0.0013
Material
Source
NBL-116
(%)
Th Tracer
230-Th solution (40 ~tg 230-Th/g) CBNM-060
230-Th
0.74
Pu Tracer
242-Pu solution (9 ~tg 242-Pu/g) (9 ~tg 242-Pu/g) (91 ~tg 242-Pu/g) (66 ~tg 242-Pu/g) 242-Pu dried nitrate
CBNM-044 (CBNM-041/1) CBNM-049 (CBNM-041-2) NBL-130
242-Pu
0.15 0.3 0.2 0.3
240-Pu solution (94 % 240-Pu)
KRI
240-Pu
0.02
244-Pu 24-Pu 244-Pu
0.2 0.3 0.15
U, 5/8-U Pu, 0/9-Pu
0.1, 0.1 0.1, 0.1
3-U, 2-Pu 3-U, 2-Pu
0.15, 0.25
243-Am
0.7
244-Pu solution (0.9 lxg 244-Pu/g) CBNM-042a (0.9 ~tg 244-Pu/g) (CBNM-042) 244-Pu nitrate NBL- 131 (prev. NBS-996) 240/242/244-Pu dried nitrate
NBL-144
U/Pu Tracer 235-U/23p-Pu dried nitrate (LSD) IRMM-1027x
Am Tracer
Impurities
233-U/242-Pu solution (1 mg 233-U + 6 ~tg 242-Pu/g)
CBNM-046a (CBNM-046-2)
243-Am
U308
NBL-133 Harwell 95/243/37 LMRI NBL-123 NBL-124 CETAMA Chanterelle CETAMA Bolet CETAMA Morille
U-metal
CETAMA Floralies
Nat-UO3
NBL 18
Ammonium uranate
CETAMA (var.)
Sodium uranate
CETAMA Grenat
Magnesium uranate
CETAMA (var.)
Nat-UF4
NBL 17b
PuO2
CETAMA (Grands Crus)
FBR (U,Pu)O2 pellets
CETAMA-VENDEMIAIRE
ThO2
NBL-66
International Safeguards
575
Table 11.4 Reference Materials needed for IAEA Safeguards Destructive Analysis and unavailable commercially Use
Material
Source
Analyte
Accuracy
Pu-Assay
PuO2 (239-Pu <70 %)
Harwell
Pu
0.05
Nd Assay
Nat. Nd oxide Nd solution
IRMM Harwell
Lu Assay
Nat. Lu203
ORNL
Cm Assay
244-Cm metal
TUI KRI
U Isotopic
233/235/236/238-U solutions (1"1"1"1) (4 mg U/g) 233/235/236/238-U solutions various ratios (1 mg U/g)
KRI OCO-95.060-89
239/240/242/244-Pu solutions (various isotope ratios)
IRMM
Pu Isotopic
Pu Alpha Spec 238/239-Pu 238-Pu) 238/239-Pu 238-Pu) 238/239-Pu 238-Pu) 238/239-Pu 238-Pu)
sol. (1 mg Pu/g, 0.5% sol. (1 mg Pu/g, 1.0%
0.05
1
KRI
AEA AEA AEA AEA
(%)
90099 09100 90101 09102
3/6, 5/6-U 8/6-U 3/6, 5/6-U 8/6-U
0.04, 0.024 0.024 0.03 -0.05 0.03 -0.05 0.01
238-Pu 238-Pu 238-Pu 238-Pu
0.09 0.06 0.08 0.17
sol. (1 mg Pu/g, 2.0% sol. (1 mg Pu/g, 4.0%
0.05
Lu Isotopic
175/176-Lu
ORNL
U Tracer
233-U solution (40 mg/g)
KRI
U
0.35
235-U solution (10 mg U/g, 99.994% 235-U)
KRI OCO-4343-88
U
0.04
236-U solution (10 mg U/g, 99.973% 236-U)
KRI OCO-95.048-88
U
0.05
236-U U308 (99.973% 236-U)
KRI OCO-4213-87
U
238-U solution (10 mg U/g, 99.9999% 238-U)
KRI OCO-95.057-89
U
0.04
233/236-U solution
KRI
U
0.04
239-Pu solution 239-Pu metal 240-Pu metal
KRI KRI
242-Pu
0.02 0.01 0.01
Pu Tracer
continued
576
Chapter 11
Table 11.4 (continuation) Use
U/Pu Tracer
Material
Source
Analyte
Accuracy (%)
242-Pu solution (98%) (50 ~tg Pu/g, 98% 242-Pu) 242-Pu metal
RI WRM 009-90
Pu, 242-Pu
0.18,0.002
KRI
0.05
242/244-Pu solution (1:1)
KRI
0.02
239/240-Pu solution
KRI IRMM
0.01
235-U/239-Pu solution 235-U/239-Pu metal
IRMM IRMM LANL Harwell
0.01 0.02
236-U/240-Pu dried nitrate (LSD) 236-U/240-Pu solution 236-U/240-Pu metal
IRMM IRMM IRMM LANL Harwell
233-U/236-U+242/244-Pu solution KRI (QS87) KRI (QS92)
0.01 0.02
U Pu
0.05 0.16
Lu Tracer
176-Lusolution
ORNL
0.05
Cm Tracer
242-Cm
TUI KRI
1
3. C o n t a i n m e n t and surveillance (C&S) a. V i d e o surveillance systems b. Seals 4. Sampling for Destructive Analysis (DA) 5. D o c u m e n t a t i o n 6. C o m m u n i c a t i o n s a. Within inspection teams b. To H Q from inspection teams c. With support aircraft 7. Geographical location T h e list is r e v i e w e d annually and equipment added or deleted based upon operational experience and n e w requirements or new equipment available. Presently the equipment n e e d e d for special and non-routine inspection activities is shown Table 11.5.
International Safeguards
577
Table 11.5 Equipment for special and non-routine inspection activities Description of equipment
Intended use
Gamma-ray detection equipment:
1.
1.1
Large volume portable coaxial HPGe detector (LOAX type)
Precise measurement of weak sources in the field (e.g. smear samples).
1.2
Large area portable detector for the isotopic analysis of weak plutonium and other samples.
Measurement of the isotopic composition of weak plutonium samples.
1.3
Standard high rate portable planar detector for the isotopic analysis of plutonium samples.
High rate plutonium and spent fuel measurements.
1.4
Special HPGe detection system which can be operated without using liquid nitrogen, large area planar detector.
Gamma measurements at remote locations where liquid nitrogen is not available.
1.5
Set of NaI detectors having different sizes of crystals.
Quick low resolution measurements.
1.6
Small size CdTe, CsI or NaI detectors.
Gamma-detection in restricted areas (such as inside fuel assemblies).
.
Neutron detection equipment:
2.1
Bubble detectors.
Spent fuel measurements.
2.2
Neutron radiation survey meter.
General neutron survey.
2.3
Highly sensitive He-3 detectors and fission chambers with moderator.
Neutron measurements.
.
General radiation monitors:
3.1
Radiation survey meter with telescopic extension.
General radiation survey.
3.2
Alpha and beta meters.
Surface contamination monitors.
3.3
High sensitivity underwater radiation survey meter Spent fuel measurements. Also for covering the range from 0.1 mr/hour to 30000 unattended monitoring (if coupled to rad.hour, computer).
4.
Data acquisition and analysis equipment:
4.1
Battery powered portable self-contained mini-multichannel analysers (5).
For use with all of Section 1 detectors for gamma spectrometry and data evaluation on the spot. Can be used with computer above for isotopic identification and analysis.
4.2
Portable X-ray fluorescence detection system.
Identification of different non-nuclear materials under field conditions. continued
Chapter 11
578
Table 11.5 (continuation)
Description of equipment .
Intended use
C&S Equipment:
5.1
Portable video surveillance system.
5.2
Set of mechanical, electronic and fibre optic seals. Maintenance of integrity of Safeguards equipment and/or data during unattended period at a facility (e.g. over night during special inspections).
6.
Equipment for documentation:
6.1
Note book computers
6.2
Video cameras.
6.3
Still photo cameras.
.
Temporary freeze of the status of facility or equipment (e.g. over night during special inspections).
Word processing, spread sheets, etc.
Equipment for communication:
7.1
Portable transceivers
Personnel communication between team members.
7.2
Satellite communication system, including FAX and encryption capabilities.
Communication with HQ in Vienna. Transmission of site situation photos and papers during consultation with HQ.
7.3
Aircraft band transceiver.
Communication with support aircraft.
8.
Miscellaneous equipment:
8.1
Magnetic sensor
To search for large metal items in a building or under ground.
8.2
Infrared camera
To search for hidden heat sources.
8.3
Ultrasonic thickness gauges.
To measure the thickness of concrete walls or steel structure for design verification.
8.4
Ultrasonic or laser length measurement device.
Design verification.
8.5
Satellite positioning system.
Determination of geographic coordinates.
8.6
Different types of high performance flashlights.
8.7
Rugged suitcases with sets of bottles for taking samples.
8.8
Binoculars.
8.9
Radiation protection means (gloves, protective clothes, individual dose rate meters, etc.).
8.10 ICVD; CVD.
Spent fuel attribute verification.
International Safeguards
579
11.3 SAFEGUARDS IMPLEMENTATION IN JAPAN This is presented as an example of a State System of Accounting and Control, SSAC, a national safeguards system. Since the launch of its nuclear use and development programme, Japan has limited nuclear uses to peaceful purposes and has managed the programme under a strict domestic control system. In response to the ratification of the NPT in 1976, the government amended the Nuclear Regulation Law and set up a national system which enabled the country to cope with the IAEA's full scope safeguards. An outline of the system is shown in Fig. 11.2. Japan has an impressive list of nuclear materials and facilities which relate to safeguards. Of the total electricity generation of 797.8x 109 k W / h at the end of fiscal 1992, 28.2%, or approximately 225x109 kW/h generated comes from nuclear power. According to energy demand forecasts, 5 0 5 • kW of power generation will be targeted for the year after 2000. Japan's nuclear source materials and nuclear fuel materials, which support this nuclear power generation, are all subject to safeguards under the Nuclear Regulation Law and the NPT. At the end of 1992, Japan held roughly 3500 t of natural and depleted uranium, about 9400 t of low enriched uranium (LEU) and 33.5 t of plutonium. Figure 11.3 indicates the locations of the country's major facilities, including nuclear power generation and a part of the nuclear fuel cycle. International inspection Reports on material accountancy, I 4 - inspection results "~ IAEA 411..Reports on conclusion of verification 9
.
,.
,
~
Goverment ~'I :1 I of Japan I
~ Consultations on implementing safeguards
National inspection
,.-
- Book audit - Item count - Measurement - Sample taking - Containment / surveillance application
Nuclear facilities
~-- Reports on material aqccountancy-~
Bilateral agreement countries USA UK France Canada Australia China
Nuclear Material Control Center
Reports on inventory ~ " and transfer lists ~tl'
Safeguard division, Notification on inventory...i P STA confirmation of transfer
- Information treatment by centralized control of material accountancy data and preparation of material accountancy - Analysis of U & Pu in samples taken
Consultation on ~ ' - safeguards
v
- Calibration & maintenance of inspection instruments
Fig. 11.2. The safeguards system in Japan (after Hayashi et al., 1994).
580
Chapter 11 14
Toma" ) ) ~ Rokkasho lga
~ ]
~Onagawa
Tsuru _ 0"/ ~ Fukusima Mihama~~ahshiaW~Z~/ui,/ _ I ~ Tokai Ohi \ \ ] x..- TOKIO,,,2~t~O-Arai T a k a C a ~ . ~ ^ ~Kuriham a Shimant~II1 -- ~ Hamaoka j i ~ ~.~AKA uenK~J~ l ~Sendai-" Fig. 11.3.Majornuclearfacilitiesin Japan. In August 1993, the number of facilities which deal with nuclear materials under safeguards reached 248, including the major facilities indicated in Fig. 11.3. The Nuclear Regulation Law mandates all of these facilities to receive permits for installation, business, use and change. In addition, the Law mandates that these facilities receive permits for procedures which stipulate the implementation of strict accountancy within these facilities. A breakdown of the types of facilities is indicated in Table 11.6 (after Hayashi et al., 1994). Characteristic of Japanese safeguard-related facilities is that uranium enrichment, conversion and fuel fabrication facilities, which support nuclear power generation through the use of low enriched uranium fuel, as well as light water reactors, are operated on a commercial scale; these facilities also cover a wide range of facilities of established nuclear fuel cycles, including fast breeder reactors, advanced thermal reactors, a spent fuel reprocessing facility and a mixed oxide (MOX) conversion and fuel fabrication facility. These aim at establishing nuclear power generation through the use of plutonium fuel. With respect specifically to the plutonium cycle, which originates in the Tokai Reprocessing Plant (TRP), an annual reprocessing capacity of approximately 90 t U is maintained. Meanwhile, the MOX conversion facility, which
International Safeguards
581
Table 11.6 Breakdown of the types of facilities (as of August 1993) (after Hayashi et al., 1994) Facility type
Number
Nuclear power plants
46
Research reactors, critical assemblies Mining and conversion facility
23 1
LEU fabrication plants
5
Enrichment plants
2
Reprocessing plant
1
Pu conversion development facility
1
Pu fabrication development facilities Research facilities Storage facility Subtotal Outside facilities Total
2 19 1 101 147 248
takes nuclear proliferation resistance characteristics into consideration, has adopted a microwave denitration method, unique to Japan, to produce MOX fuel as a joint effort with TRP. Furthermore, fuel for the MONJU fast breeder reactor (FBR) is being produced at the Plutonium Fuel Production Facility (PFPF). Located within the Tokai Works of the Power Reactor and Nuclear Fuel Development Corporation (PNC), these facilities are controlled by strict safeguards measures. The MONJU FBR, built in Tsuruga by the Sea of Japan, is currently in the final commissioning stage before criticality. Concerning safeguards, the government of Japan and the IAEA were able swiftly to arrange facility attachment before fuel delivery. Future projects include the Rokkasho Reprocessing Plant (RRP), which is to become the base for Japanese commercial use of plutonium. The facility is currently under construction to meet the planned operational launch date of 2000. Regarding safeguards for this facility, following the conclusion reached at the recent large-scale reprocessing (LASCAR) meeting, a study of specific approaches to be applied, and authentication and production of safeguards equipment which are expected to be adopted are being implemented with the co-operation of the IAEA. Japan has set up during the past 22 years a unique SSAC, in which the government' s administrative organisations have a very close business relationship as regards technical aspects of the implementation of safeguards. As far as the processing of safeguards accountancy information is concerned, the government has authorised the Nuclear Material Control Center (NMCC) to conduct business on behalf of the Government in accordance with the Nuclear Regulation Law. The Safeguards Analytical Laboratory located at Tokai-mura has been constructed by the Government and is being operated
582
Chapter 11
by the NMCC under contract to the Government. Technical support, including maintenance and operation of the non-destructive assay equipment at the inspection sites, is also entrusted to the NMCC. Japan's safeguards system begins with the approval of the control and accountancy procedures, applications for which had been submitted by facility operators before the start of operation of each nuclear facility. This entails scrutiny and approval of the procedures on the control and accountancy system, including records, the reporting system, the system for controlling measurement errors and the system for calculating material unaccounted for (MUF). Facilities for which approval has been granted are required to report to the Science Technology Agency (STA), at each designated period, on the handling of nuclear source materials, nuclear fuel materials and nuclear equipment under international control. These collected reports are transmitted to the NMCC and checked for the self-consistency of each operator' s report as well as the mutual consistency among different facilities. After this has been confirmed, the reports are modified to IAEA formats and sent to the IAEA via diplomatic channels. At the same time, submission of information on the country specific inventory based on bilateral nuclear co-operation agreements is mandatory. This information is also periodically sent to supplier countries. The major aim of the country's control and accountancy system is to maintain the quality assurance of accountancy information that is vital to enhance the reliability of safeguards (Hayashi et al., 1994).
11.4 UN SECURITY COUNCIL RESOLUTIONS 687 AND 707 On 3 April 1991, the United Nations Security Council passed Resolution 687 (and a number of subsequent resolutions) on Iraq, which placed an additional and new demand on the IAEA resources and subsumed all other safeguards activities in that country. For the first time a Member State with declared nuclear facilities (which had been regularly inspected by the IAEA under a Safeguards Agreement) had clandestinely established additional nuclear facilities and had begun to produce nuclear material in violation of the Agreement. By Security Council resolution 687 the IAEA was entrusted, inter alia, with the task of carrying out immediate on-site inspections of Iraq's nuclear capabilities based on Iraq's declarations and on the designation of additional locations by the Special Commission established pursuant to Paragraph 9(b) of that resolution. Pursuant to the resolution, Iraq was to submit to the Secretary General of the United Nations and to the Director General of the IAEA within 15 days of adoption of the resolution, a declaration of the locations, amounts and types of nuclear weapons, nuclear weapons-usable material and any subsystems or components and any research, development, support or manufacturing facilities related to nuclear weapons or nuclear weapons-usable material. By letters of 18 and 27 April, Iraq submitted to the Secretary General of the United Nations and to the Director General of the IAEA a list that included all material
International Safeguards
583
previously declared to the IAEA under the Safeguards Agreement between Iraq and the Agency for the Application of Safeguards in Connection with the Treaty on the Non-Proliferation of Nuclear Weapons. Iraq also reported a uranium concentrate production plant. The first Agency inspection under resolution 687 took place from 14 to 22 May. The second inspection mission arrived in Iraq on 22 June. During the course of that inspection, the inspection team was denied access on 23, 25 and 28 June to sites designated by the Special Commission. At the request of the Security Council, a high-level mission composed of the Director General, the Chairman of the Special Commission and the Under Secretary General for Disarmament Affairs went to Iraq on 30 June to secure immediate and unimpeded access to all sites and objects which the team had endeavoured to inspect. Pursuant to the mission' s visit, Iraq indicated its intention to submit an additional list of nuclear items relevant to Security Council resolution 687. By a letter dated 7 July 1991 from the Foreign Minister of Iraq to the Secretary General and copied to the Director General, Iraq submitted an additional list of nuclear equipment and material in its possession. A review of the 1 July letter suggested that Iraq had been in non-compliance with its obligation under its Safeguards Agreement with the Agency, in particular with regard to Article 34(c), which requires nuclear material of a composition and purity suitable for fuel fabrication or isotopic enrichment, and any nuclear material produced at a later stage in the nuclear fuel cycle, to be subject to all of the safeguards procedure provided for in the Agreement; with regard to Article 34(b), which requires notification to the Agency of imports of material containing uranium or thorium which has not reached the stage provided for in Article 34(c), unless the material is imported for specifically non-nuclear purposes; and with regard to the requirements set forth in Article 42 of the Agreement and Code 3.1.2 of the Subsidiary Arrangements General Part concerning submission of design information in respect of facilities. On 9 and 11 July, the Director General wrote to the Foreign Minister of Iraq requesting urgent comments on the matter. On 12 July 1991, the Director General received from the Iraqi Resident Representative a letter from the Foreign Minister responding to the 9 July letter of the Director General, and providing observations and clarifications with regard to the enrichment activities in this context. In that letter the foreign Minister stated that the enrichment activities were not subject to safeguards because they were still in the early stage of research and development, because there was no facility for isotope separation as defined in Article 98 of the Safeguards Agreement, and because the amount of material produced was less than a significant quantity. He further stated that in the interpretation of Iraq, Article 34(c) referred only to imports of nuclear material of a composition and purity suitable for fuel fabrication, and that as the material in question (a half kilogramme of uranium enriched to 4% 235U) had been produced domestically, Article 34(c) was not applicable. Alternatively, it was maintained, as the material had not been produced in a "facility" as defined under the Agreement, Article 34(c) was not applicable.
584
Chapter 11
On 13 July 1991, the IAEA received a further communication from the Foreign Minister responding to the Director General's letter of 11 July. In that communication, it was stated that, as regards the yellowcake and uranium dioxide referred to in Table 9 of the Iraqi declaration of 7 July, the material was meant for non-nuclear use, and accordingly, the Agency had not been notified of such material in accordance with Article 34(b) of the Safeguards Agreement. With respect to the uranium hexafluoride, Iraq's view was that the quantity was insignificant and, hence, was not required to be notified or inspected. With regard to the uranium tetrachloride, it was maintained that the amount was the normal amount required for feeding experimental separators using the electromagnetic enrichment process and did not require notification. Finally, it was noted that the government of Iraq considered its letter of 7 July 1991 a "corrective" measure meeting the requirements of Article 19 of the Safeguards Agreement. The Safeguards Agreement between Iraq and the Agency, a standard INFCIRC/ 153-type agreement, requires the application of safeguards to all nuclear material in all peaceful nuclear activities in the State. Pursuant to Article 34(b), when any material containing uranium or thorium which has not reached the stage of the nuclear fuel cycle described in Article 34(c) is imported, the State is required to inform the Agency of the quantity and composition of such material, unless the material is imported for specifically non-nuclear purposes. Article 34(b) applies, inter alia, to imports of yellowcake. Pursuant to Article 34(c) of the Agreement, when any nuclear material of a composition and purity suitable for fuel fabrication or for isotopic enrichment leaves the plant or processing stage in which it has been produced, the nuclear material becomes subject to the other safeguards procedure specified in the Agreement. In addition, Article 34(c) requires the application of such other safeguards procedures to nuclear material of the composition and purity suitable for fuel fabrication or for isotopic enrichment, or any other nuclear material produced at a later stage in the nuclear fuel cycle, which is imported into the State. Article 34(c) would include, therefore, any domestically produced or imported uranium dioxide of a purity suitable for fuel fabrication, uranium hexafluoride and enriched uranium, as well as any nuclear material produced therefrom. Peaceful nuclear activities include research and development activities in the nuclear field. The Safeguards Agreement provides no automatic exemption from the application of safeguards to nuclear material which is used or intended for use in such activities. Although the Agreement provides a mechanism for exempting nuclear material of certain quantities (Article 36) and for certain uses (Article 37), both mechanisms require the material to have been subject to safeguards and then, upon request by the State, to be exempted therefrom by the Agency. The production of nuclear material in a location which is not a "facility", as defined in Article 98(1) of the Safeguards Agreement, does not exempt the State from reporting such material to the IAEA. Article 49 of the Agreement requires that when nuclear material is customarily used outside facilities, the State must provide the Agency with,
International Safeguards
585
inter alia, a general description of the use of such material, its geographic location and the user' s name and address for routine business purposes. The State is also obliged to provide a general description of the existing and proposed procedures for nuclear material accountancy and control. The technical objective of the safeguards procedures set forth in the Agreement, as described in Article 28, which relates to the timely detection of diversion of significant quantities of nuclear material, and deterrence of such diversion by the risk of early detection, does not imply that any minimum quantity of nuclear material is required for the application of safeguards. On the basis of the information and comments provided to the Agency by the letters of 7, 10 and 12 of July, and on the basis of reports by the Agency Chief Inspector, it was concluded that Iraq failed to report to the Agency the existence in Iraq of nuclear material of the composition and purity provided for in Article 34(c), including the uranium dioxide, the uranium tetrachloride and the uranium hexafluoride identified in the annexes to the 7 July letter. In addition, it was concluded that Iraq failed to report the half kilogramme of enriched uranium as nuclear material past the stage provided for in Article 34(c). It was concluded, therefore, that Iraq had not been in compliance with its obligations under the Safeguards Agreement with the IAEA, in particular with respect to the obligation to accept safeguards on all nuclear material in all peaceful nuclear activities in Iraq. The United Nations Security Council Resolution 687, the so-called cease-fire resolution, inter alia, mandated the destruction of all weapons of mass destructionm chemical, biological, ballistic and nuclear--existing in Iraq. The IAEA was given sole responsibility under this resolution to destroy, remove or render harmless not only nuclear weapons but also any existing capability to acquire them, including prohibited precursor materials such as enriched uranium, plutonium and all facilities, equipment and materials used for their production. On 15 April 1991 the Director General created an Action Team to implement Resolution 687: specifically to pursue the identification and elimination of activities in Iraq aimed at the acquisition of nuclear weapon capability and the ongoing monitoring and verification of Iraq to ensure no resumption of a nuclear weapons programme (Zifferero, 1994). It must be noted that the cease-fire resolution, which Iraq was obliged to accept, provided the IAEA with the legal basis to implement a countrywide search of a highly intrusive nature (the any-place-any-time approach), with inspection teams being granted fights similar to those of an occupying army. Any attempt made by the Iraqi side to limit or ignore these rights has met with an immediate, firm reaction by the UN Security Council, and, in a couple of cases, with military retaliation. The exercise of these rights has played an important role in the discovery and dismantling of the clandestine programme. While it is not conceivable that similar rights might be granted under ordinary circumstances, there are other ingredients which contributed to success in Iraq, the inclusion of which in the search for undeclared activities and material in other States may be less objectionable. As examples, one may mention the access to
Chapter 11
586
information provided by third parties and the extensive use of modem analytical techniques in environmental sampling. Many nuclear inspection teams have visited Iraq to inspect facilities, interview key personnel, inventory nuclear materials, identify prohibited items and carry out destruction and removal operations. The Iraq nuclear weapons programme was carried out at nine dedicated sites. On these sites there were dozens of major processing buildings that represented an investment equivalent to several billion US dollars. Most of these buildings were either destroyed during the war or demolished under inspection teams' supervision. What now remain at these sites are offices, warehouses and light buildings with no unique capabilities (see map of Iraq for the sites of interest, Fig. 11.4). The IAEA will gradually phase in all the elements of the ongoing monitoring and verification plan, which consists of a package of activities and techniques such as: on-site inspections, both routine and unannounced; 9 extended and continuous presence of inspectors in Iraq; 9 close control of the natural uranium inventory remaining in Iraq; 9 use of seals, tags and video surveillance; 9 monitoring dual-use items (equipment and materials); 9 periodic environmental sampling of air, water, soil, biota and vegetation; 9 use of advanced sensors and systems to detect signatures of prohibited activities (enrichment, processing, weaponization) on mobile land-based and aerial platforms; 9 satellite and low-altitude imagery; 9 import/export monitoring; 9 power line monitoring; 9 interviews with key personnel; 9 assessment and follow-up of information received from Member States. 9
On 15 August 1991 a new resolution was adopted by the U.N. Security Council" Resolution 707 which obliges Iraq, inter alia, to "halt all nuclear activities of any kind, except for use of isotopes for medical, agricultural or industrial purposes, until the Security Council determines that Iraq is in full compliance with resolution 707 and with paragraphs 12 and 13 of resolution 687, and the IAEA determines that Iraq is in full compliance with its safeguards agreement with that Agency". So long as the proscriptions under resolution 707 remain operational, the Agency will secure the nuclear material, equipment and facilities which Iraq is allowed to keep and use under the terms of resolution 687 and verify that they are not used for any nuclear activity except as permitted under resolution 707. The Agency will also verify that nuclear material and isotopes are not produced indigenously by Iraq, and that isotopes held or imported by Iraq are used only for medical, agricultural or industrial purposes. The list of peaceful applications of isotopes imported from other states after prior approval by the IAEA which are permitted includes:
International Safeguards
587
Fig. 11.4. Iraq: uranium mining, production and processing sites; sites related to uranium enrichment; sites related to weaponization. Uranium mining, production, processing sites: A1 Tuwaitha Nuclear Research Centre, A1Jesira, A1 Mosul (processing), A1 Qaim, Tikrit (yellowcake storage), Akashat (phosphate and uranium processing). Sites related to uranium enrichment: A1 Tuwaitha Nuclear Research Centre, A1 Tarmiya (EMIS), Ash Sharqat (planned), A1Jesira, A1Radwan, A1Amir, A1Furat (centrifuge production), Daura (manufacturing), Badr (engineering complex), Salladine, Nassar Works. Sites related to weaponization: A1Tuwaitha Nuclear Research Centre, A1Atheer, AI Qa Qaa, Hatteen (high explosive test site), A1 Hadre (high explosive test site).
588
Chapter 11
1. Agricultural applications 1.1 Soil fertility, irrigation and crop production 1.2 Plant breeding and genetics 1.3 Animal production and health 1.4 Insect and pest control 1.5 Food preservation 1.6 Other uses as approved by the IAEA 2. Industrial applications 2.1 Radiography and other non-destructive testing methods 2.2 Industrial process control and quality control 2.3 Radiotracer applications in oil, chemical and metallurgical processes 2.4 Development of water and mineral resources 2.5 Industrial radiation processing 2.6 Other uses as approved by the IAEA 3. Medical applications 3.1 Diagnostic and therapeutic medicine including dosimetry 3.2 Radiotherapy by teletherapy and brachytherapy 3.3 Nutrition and health-related environmental studies 3.4 Other uses as approved by the IAEA It is of interest is to present here a list of items which were to be reported to the IAEA. Security Council Resolution 707 demands that Iraq, inter alia, halt all nuclear activities of any kind, except for certain uses of isotopes, until the Security Council determines that Iraq is in full compliance with the provisions of Resolution 707 and paragraphs 12 and 13 of Resolution 687 and that the IAEA determines that Iraq is in full compliance with the provisions of its safeguards agreement with the IAEA. Once these determinations have been made affirmatively by the Security Council and by the IAEA, Iraq may seek to initiate the nuclear activities which are not prohibited by Resolution 687. Approval by the Security Council for Iraq to initiate one or more of these nuclear activities may necessitate a corresponding amendment to this list. The list contains articles specifically prohibited to Iraq under Resolution 687 as well as others which may be prohibited if they are used, or are to be used, in activities prohibited under Resolution 687. 1. Source materials Uranium containing the mixture of isotopes occurring in nature; uranium depleted in the isotope 235; thorium; any of the foregoing in the form of metal, alloy, chemical compound or concentrate. 2. Special fissionable materials Special fissionable materials which fall within the definition of nuclear-weaponusable materials are prohibited. Plutonium-239; uranium-235; uranium-233; uranium enriched in the isotopes 235 or 233; any material containing one or more of the foregoing.
International Safeguards
589
3. Nuclear-weapon-usable materials Nuclear material that can be used for the manufacture of nuclear explosive components without transmutation or further enrichment, such as plutonium containing less than 80% plutonium-238, uranium enriched to 20% uranium-235 and uranium-233 or more; any chemical compound or mixture of the foregoing. Plutonium, uranium-233 and uranium enriched to less than 20% uranium-235 contained in irradiated fuel do not fall into this category. 4. Equipment or materials referred to in Section 2 of Memorandum B of INFCIRC/ 209/Rev. 1 and in the Annex to INFCIRC/209/Rev. 1 All items included in INFCIRC/209/Rev. 1 which are used for enrichment and reprocessing are prohibited. Any item to be used in any activity listed in items 2.1 to 2.9 of Annex 1 is also prohibited. 5. Equipment and materials used in uranium enrichment including 5.1 Rotor fabrication and assembly equipment and bellows-forming mandrels and dies (a) Rotor assembly equipment specially designed or prepared for assembly of gas centrifuge rotor tube sections, baffles, and end caps. Such equipment includes specially designed precision mandrels, clamps, and shrink fit machines. (b) Rotor straightening equipment specially designed or prepared for alignment of gas centrifuge rotor tube sections to a common axis. (c) Bellows-forming mandrels and dies, two-piece cylindrical with a single indented circumferential convolution bisected by the two halves. 5.2 Centrifugal balancing machines Centrifugal balancing machines, fixed or portable, horizontal or vertical. 5.3 Filament winding machines Filament winding machines in which the motions for positioning, wrapping, and winding fibres are coordinated and programmed in three or more axes, specially designed to fabricate composite structures or laminates from fibrous and filamentary materials and capable of winding cylindrical rotors. 5.4 Centrifuge housing/recipients Components to contain the rotor tube assembly of a centrifuge enrichment machine 5.5 Aluminium, high-strength tube Cylindrical tubing in semifabricated or finished forms made of aluminium alloy. 5.6 Fibrous and filamentary materials (high strength) Fibrous and filamentary materials for use in composite structures 5.7 Maraging steel Maraging steel (high strength) with an ultimate tensile strength of 2.050• 109 N/m 2 (300.000 psi) or more.
Chapter 11
590
5.8 Titanium Cylindrical tubing in semi-fabricated forms made of high-strength titanium alloys 5.9 Spin-forming and flow-forming machines specially designed or adopted for use with numerical or computer controls and specially designed parts and accessories therefor. 6. Chlorine trifluoride 7. Electrolytic cells for fluorine production and specially designed parts and accessories therefore 8. Mass spectrometers for uranium hexafluoride as follows. 8.1 Mass spectrometers, magnetic or quadruple: 8.1.1 Instruments having all of the following characteristics: (a) Resolution of less than 1 atomic mass unit (amu) for molecular masses greater than 320 amu; and (b) Electron-bombardment ionisation source; and 8.1.2 Instruments having any of the following characteristics: (a) Molecular beam ion sources; (b) Ion source chambers constructed of or lined with nichrome or monel, or nickel plated; (c) A collector system suitable for simultaneous collection of two or more isotopic species; and 8.2 Source for mass spectrometers having any of the following characteristics: (a) Molecular beam source; (b) Ion source chambers constructed of or lined with nichrome or monel, or nickel plated; or (c) Sources for mass spectrometers designed especially for use with UF 6. 9. Uranium hexafluoride-resistant gauges 10. Uranium hexafluoride-resistant valves Valves, with a bellows seal, wholly made of or lined with aluminium, nickel, or alloy containing nickel, either manually or automatically operated and specially designed parts or accessories therefor. 11. Lasers and equipment containing lasers as follows (a) Copper vapour lasers with 40 W average output power; (b) Argon ion lasers with greater than 40 W average output power; (c) Nd:YAG lasers that can be frequency doubled and after doubling have an average power output at the doubled frequency greater than 40 W; (d) Tuneable pulsed dye laser amplifiers and oscillators, except single-mode oscillators, with an average power greater than 30 W, a repetition rate greater than kHz and a wavelength between 500 nm and 700 nm;
International Safeguards
591
(e) Tuneable pulsed single-mode dye oscillators capable of an average power greater than 1 W, a repetition rate greater than 1 kHz, a pulse width less than 100 ns, and a wavelength between 500 nm and 700 nm. (f) Alexandrite lasers with a bandwidth of 0.005 nm or less, a repetition rate greater than 124 Hz, and an average output power greater than 30 W; (g) Pulsed carbon dioxide lasers with a repetition rate greater than 250 Hz, an average output power greater than 2.5 kW, and a pulse length less than 200 ns; (h) Pulsed excimer lasers (XeD, XeC1, KrF) with a repetition rate greater than 250 Hz and an average output power greater than 500 W; (i) Free electron lasers. 12. Pipes, valves, fittings Pipes, valves, fittings, heat exchangers, or magnetic, electrostatic, or other collectors made of graphite or coated in graphite, yttrium, or yttrium compounds resistant to the heat and corrosion of uranium vapour. 13. Resins and organic complexing agents capable of separating isotopes of uranium Chemical exchange resin developed for the separation of isotopes of uranium and other fissile materials and organic complexing agents developed for the same purpose. 14. Solvent extraction equipment suitable for use in the separation of uranium isotopes. 15. Ordinary and superconducting electromagnets capable of creating magnetic fields of more than 2 teslas (20 kilogauss) as follows. (a) ordinary and solenoidal superconductive electromagnets or more than 300 mm inner diameter except such magnets shipped as integral parts of medical nuclear magnetic resonance (NMR) imaging systems. (b) ordinary and superconductive electromagnets with a diameter of 500 mm or greater except such magnets shipped as integral parts of NMR systems. 16. Process control systems configured for use in uranium enrichment, as follows: (a) Computer systems configured to read process variables, compute control levels, and automatically adjust process variables for such units; (b) Arrays of instrumentation for monitoring process variables such as temperature, pressure, pH, fluid level, and flow rate selected for specific production process and designed to operate in the hostile environment required by each process. 17. Equipment specially designed for the preparation of feed materials for enrichment processes, including the preparation of UF 6 and UC14. 18. Feed materials for enrichment processes including UF 6 and UC14. 19. Nuclear reactors, including critical and sub-critical assemblies, reactor equipment and materials. Reactor systems, sub-systems, equipment and components 19.1 Reactor vessels: Reactor vessels, including pressurised and unpressurised types.
Chapter 11
592
19.2Reactivity control mechanisms, devices and systems: .Reactivity control mechanisms, devices and systems, including manual, electro-mechanical, hydraulic, pneumatic and chemical injection/removal-type systems. 19.3 Reactor process monitoring, measurement and control systems: Reactor process monitoring, measurement and control systems, sub-systems and components. All analog and digital process control computers and hydraulic and pneumatic process monitoring and control instruments and equipment. 19.4 Reactor fuel charging and discharging systems: Reactor fuel charging and discharging systems and equipment, including manual, electro-mechanical, hydraulic and pneumatic systems and components. 19.5 Calandrias: Calandrias, calandria tubes, pressure tubes and other fuel channel assemblies and components. 19.6Primary and secondary heat transport and removal systems: Primary and secondary heat transport and removal systems, including steam generators, heat exchangers, coolant purification, coolant recovery, high and low pressure injection and circulating pumps, pressure relief devices and other pressureretaining components especially designed, manufactured or prepared for use in such systems. 20. Plants and equipment used in reprocessing. Process control systems configured for use in reprocessing, as follows: (a) Computer systems configured to read process variables, compute control levels, and automatically adjust process variables for such units; (b) Arrays of instrumentation for monitoring process variables such as temperature, pressure, pH, fluid level, and flow rate selected for specific production process and designed to operate in the hostile environment required by each process. 21. Hot cells and associated equipment Hot cells and associated equipment for the handling and processing of irradiated fuel on any scale. 22. Other equipment for the reprocessing of irradiated fuel Equipment for the reprocessing of irradiated fuel by methods other than solvent extraction, e.g., ion-exchange, fluoride volatility, pyrometallurgical. 23. Reprocessing waste treatment Plant and equipment for the treatment of wastes from reprocessing. 24. Plants and equipment used for the following processes (a) Prospecting for ores containing source materials; (b) Mining of ores containing source materials; (c) Separation of source material from ores and other naturally occurring materials to form concentrates; (d) Preparation of metals, alloys, or any chemical compound containing source material or uranium enriched to less than 20% in uranium-235;
International Safeguards
593
(e) Fabrication of source material or uranium enriched to less than 20% in uranium-235 into a form suitable for irradiation in a nuclear reactor; (f) Treatment of wastes from mining, conversion and fabrication processes and plants. 25. Turning machines Turning machines (lathes) having one or more of the following characteristics: (a) Vacuum chucks suitable for holding hemispherical parts; (b) Machines installed within glove boxes. 26. High temperature furnaces Vacuum or controlled environment (inert gas) furnaces including induction, arc, plasma and electron beam furnaces, capable of operation above 700~ and especially designed power supplied therefor. 27. Crucibles resistant to liquid fissile metals Crucibles made of materials resistant to liquid fissile metals and designed to avoid nuclear criticality. 28. Isostatic presses Isostatic presses capable of achieving a maximum working pressure of 69 MPa or greater and specially designed dies and moulds, components, accessories and controls and "specially designed software" therefor. 29. Beryllium as follows: (a) Metal; (b) Alloys containing more than 50% of beryllium by weight; (c) Compounds containing beryllium; (d) Manufactures thereof; and (e) Waste and scrap; except (a) Metal windows for X-ray machines, (b) Oxide shapes in fabricated or semi-fabricated forms specially designed for electronic component parts or as substrates for electronic circuits; (c) Naturally-occurring compounds containing beryllium. 30. Calcium High purity calcium containing both less than 0.1% by weight of impurities other than magnesium and less than 10 ppm (parts per million) of boron. 31. Lithium (a) Lithium enriched in lithium-6; (b) Facilities or specialised equipment for the separation of the lithium-6 isotope; except for use in thermoluminescence dosimetry.
594
Chapter 11
32. Magnesium High purity magnesium containing both less than 0.02% by weight of impurities other than calcium and less than 10 ppm (parts per million) of boron. 33. Tantalum Tantalum sheet with a thickness of 2.5 mm or greater. 34. Plutonium, uranium-233 and enriched uranium contained in irradiated fuel. 35. Tungsten Parts made of tungsten, tungsten carbide, or tungsten alloys (greater than 90% tungsten) having a mass greater than 20 kg; except parts specifically designed for use as weights or gamma-ray collimators. 36. Hafnium Hafnium in any metallic, alloy or oxide form. 37. Boron Boron and boron compounds and mixtures in which the boron-10 isotope is more than 20% of the total boron content. 38. Hydrodynamic testing facilities Hydrodynamic test facilities capable of handling the detonation of high explosive charges of 1 kg or greater and suitable for use of appropriate diagnostic instrumentation. 39. Computers Computer centers and networks using hydrodynamics codes, neutronic codes, and/or equation-of-state and nuclear data files. 40. Flash X-ray equipment Flash X-ray generators or pulsed electron accelerators with peak energy of 500 keV or greater. 41. Gun Systems for large masses Gun systems for accelerating large masses (greater than 5 kg) to low velocity using chemical propellants similar to those used in artillery, usually associated with timing, velocity, and other diagnostics. 42. Shells, hollow spheres or portions thereof Shells, hollow spheres, or portions of spheres made of high explosives or metals listed in 2 and moulds for such parts. 43. Photographic equipment (a) Mechanical framing cameras with recording rates greater than 225.000 frames per second; streak cameras with writing speeds greater than 0.5 mm per microsecond; and parts and accessories thereof, including synchronising
International Safeguards
595
electronics specially designed for this purpose and rotor assemblies (including turbines, mirrors, and bearings). (b) Electronic streak cameras capable of 50 ns or less time resolution and framing cameras capable of 50 ns or less frame exposure time including single-frame cameras, and streak and framing tubes usable in such cameras. 44. Transient recorders and/or digital oscilloscopes Transient recorders and/or digital oscilloscopes using analog-to-digital conversion techniques capable of storing transients by sequentially sampling one-short input signals at successive intervals of less than 20 nanoseconds (greater than 50 million samples per second), digitising to 8 bit or greater resolution, and storing 256 or more samples. 45. Analog oscilloscopes and cameras (a) Analog oscilloscopes suitable for triggered single-sweep operation with a bandwidth greater than 250 MHz and associated oscilloscope cameras; (b) Analog oscilloscopes in the 30-250 MHz range and associated oscilloscope cameras. 46. Specialised equipment for hydrodynamic experiments 47. Detonators and multipoint initiation systems: (a) Electrically driven explosive detonators of the types exploding bridge (EB), exploding bridgewire (EBW), slapper, or exploding foil initiators (EFI); (b) Specially designed parts or bodies for any of the detonators described above; or (c) Arrangements of multiple detonators designed to nearly simultaneously initiate an explosive surface from a single firing signal. (d) Explosive lenses designed to uniformly initiate the surface of a high-explosive charge. 48. Firing sets and equivalent high-current pulsers (for controlled detonators) (a) Explosive detonator firing sets designed to drive multiple controlled detonators; (b) Modular electrical pulse generators (pulsers) designed for portable, mobile, or rugged use (including xenon flashlamp drivers) having the following characteristics: 9 capable of delivering their energy in less than 15 microseconds; 9 having an output greater than 500 A; and 9 having a rise time of less than 10 microseconds into loads of less than 5 ohms. 49. High explosives including the following: (a) Cyclotetramethylenetetranitramine (HNM); (b) Cyclotrimethylenetrinitramine (RDX); (c) Triaminotrinitrobenzene (TATB); (d) Pentaerythritoltetranitrate (PETN), except when contained in pharmaceuticals;
596
Chapter 11
(e) Hexanitrostilbene (HNS), except when contained in pharmaceuticals. 50. Neutron generator systems utilising electrostatic acceleration to induce a tritiumdeuterium nuclear reaction; and specially designed parts (including tubes) thereof. 51. Tritium and tritium related plants, equipment, and materials (a) Tritium, including compounds and mixtures containing tritium in which the ratio of tritium to hydrogen by atoms exceeds 1 part in 1000. except tritium in luminescent devices (e.g. safety devices installed in aircraft, watches, runway lights) (b) Facilities or plants for the production, recovery, extraction, concentration, or handling of tritium, and equipment and materials suitable for use therein, including the following: 9 Tritium storage, separation, purification, and pumping systems using metal hydrides as the storage, pumping or purification medium; 9 Pumps or compressors that are constructed without plastic parts and which are designed so that lubricating oils are not in contact with the process gas. 52. Deuterium and deuterium-related plants, equipment and materials (a) Deuterium including compounds and mixtures containing deuterium in which the ratio of deuterium to hydrogen by atoms exceeds 1 part in 5000. (b) Facilities or plants for the production, recovery, extraction, concentration or handling of deuterium, and equipment and materials suitable for use therein (c) Compressors and blowers specially designed or prepared to be corrosion resistant of HzS and having all of the following characteristics: i) an inlet operating pressure of 260 to 280 psi-gauge, with a differential pressure between outlet and inlet of approximately 30 psi; ii) suction volume of 120.000 scfm; iii) capable of sustaining the above inlet pressure and suction volume in HzS gas saturated with water vapour. (d) Specialised packings made of phosphor bronze mesh designed for use in vacuum distillation towers, suitable for use in separating heavy from light water. 53. All alpha-emitting radionuclides and equipment containing alpha-emitting radionuclides meeting all of the following specification: (a) The radionuclides have an alpha half-life of 10 days or greater but less than 200 years; (b) The radionuclides are contained in compounds or mixtures with a total alpha activity of 37 GBq per kilogram (1 curie per kilogram) or greater; and (c) The radionuclides have a total alpha activity of 3.7 GBq (100 millicuries) or greater;
International Safeguards
597
except radionuclides in medical implant devices. 54. Photomultiplier tubes of the following descriptions: (a) An anode pulse rise time of less than 1 ns; or (b) Containing microchannel plate electron multipliers. 55. Capacitors with either of the following sets of characteristics: A voltage rating greater than 1.4 kV having all of the following characteristics: (1) Energy storage greater than 10 J; (2) Capacitance greater than 0.5 ~tF; and (3) Series inductance less than 50 nil; or A voltage rating greater than 750 V having both of the following characteristics" (1) Capacitance greater than 0.25 ~tF; and (2) Series inductance less than 120 nil. 56. High-purity (99.99%) bismuth with very low silver content (less than 10 parts per million) 57. "Robots" and specially designed robot controllers, and robot "end-effectors" having any of the following characteristics: (a) Specially designed to comply with national safety standards applicable to explosive environments (for example, meeting electrical code ratings for explosive environments); (b) Specially designed or rated as radiation hardened more than necessary to withstand normal industrial (i.e., non-nuclear industry) ionising radiation. 58. Pulse amplifiers with gain greater than 6 decibels and with a baseband bandwidth greater than 500 megahertz (having the low frequency half-power point at less than 1 MHz and the high frequency half-power point at less than 1 MHz and the high frequency half-power point greater than 500 MHz) and output voltage greater than 2 volts into 55 ohms or less (this corresponds to an output greater than 16 dbm in a 50 ohm system). 59. Switching devices, as follows: (a) Cold-cathode tubes (including gas krytron tubes and vacuum sprytron tubes), whether gas filled or not, operating similarly a spark gap, containing three or more electrodes, and having all of the following characteristics: (1) Anode peak voltage rating of 2500 V or more; (2) Anode peak current rating of 100 A or more; and (3) Anode delay time of 10 microseconds or less; (b) Triggered spark-gaps having an anode delay time of 15 microseconds or less and rated for a peak current of 500 A or more: (c) Hydrogen~ydrogen-isotope thyratron of ceramic-metal construction and rated for a peak current of 500 A or more.
598
Chapter 11
60. Vibration test equipment using digital control techniques and feedback or closed loop test equipment and software therefore capable of vibrating a system at 10 g RMS or more between 20 Hz and 2000 Hz, imparting forces of 50 kN (11,250 lbs) or greater. 61. Electronic digital computers with a "composite theoretical performance" (CTP) of 12.5 million theoretical operations per second (Mtops) or greater except: (a) Computers contained in or associated with other equipment or systems where the computers are essential for the operation of the other equipment or systems and the computers are not the principal element of the other equipment or systems, or (b) Computers essential for medical applications and incorporated in equipment or systems designed or modified for an identifiable and dedicated medical applications. 62. Electric equipment for time delay generation or time interval measurement: (a) Digital time delay generators with a resolution of 500 nanoseconds or less over time intervals of 1 microsecond or greater; (b) Multichannel (three or more) or modular time interval meters and chronometry equipment with time resolution less than 50 nanoseconds over time ranges greater than 1 microsecond.
11.5 ANALYSIS OF SAMPLES FROM INSPECTIONS OF IRAQ The IAEA Action Team succeeded in implementing, on very short notice, a comprehensive program of inspection activities by calling upon an impressive range of technical and administrative resources within the IAEA and its member states. The IAEA laboratories in Seibersdorf, Austria, played an important role in this effort by performing hundreds of analytical measurements on samples brought back by inspectors. These analytical results were used to plan subsequent inspections and to verify the declarations made by the Iraqi authorities about their activities prior to the war. Two specific laboratories in the IAEA' s Department of Research and Isotopes were responsible for supporting the Action Team. The Chemical Analysis and Isotopic Analysis Units of the Safeguards Analytical Laboratory (SAL) and the Chemistry Unit of the Physics, Chemistry, and Instrumentation (PCI) Laboratory performed the majority of the measurements. Additional analytical support was provided by government laboratories in the member states, the Austrian Research Center in Seibersdorf, the Atom Institute of the Austrian Universities in Vienna, and a commercial analytical laboratory. In the initial phases of the project, inspectors and analytical chemists were confronted with the need to define what, if any, undeclared nuclear activities had taken place; where and, if possible, when they took place; and by what means they were accomplished. To answer these questions, it was important to consider all aspects of the suspect process when the sample collection and analysis system was designed.
International Safeguards
599
Considering the degree of destruction of the Iraqi facilities and the Iraqis' efforts to conceal certain aspects of their program, it was necessary to collect a large number of environmental or construction material samples to search for unusual traces of nuclear materials. To facilitate this sampling, inspectors were supplied with materials such as filter papers for smear samples and plastic or glass bottles for bulk samples. At the laboratory, the samples were coded and split into three parts: one sample for assay in the IAEA laboratories, one for possible analysis in a laboratory outside the IAEA, and one for archival purposes. At the Agency's laboratories in Seibersdorf, Austria, scientists and technicians have focused their efforts on the measurement and analysis of hundreds of samples collected during inspections. They include smear samples taken from various sites in Iraq to detect potential undeclared nuclear activities; samples of uranium and plutonium compounds; samples of construction materials such as graphite, steel, and beryllium; and samples of soil, vegetation, water, rocks, and ores (see Tables 11.7 and 11.8). The results helped inspectors to map the Iraqi nuclear programme, both with respect to declared and undeclared activities. The techniques selected for analysis by the SAL (Rodgers, 1983) require high precision and accuracy, high selectivity for U or Pu and, in some cases, high sensitivity (because of the problems associated with shipping large samples). Table 11.9 lists the analytical techniques applied at the SAL for analysis of the samples obtained by the Action Team. Certain techniques such as high-resolution gamma ray spectrometry and X-ray fluorescence (XRF) spectrometry can be used to measure a large number of isotopes or elements. For this reason, these methods were used extensively in screening the non-nuclear material samples, whereas the more traditional safeguards analytical techniques were used for the nuclear material samples. The PCI Laboratory performs a broad range of measurements in support of IAEA programs. The activities of the PCI Laboratory range from the measurement of radionuclides in the environment (e.g., the international Chernobyl project) to the provision Table 11.7 Samples from Iraq processed at Seibersdorf Laboratories in 1991. Inspection
Non-nuclear materials
Nuclear materials
1st
48
31
2nd
35
0 51
3rd
139
4th
41
0
5th
49
61
6th
7
0
7th
139
141
8th Total
6
105
464
389
Chapter 11
600 Table 11.8 Type of samples and analysis performed in Agency's Laboratories in 1991 Sample category
Sample types
Analysis
Non-nuclear materials
Smears Vegetation Soil
Presence of U, Pu or radionuclides Amount of U, Pu
(Environmental)
Debris Rocks, Ores Water
Presence of F, C1 U, Pu isotopics Presence of high explosives
(Materials of construction)
Graphite Steels Beryllium Unknown metals
Purity, type or identity
Nuclear materials
Uranium metal Uranium compounds Plutonium compounds Polonium U, Pu waste & scrap
Amount of U, Pu U, Pu isotopics Amount of polonium Compounds of U, Pu Trace elements in U compounds
Table 11.9 Techniques applied at IAEA Safeguard Analytical Laboratory on Iraq samples Analytical method
Measurement- Safeguards
High-resolution gamma-ray spectrometry
Pu isotopic abundances Amount of 241Am, 237Np (Presence of radionuclides)
Alpha-particle spectrometry
238pu abundance (Presence of 21~
X-ray fluorescence spectrometry
Major, minor, trace elemental analysis
K-edge densitometry Hybrid XRF K-edge
Amount of U, Pu, Th, Np in solutions
McDonald/Savage potentiometric titration
Amount of Pu in pure nuclear materials
MBL modified Davies/Gray potentiometric titration
Amount of U in pure nuclear materials
Optical emission spectrometry
Trace elements in U compounds
Thermal ionisation mass spectrometry
U, Pu isotopic abundances
Isotope dilution mass spectrometry
Amount of U, Pu in small samples
International Safeguards
601
Table 11.10 Techniques applied in IAEA Physics-Chemistry-Instrumentation Laboratory
Non-Destructive Analysis Analytical Method Neutron activation analysis (NAA)
Measurement amount of F,C1,U and elemental composition
Gamma-ray spectrometry
Amount of U and radionuclides (fission prod.)
X-ray fluorescence spectrometry Conductivity and pH
Amount of U and elemental composition Ionic concentration of solutions
Destructive Analysis Laser-excited optical fluorimetry Inductively-coupled plasma atomic emission spectrometry Alpha-particle spectrometry
Amount of U Amount of U and trace elements Amount of U and Pu
of Quality control standards under the Analytical Quality Control Services program. Techniques for the screening and analysis of inspection samples from Iraq, which are divided between destructive and nondestructive analysis methods, are shown in Table 11.10. The analytical schemes applied to the non-nuclear material samples were specially developed to allow the inspectors to make rapid and selective measurements without demanding optimum performance in terms of precision. Another important objective was to obtain as much information as possible from a sample before it was destroyed by further chemical processing. Usually, a preliminary measurement was performed to screen samples for the presence of important components such as uranium, plutonium, or other radionuclides. Initial screening for uranium was performed by technicians at the Austrian Research Center in Seibersdorf. They used alpha particle counting in an ionisation chamber for the highest sensitivity. In both the SAL and the PCI laboratory, additional screening was carried out with high-resolution gamma ray spectrometry and energy-dispersive XRF spectrometry. Gamma ray spectrometry has a high sensitivity (nanogram to microgram levels) for radionuclides with relatively short half-lives, as is the case for many fission products and certain isotopes of Pu. The XRF method was used to screen for the presence of uranium, with a detection limit of 1 ~tg/cm2. Because uranium is a naturally occurring element that is present in soil at a concentration of 1 ~tg/g, there were certain analytical problems associated with the "blank" levels, primarily in the environmental samples. Following initial screening measurements, samples that showed elevated levels of U or Pu were measured by other nondestructive methods, such as NAA (for F and C1 content), before being submitted for chemical dissolution and further destructive analysis. Care was taken in the chemical treatment of the samples to minimise the danger of contamination. The primary methods used for the determination of U
Chapter 11
602
content~laser fluorometry and isotope dilution MS along with other complementary methods--were especially valuable in the analysis of uranium ores from A1 Qaim. The chemical treatment of samples containing U or Pu depended on the matrix. Filter paper smears were digested in nitric acid; soils, ores, and rocks were dissolved in nitric/hydrofluoric acid; and graphite or metal pieces with suspected surface contamination were leached with nitric or hydrochloric acid. Certain analytical techniques such as isotope dilution MS required that the samples undergo further chemical processing steps. Thus the analysis of one sample could involve a significant amount of effort and measurements with several analytical techniques. More than 500 environmental or smear samples and nearly 600 nuclear material samples from the 15 Action Team inspections in 1991 and 1992 were analysed at IAEA Laboratories in Seibersdorf. At Tuwaitha and Tarmiya the amount of 238Uwas successfully separated by isotope separators called calutrons. The Action Team was shown parts of the isotope separators that had been dismantled, destroyed, and buried in an attempt to conceal this program. A few parts from the ion sources and collectors of the calutrons were brought back for analysis. In addition, samples were taken from the declared product batches, coveting a range of 235Uenrichment from depleted <0.1 wt% to -6 wt% (see Table 11.11). The revelation that Iraq had been using the electromagnetic isotope separation (EMIS) process for enriching 238U came as a surprise to many people in the scientific community. It was not until the third inspection that the Iraqi authorities admitted to the existence of this program and described their activities in detail. A calutron is a massive electromagnetic machine developed in the early 1940s to enrich uranium. It is so old that most of its design plans have long been declassified. Not only is information freely available, but calutrons are also easier to build than equipment for other, more advanced enrichment technologies such as those based on
Table 11.11 Results of isotopic measurements for calutron places Sample I.D.
235Uabundance
Ion source 1
0.71
Natural U
Ion source 2
0.71
Natural U
Collector 1.1
0.76
Slightly enriched
Collector 1.2
5.82
Enriched
Collector 1.3
4.76
Enriched
Collector 1,4
0.39
Slightly depleted
Collector 1.5
6.84
Enriched
Collector 2.1
0.06
Highly depleted
Collector 2.2
5.94
Enriched
Collector 2.3
4.22
Enriched
Collector 2.4
0.79
Slightly enriched
(wt %)
Remarks
International Safeguards
603
Fig. 11.5. Inside the calutron. Schematic diagram of a beta calutron. Magnetic field is perpendicular to the page; power supplies, gas supply to low sources are not shown.
gaseous diffusion or gas centrifuges. The U.S. government never controlled export to specific calutron components, as it did for other enrichment technologies. At the same time, however, electromagnetic isotope separation technology presents several major challenges that are not easily overcome. An early problem with calutrons, for example, was that ion source components, such as iron, nickel, chromium, and copper could partially vaporise and create unwanted beams that literally cut the operating equipment to pieces. Ernest O. Lawrence invented the calutron electromagnetic isotope separator at the University of California in the early 1940s. It was based on the cyclotron; the name is a shortening of "California University cyclotron". The calutron separates the rare, fissile 235 isotope of uranium from the more plentiful, non-fissile uranium-238 by injecting an ionised beam of high-energy uranium atoms into a large magnetic field (see Fig. 11.5). Because the two isotopes have different masses, they follow slightly different trajectories, and a collector in the fight position will theoretically admit only uranium235. But the beams are imperfect, so some uranium-238 becomes mixed with the uranium-235. Little uranium-235 ends up in the uranium-238 collector, however, which means that the extremely "depleted" uranium from this collector can be a signature of calutron activity. U.N. inspectors have collected many samples of soil and material near the suspected Iraqi calutron sites, looking for this evidence, which would lend proof that Iraq has enriched uranium. A few parts from the ion sources and collectors of the calutrons were brought back to Laboratories in Seibersdorf for analysis. In addition, samples were taken from the declared product batches, coveting a range of uranium-235 enrichment from depleted uranium less than 0.1 wt% up to around 6 wt%.
604
Chapter 11
The ion source and collector pieces were from the large (1200 mm) calutrons which were declared to have been in operation at Tarmiya. These graphite pieces were scraped with a razor blade to remove about one gram of powder from the surface which was then leached in nitric acid to dissolve the uranium. This was followed by isotopic measurements using a mass spectrometer at Seibersdorf's Safeguards Analytical Laboratory. Results showed that the ion sources sampled for analysis contained only natural uranium (see table 11.11). The data from the sampled collector pieces show enrichments not exceeding about 6%. During World War II, the Manhattan Project spent the equivalent of about $5 billion in 1990 dollars to build a calutron production installation at the Y-12 plant at Oak Ridge, Tennessee. In 1944, the plant began producing the highly enriched uranium used in the bomb dropped on Hiroshima. At its peak in 1945, the program employed nearly 25,000 people and had over 1100 separating units in nine buildings. Eight electrical substations at Oak Ridge used more electricity than Canada produced during World War II. The plant had two types of calutrons. The larger "alpha" units enriched natural uranium to 10-30 percent uranium-235; the smaller "beta" calutrons enriched this product further, up to about 90 percent, for use in weapons. The calutrons were connected into "racetracks" of about 100 units to use the magnets and power supplies more efficiently. A small number of these calutrons were used after the war to purify stable isotopes for medical purposes and scientific research, but the technology was abandoned for making weapons material because it was extremely slow and costly and required enormous quantities of electrical energy. A calutron consists essentially of an intense source of uranium ions, a way to accelerate the ions to high energy within a vacuum system, and a way to collect the uranium-235 and uranium-238 ions after they have moved in separate arcs between the poles of a very large electromagnet. The components at the heart of the system are ion sources, collectors, and high-voltage, regulated direct-current power supplies (Lore, 1973; London, 1961; USDoE, 1980). The large vacuum chamber is situated between the pole faces of the electromagnet. "Forepumps" are used to begin pressure reduction; vacuum is maintained primarily by one or two high-capacity diffusion pumps with pipe-throat diameters of 15-20 inches. Special disposable stainless steel, water-cooled liners are often used in the vacuum chamber to simplify recovery of the large amounts of uranium that end up on the chamber surfaces. A calutron electromagnet has two circular poles, separated by a gap 30-60 cm wide in which the vacuum chamber is inserted. The magnet is typically about 1-2 m meters in diameter, weighs about 10-20 tons, and contains about a quarter-mile of thick copper wire. These extremely powerful magnets use one-third to one-half of the energy consumed by calutrons, and require cooling. The power supply for the magnets requires a direct-current capacity of about 1000 amps at 300--800 volts~similar to that used to power elevators. But the ones for calutron magnets must also be regulated precisely to produce a stable magnetic field.
International Safeguards
605
Table 11.12 Uranium determination in phosphate ores Sample
U content by laser fluorimetry (ppm)
U content by IDMS (ppm)
U content by gamma-ray spectrometry (ppm)
Ore-1
63.8
56.7
61.5
Ore-2
73.0
72.8
69.0
Ore-3
84.5
-
87.7
Ore-4
160.0
-
175.0
The associated chemical processing area is usually one of the largest parts of the plant. In this area the collectors are burned and the uranium is recovered from the ashes. In addition, the waste uranium must be recovered from the other calutron components, by scrubbing with nitric acid, soaking in acid baths, or burning or dissolving away disposable components. The power supplies, magnets, and other components require extensive cooling with oil or purified water. A large fraction of the uranium starting materials for the EMIS programme came from domestic uranium mines at A1 Qaim. The amount of uranium obtained can be estimated from the weight of ore processed and the concentration of uranium in the ores. This information would represent an upper limit on the amount of indigeneously produced uranium starting materials which were available for the EMIS or other processing in Iraq. Several ore samples were brought back to Laboratories in Seibersdorf by the inspection team; the uranium content was determined by gamma-ray spectrometry and, after dissolution, by laser-excited optical fluorescence at PCI and by isotope dilution mass spectrometry at SAL. Considering the low concentration of uranium in the samples, the agreement between the various techniques is quite good (see Table 11.12). 11.6 S T R E N G T H E N I N G THE SAFEGUARDS SYSTEM When Israel attacked Iraq's Osirak reactor at Tuwaitha in 1981, the State Department denounced the raid. So did the International Atomic Energy Agency (IAEA), which passed a resolution condemning the action. The Israeli claim that Iraq planned to use the facility to produce nuclear weapons material was dismissed~the facility was under IAEA safeguards (international inspection), and Iraq was fulfilling its commitments to the Non-Proliferation Treaty (NPT) and its safeguards agreement with the IAEA. Iraq could have put into place every element of a nuclear weapons program, including production facilities, without violating a single provision of the NPT or of its safeguards arrangements with the IAEA. When Saddam Hussein was ready to produce the weapons, he could have invoked the NPT clause that allows a signatory to withdraw from the treaty after giving 90 days notice by declaring that "extraordinary events, related to the subject matter of this Treaty, have jeopardised [its] supreme interests."
606
Chapter 11
What happened in Iraq is not because the inspectors have been lax. They employed an impressive array of mechanisms to make sure that materials used to generate nuclear power or for other peaceful purposes are not diverted to bomb development. In 21 years, the inspectors, who have lately run more than 2000 inspections a year, have never found even a single case of material diverted from peaceful use. But the inspectors can visit only those facilities they know about. The way to start a successful bomb-building program is simply to carry it on at highly secret sites completely separate from all publicly known power-generation or research activities. The IAEA did not have a theoretical fight to conduct "special inspections" of undeclared plants but only if another member country supplies intelligence information indicating that such a nuclear facility exists--and until the gulf-war cease-fire that had never happened. A move has been made to amend the treaty to make such "challenge" inspections mandatory and not voluntary. At its March 1995 meetings, the IAEA Board of Governors reiterated that under comprehensive safeguards agreements verification by the Agency should be so designed as to cover the correctness and completeness of States' declarations, so that there is credible assurance of the non-diversion of nuclear material from declared activities and of the absence of any undeclared nuclear activities. A proliferation critical path has been developed by IAEA experts as a means to structure both the information requirements and the analysis requirements. The proliferation critical path is designed to include all known pathways for the production of weapons-usable material and subsequent weaponisation. It can be represented graphically as a series of increasingly specific and detailed diagrams indicating the processes required to develop a nuclear weapon. The first two levels are illustrated in Figs. 11.6 and 11.7. The top-level diagram (Fig. 11.6) contains all the main steps involved in proliferation. Each block in the top-level diagram is broken down into more specific routes or processes. For example, the enrichment block (Fig. 11.7) is broken down into nine possible processes (gas centrifuge, electromagnetic, aerodynamic, gaseous diffusion, molecular laser, atomic vapour laser, plasma separation, chemical exchange and ion exchange) as illustrated for the second level of the proliferation critical path. The third level of proliferation critical path, for example of gaseous diffusion enrichment is composed of the following: Equipment indicators 9 gaseous diffusion barriers 9 diffusion housings 9 gas blowers 9 etc. Material indicators 9 uranium nexafluoride 9 chlorine trifluoride 9 etc.
International Safeguards
607 :
Mining & Milling
Conversion
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Enrichment
. . . . . . . . . . . . . .
Heavy Water Production
Fuel Fabrication
HEU Processing
Reactor
Weaponization
Reprocessing ~
Weapon
Pu processing
Fig. 11.6. Proliferation critical path (top level).
Environmental monitoring indicators 9 9 9 ~
hydrogen fluoride or fluorinated compounds derived from uranium hexafluoride evidence of perturbed uranium isotopics large heat increases in air or water etc.
Other indicators 9 large amounts or power going into a large facility 9 etc. Each process is then characterised by indicators of the existence or development of the process, such as specialised equipment, dual-use equipment, nuclear and non-nuclear
608
Chapter 11
Gas Centrifuge Enrichment
Gaseous Diffusion Enrichment
Electromagnetic Enrichment
Atomic Vapor Laser Enrichment
Molecular Laser Enrichment
Chemical Exchange Enrichment
Aerodynamic Enrichment
Ion Exchange Enrichment
Plasma Separation Enrichment Fig. 11.7. Proliferation critical path (second level detail of enrichment).
materials, training and environmental signatures. These indicators represent the third level of the proliferation critical path. The weaponisation-related blocks in the top-level diagram involve such processes as the production of high explosives, tritium, enriched lithium and alpha-emitting radionuclides. The indicators will be largely comprised of equipment, material and environmental signatures. The proliferation critical path structure serves a number of purposes. The model expanded declaration will evolve in a way that reflects this same structure. Thus the nuclear programme of each country, as described at the activity level in the expanded declaration, will be structured and reflected to the same level of detail as illustrated in Fig. 11.8. The proliferation critical path structure then becomes a framework whereby
609
International Safeguards
Proliferation Critical Path (rule-based system)
NUCLEAR ACTIVITY PROFILES Fig. 11.8. Integrated Rule-Based System. the analyst can "visualise" the whole of a State' s nuclear programme in a coherent and connected way. The analyst can then explore individual activities at increasing levels of detail depending upon the information provided by the State and information deriving from the inspection process and other sources. The expanded declaration can also be "visualised" geographically where each activity is further identified by its location. Here again the information is nested so that the analyst can follow an activity from its location in a country to a specific point in a facility (facility line drawing) and then to a specific point in a building (design information). The system also has an inspection file capability. This allows an inspector, during the inspection planning process, to create a file of observables (equipment, nuclear and non-nuclear materials, infrastructure, etc.)--some from information provided by the State, some from other sources~to prepare the inspectors for what they should see in the course of a particular inspection. With the help of experts, the proliferation critical path is being formulated as logically connected "if-then" rules. The primary purpose of this formulation is to recognise and place information (e.g., export data) in the appropriate place(s) in the critical path structure. The critical path takes into account the possibility of shortening any of the paths to weaponisation at each of the fuel-cycle steps through external procurement (e.g., procurement of source material, UF 6, enriched uranium, etc.) and technical assistance. Nuclear activity profile for a given country is then obtained from a model presented in Fig. 11.8. A summary of measures required for an effective safeguards system is presented in Table 11.13.
610
Chapter 11
Table 11.13 A summary of measures required for effective safeguards Broad Access to Information
Expanded Declaration
1. Information on the SSAC (State System of Accountancy and Control (of nuclear material)) 2a. Information on past nuclear activities (to the extent necessary to enable the Agency to verify the completeness and correctness of the State' s declarations) through access to existing records on production of nuclear material and on related facilities 2b. Information presently provided routinely: (i) design information and modifications thereto, including closed-down and decommissioned facilities; (ii) accounting and operating records; (iii) accounting and special reports; (iv) operational programme 2c(i). Description of the nuclear fuel cycle, and of other activities involving nuclear material 2c(ii). Description of nuclear fuel cycle-related R&D (hereinafter referred to as nuclear R&D) activities 2c(iii). Information, to be agreed with the State, on operational activities additional to that required under INFCIRC/153 (see 2b(iv) above) 2c(iv). Nature of each of the buildings on the sites on which are located nuclear facilities, LOFs or nuclear R&D activities 2c(v). Nature of any other locations directly related to the operation of nuclear facilities, LOFs or nuclear R&D activities 2c(vi). Location and status of known U and thorium ore deposits and mines 2c(vii). Domestic manufacturers of major items of nuclear equipment or materials 2c(viii). Information identified in GOV/2629 (voluntary reporting on nuclear material and specified equipment and non-nuclear material 3a. Early provision of design information 3b. Plans for the further development of the nuclear fuel cycle 3c. Description of planned nuclear R&D activities
Environmental Sampling
For ad hoc inspections, at locations where the initial report, or inspections carried out in connection with it, indicates that nuclear material is present For routine inspections at strategic points
International Safeguards
611
For special inspections, at the locations where these take place For design information verification, at any location to which the Agency has access to carry out design information verification
Increased Physical Access
Improved Analysis of Information
Improvements in the Agency's information analysis methods
Broad Access
Access to locations beyond strategic points in nuclear facilities of LOFs, but within the sites containing such facilities or LOFs Access to other locations identified in the Expanded Declaration Access to other locations which may be of interest to the Agency, under voluntary arrangements with the State
No-notice Inspections
Unannounced (no-notice) routine inspections Unannounced (no-notice inspections at other locations identified in the Expanded Declaration
Optimal Use of the Present System
SG Technology Advances
Use of unattended equipment Remote transmission of inspection data Remote monitoring of safeguards equipment
Increased Co-operation The SSAC carries out activities that enable the Agency with States and SSACs to conduct inspection activities The Agency and the SSAC may carry out selected inspection activities jointly The Agency and the SSAC carry out selected support activities jointly Use of simplified procedure for designation of inspectors Multiple-entry visa, long-term visa or visaless entry for inspectors on inspection Use of available systems for direct communication (including satellite systems) between inspectors and installations in the field and Headquarters SG Implementation Parameters
Significant quantities of nuclear material Conversion/detection times Starting point of safeguards
612
Chapter 11
REFERENCES Agu, B. and Iwamoto, H., Recent advances in safeguards operations, Nuclear Safeguards Technology 1982 (Proc. Symp. Vienna, 1982), Vol. 1, IAEA, Vienna (1983) 15. Agu, B., McManus, J. and Schuricht, V., Recent advances in safeguards operations, Nuclear Safeguards Technology 1986 (Proc. Symp. Vienna, 1986), Vol. 1, IAEA, Vienna (1987) 57. Deron, S., Donohue, D., Bagliano, G., Kuhn, E. and Sirisena, K., The IAEA's analytical capabilities for safeguards IAEA-SM-333/221, p. 717, Vienna, 1994. Donohue, D.L. and Zeisler, R., Analytical chemistry in the aftermath of the Gulf war. Anal. Chem., 65 (1993) 359A. Donohue, D.L. and Zeisler, R., Behind the scenes: Scientific analysis of samples from nuclear inspections in Iraq. IAEA Bulletin 1 (1992) 25. Elbaradei, M., 1995 non-proliferation treaty review and extension conference, IAEA-SM-333/205, p. 13, Vienna, 1994. Handbook of Nuclear Safeguards Measurement Methods; Rodgers, D.R., Ed.; U.S. Nuclear Regulatory Commission: Washington, DC, 1983; NUREG/CR-2078. Harms, N. and Rodriguez, P., Safeguards at light-water reactors: current practices, future directions. IAEA Bulletin 4 (1996) 16. Hayashi, M., Asai, S., Motoda and Y., Kikuchi, M., Present status of safeguards implementation in Japan, IAEA-SM-333/39, p. 59, Vienna (1994). London, H., Separation of Isotopes. George Newnes Ltd., London, 1961. Love, L.O., Electromagnetic separation of isotopes at Oak Ridge. Science (Oct. 26, 1973), 343-352. Nilsson, A., Safeguards at LEU facilities: current practices, future directions. IAEA Bulletin 4 (1996) 11. Scheinman, L., The non-proliferation treaty: on the road to 1995. IAEA Bulletin, 1 (1992) 33. Schreifer, D., Perricos, D. and Thorstensen, S., IAEA Safeguards experience, IAEA-SM-222/217, p. 35, Vienna, 1994. Schriefer, D., New safeguards measures: initial implementation and experience. IAEA Bulletin, 4 (1996) 7. U.S. Department of Energy, Nuclear Proliferation and Civilian Nuclear Power, Vol. II: Proliferation Resistance (1980); Zifferero, M., IAEA activities and experience in Iraq under the relevant resolutions of the United Nations Security Cancil, IAEA-SM-330/220, p. 211, Vienna, 1994. Zuccaro-Labellarte, G. and Fagerholm, R., Safeguards at research reactors: current practices, future directions. IAEA Bulletin, 4 (1996) 20.
613
CHAPTER 12
Environmental Monitoring for Safeguards
The objective of an environmental monitoring and sampling programme is to detect undeclared nuclear activities. Although the ability to detect any type of nuclear activity would be important, the first priority is to detect the production of separated plutonium or HEU, i.e. material that is in, or can be processed relatively easily into, a form in which it can be incorporated directly into a nuclear weapon. Thus the ability to detect undeclared reprocessing, or conversion of the plutonium nitrate product to Pu metal, enrichment of uranium to HEU or conversion of HEU product to HEU metal is the key requirement. There is also an issue as to whether or not to introduce the concept of goal quantities. Whilst it is clear that the detection of, say, the separation of a few grammes of plutonium would be very difficult, and the possession of such would not allow the construction of a nuclear weapon, the fact remains that any undeclared peaceful nuclear activity would constitute a breach of a Comprehensive Safeguards Agreement. The aim therefore is expressed in terms of detecting the undeclared activity, rather than any particular amount of nuclear material. From the IAEA safeguards point of view, there are two options for a proliferation in carrying out undeclared activities: (a) In a declared nuclear site that is subject to safeguards inspections; or (b) elsewhere in the State. Environmental monitoring and sampling (EMS) can be used, in principle, to detect both categories, via short- or long-range monitoring. The long-range EMS, may give rise to suspicion of a particular site or area outside of a declared nuclear site. These considerations can be combined to give three possibilities for EMS for safeguards: (i) Short-range--at a declared nuclear facility; (ii) Short-range~at a specific site or area not declared as having a nuclear activity; (iii) Long-range--no specific target. Before looking separately at the detection of plutonium and HEU production, there are some general considerations which apply to both. Some of the longer lived isotopes of interest may be detected in the environment even in the absence of undeclared activities. Weapons-testing fallout and (declared) nuclear industrial activities can both
614
Chapter 12
contribute to giving a measurable concentration. It would be important, therefore, to have a thorough understanding of these levels, and the variation in them, in a particular State before implementing an EMS programme. It can be expected that a State (or a well organized group) would take countermeasures to avoid detection. These counter-measures could take several forms, e.g.: 9 underground location; 9 strict secrecy; 9 trapping effluents and storing them on site; 9 masking, i.e. deliberately releasing large amounts of certain isotopes to obscure others; co-location with a similar, declared, activity; 9 in the case of reprocessing, cooling time to allow some potential fission product signatures to decay. The detection of an undeclared activity would not necessarily allow a deduction as to how long that facility had been operating and therefore how much nuclear material had been produced. Therefore in this chapter we shall consider the environmental signatures likely to arise as a result of nuclear fuel reprocessing on a relatively small scale as part of a covert nuclear weapons programme. Nuclear weapons can be constructed from highly enriched uranium (HEU) produced by enrichment of natural uranium or from 239pu which is produced in nuclear reactors by neutron transmutation of 238U. The approximate quantities of HEU and Pu required to manufacture a nuclear explosive device are quoted in the IAEA Safeguards Glossary (1987), and are referred to as "Threshold Amounts". The Threshold Amounts (TA) are as follows: 9 HEU (235U > 90-95 wt.%) 25 kg 9 Pu (239pu > 95 wt.%) 8 kg The technology required to reprocess nuclear fuel and extract plutonium is much simpler than that required to enrich uranium; however the use of HEU as a weapons material does not necessarily require the use of a reactor. HEU can be obtained by reprocessing MTR fuel which is available in many countries. The use of a nuclear reactor, the generation of a range of highly radioactive fission products and the use of a number of organic chemicals is likely to make the detection of a 239pu-based weapons programme simpler than detection of a 235U-based programme. 9
12.1 S I G N A T U R E S OF U N D E C L A R E D A C T I V I T I E S There are many different processes that a potential proliferant could pursue to obtain weapons grade nuclear materials. There are many places within the cycle that effluents may be released into the environment and/or be present at trace levels within facilities to be inspected. Table 12.1 attempts to suggest several compounds which should be analysed as part of an upgraded safeguards program. This list is only an example of what are the key species of interest and does not represent an exhaustive and refined tabulation. In considering applications of environmental sampling, priority should be given to those facilities that are suspected of producing weapons-grade material.
Environmental Monitoringfor Safeguards
615
Table 12.1 Suggested chemical species for environmental sampling and analysis of potential undeclared nuclear facilities (after Raber, 1993) Uranium enrichment/gaseous diffusion
UF 6 (UO2F2 after hydrolysis), F 2, HF, U-isotopics, (235, 238) PTFE, freon
Jet nozzle enrichment
UF6, H 2, HF, U-isotopics (235,238)
Calutron
UC14, hexachloropropene, U-isotopics (235,238), COC12
Gas centrifuge
UF 6, F 2, U-isotopics (235,238), marging steel 131Xe, 85Kr, 129/131I,Cs-isotopics, 141/144Ce,89/9~ tributylphosphate, kerosene, freon, HF, NaOH, HNO 3, NO/NO 2, 12 (associated iodides), HT, HTO, C-14 compounds, U- and Pu-isotopics, 237Np, 241Am, H6TeO 6, H6MoO 6
Reprocessing solvent process
Mining
Rn, U-oxides (as dust only)
Milling
Rn, NH 3, SO 2, H2S, U-oxides (UO 3, UeO), NHn)2U207, Th-oxides
Refining/con version
U-oxides, CaF 2, tributylphosphate, NH 3, NO/NO 2
High explosive fabrication and testing
HMX, RDX, TNT, TATB
Plutonium and HEU metal production
Ca compounds, low burn-up Pu, 90% or greater 235U
Others
6Li, 2~
21~
3H, specialized polymer resins, graphite, Be
12.1.1 Signatures of an undeclared reactor
Any change in 234U or 235U abundance is unequivocal evidence that man-altered uranium is present. The ratio of 234U to 235U can provide some information about the type of enrichment technology used. The presence of 236U would indicate the presence of uranium that had been exposed to a neutron flux. Detection of an undeclared reactor by environmental sampling is possible by sampling short-lived atmospheric radioactive gases and transported radionuclides in surface water (streams, rivers, lakes, oceans). Effectiveness of detection techniques decrease with distance from the reactor. On-site and short-range sampling of short-lived fusion gases is key to the detection of an undeclared reactor, however, the reactor must be operating. Aqueous effluents containing trace activation and fission products are also excellent sources for effective on-site and short-range detection. Aqueous effluents also have the advantage of long-range detection because of their traceable and columnated transport. Furthermore the reactor does not have to be operating and many effluent signatures will remain in the stream sediments for several years gradually migrating downstream where they concentrate in deposited sediments and aquatic biota. Tables 12.2 and 12.3 outline a variety of signatures and the associated sampling and detection technologies, and their degree of effectiveness.
616
Chapter 12
Table 12.2 Detection of an Undeclared Reactor Signatures
Sampling techniques
Analytical techniques
Effectiveness
mobile lab 9 7-spec 9 3H liq. scint 9 noble gas y-det samples screen sample coll.: 9 wipes 9 effluents 9 protect clothes
radiometry
high
7-spec chem., GC, lab. anal.
mod high
mobile lab 9 y-spec 9 3H liq. scint 9 noble gas 7-det
radiometry
high
sample coll.: 9 stream water 9 air filters
lab. anal.
high
sample coll. 9 water
lab. anal.
high
mobile lab 9 7-spec 9 3H liq. scint 9 noble gas 7-det sample screen sample coll., wipes effluents protect clothes
radiometry
mod
3H sampler sample coll: 9 water 9 air filters
lab. anal.
mod
sample coil: 9 water
lab. anal.
mod
Without counter measures
On site
short-lived fission gases* 3H, 14C volatile isotopes AQ short-lived ACT, FP long-lived ACT, FP AQ U, Pu isotopics Thermal sig.
10 km
short-lived fission gases* 3H, 14C volatile isotopes AQ short-lived ACT, FP AQ long-lived ACT, FP AQ U, Pu isotopics Thermal sig.
Long Range
3H AQ short-lived ACT, FP AQ long-lived ACT, FP AQ U, Pu isotopics Thermal sig.
With Counter Measures
On Site
short-lived fission gases* 3H, 14C volatile isotopes AQ short-lived ACT, FP AQ U, Pu isotopics Thermal sig.
10 km
3H,
14C
AQ long-lived ACT, FP AQ U, Pu isotopics Thermal. Long Range
AQ long-lived ACT, FP AQ U, Pu isotopics Thermal.
*Reactor operating
Environmental Monitoringfor Safeguards
617
T a b l e 12.3 Reactor signatures Reactor type
G (LWC)
Magnox
HW (LWC)
BWR
PWR
CANDU
R e l e a s e to air d u r i n g n o r m a l r e a c t o r o p e r a t i o n . 41Ar
x
XX
x
x
x
x
88Kr
x
x
x
x
x
x
85mKr
x
x
x
x
x
x
87Kr
x
x
x
x
x
x
133Xe
x
x
x
x
x
x
135Xe
x
x
x
x
x
x
14C
XX
XX
x
x
x
x
3H ( H T O )
x
x
XX
XX
x
XX
R e l e a s e to air due to fuel f a i l u r e 1321 XX XX
XX
x
x
x
133I
XX
XX
XX
x
x
x
131I
XX
XX
XX
x
x
x
129I
x
x
x
x
x
x
85Kr
x
x
x
x
x
x
l~
x
x
x
x
x
x
l~
x
x
x
x
x
x
x
XX
x
x
X
X
X
X
R e l e a s e to w a t e r 3H
x
Short-lived (days) activation products 24Na x 56Mn
x
64Cu
x
76As
x
51Cr
x
72Ge
x
56Fe 32p
x x
239Np
x
L o n g - l i v e d a c t i v a t i o n p r o d u c t s (30 days) 6~
x
X
X
X
58C0
x
X
X
X
57C0
x
X
X
X
65Zn
x
x
54Mn 465c
x x
x
63Ni
x
55Fe
x
56Fe
x
618
Chapter 12
Reactor fuel for the production of plutonium and target manufacture for the production of tritium involves the use of material and machinery common to other manufacturing industries with the exception of uranium and possibly enriched lithium. Aqueous waste from such operations may find its way to surface water systems and is the primary source for short- and long-range detection.
12.1.2 Undeclared HEU production HEU processes can be identified through the use of routine environmental monitoring. Regardless of the chemical form in the environment, the detection of altered isotopic ratios in uranium is an unambiguous indicator that enrichment activities have occurred. Several technologies have been used for production of highly enriched uranium. A first priority should be detection of HEU production by the technologies with proven capability of producing HEU in quantities sufficient for the assembly of nuclear weapons. These technologies are: 9 gaseous diffusion, 9 gas centrifuge, 9 vortex tube, and 9 electromagnetic isotope separation. Uranium mining operations could be carried out for production of power reactor fuel and declarations of uranium mining activities are not required for full scope safeguard states. Therefore detection of mining operations by environmental monitoring is not a safeguards objective. Detection of conversion of uranium oxide to UF 6 is of interest as it indicates preparation for uranium enrichment activities. Other separation technologies may use different feed materials and different signatures must be identified for each separation technology to be identified. Uranium with isotopic abundances different from that of natural uranium is the primary signature for HEU production activities. In any separation technology some enriched uranium will inevitably be released to the environment. Environmental samples taken at or near an enrichment facility can contain some of the enriched material altering the uranium isotopic abundance. Analysis of samples of vegetation, water and soil for uranium isotopic content using a sensitive analytical technique, such as thermal ionization mass spectrometry is recommended as the primary technique for the detection of HEU production. Table 12.4 lists the signatures for processes using UF 6 gas. In the conversion step, one could look for uranyl fluoride, but the isotopic ratio would not change from uranium ore (i.e., no fractionation). However, an analytical method that yielded oxidation states of atomic uranium and fluoride, such as ESCA (electron spectrometry for chemical analysis), could be used to identify the uranyl compound. Isotopic analysis of uranium is the best chance of detecting uranium enrichment. The sensitivity of the method is sufficient to detect the operation if samples can be taken in close proximity to the facility.
Environmental Monitoring for Safeguards
619
Table 12.4 Signatures from enrichment processes using UF 6 gas Process
Signature
Sample matrix
Measurement technique
Conversion
UO2F 2
Air, vegetation, soil
ESCA
Enrichment
HF
Air, capture gas in impinger
IC, SIE
F-
Biota, water
IC, SIE
Uranium conc.
Vegetation, water
TIMS
Uranium isotope ratios
Vegetation, water, soil, sediments, swipes
TIMS
UOzF 2
Swipes, vegetation
ESCA
ESCA = electron spectrometry for chemical analysis; IC = ion chromatography; SIE = selective ion electrode; TIMS = thermal ionisation mass spectrometry.
Table 12.5 shows expected isotopic uranium contents which might be expected as product from cascade type (gaseous diffusion or gas centrifuge) enrichment facilities. While the highly enriched U is actually the target of these efforts one should also be able to verify the production of low enriched uranium. Columns 5 and 6 of Table 12.5 show the isotopic enrichments which would result from a 1000-1 mixture of natural U and low or high enriched U. Table 12.6 shows the ratios of the isotopes 234U and 238U relative to the 235Uin each mixture. The 0.6% isotopic shift in the 238/235Uratio in the case of LEU-MIX should be detectable by today's technology. The 0.13% in 234/235Uratio would be at the edge of measurement accuracy for nearly all laboratories. In this case Table 12.5 Uranium isotopic abundances in cascade type enrichment facility atom fractions of source materials U isotope
Nat U
LEU
HEU
MIX 1 (LEU)
MIX 2 (HEU)
234
0.000054
0.000300
0.008000
0.00005424
0.0000620
235
0.007200
0.050000
0.900000
0.007243
0.008092
238
0.992746
0.949700
0.092000
0.992703
0.991846
Table 12.6 Uranium isotopic ratios 4/5
% Shift
8/5
Nat U
0.00750
LEU MIX
0.00749
-0.13
137.06
HEU MIX
0.00765
2.0
122.57
Required precision/accuracy (rsd%)
% Shift
137.88
0.04 %
0.60 12.6 0.15 %
620
Chapter 12
detection limits are driven by accuracy limitations rather than instrument sensitivity. It is however true an that ultimate sensitivity requirement for 2x 107 atoms of U is still needed for blanking or reagents and some smear analyses. 12.1.3 Signatures from reprocessing of Pu
Sufficient 239pu(TA = 8 kg) could be obtained from 5-15 t of low bum-up natural U fuel. The problem of confirming the presence of a covert nuclear fuel reprocessing facility, within any country can be subdivided. The following scenarios can be considered: 1. A specific site is suspected as the location of a covert reprocessing facility which is operational or has recently been used. 2. It is suspected that a reprocessing facility is in a specific area, a few tens of miles in diameter. 3. It is suspected that a reprocessing facility exists within a country but its location is unknown. Clearly, these represent very different problems. Further information may also be required. For example, if an environmental signature is detected, the following questions might arise: 1. Is the signature due to fuel reprocessing activities? 2. What is the scale of that activity? 3. When was the fuel reprocessed? 4. What type of fuel was reprocessed (and what was its level of burn-up)? The reprocessing term is process-dependent and will therefore encompass a wide range of active and inactive components. The environmental signatures will depend on the release fraction, plume transport and dispersion and, in the case of organic materials, their reactivity in the environment. Environmental signatures may consist of the radioactive and non-radioactive species listed below: Radioactive:
9 9 9 9
gaseous and volatile products: 85Kr, 133Xe, 3H, 14C, l~176 fuel-derived particulates: 95Zr, 95Nb, 134/133/137Cs,144Ce, 99Tc; two years' cooling will effectively remove 133Xeand 1311by decay, and conversion derived particulates: characteristic Pu isotopic, 24~Am and 99Tc.
Non-Radioactive:
9 9 9 9
NO x from aqueous dissolution; secondary(tertiary) amines with C chains C8-C~2; organo-phosphorus compounds (phosphates, phosphonates) with C chains C4-C6; diluents, such as C12 Alkanes, aromatic hydrocarbons, brominated/chlorinated hydrocarbons; 9 reducing agents such as sulphonate, hydroxylamine, and 9 HF, H20 2, (C204)2-, 12 from conversion. The detectability of a signature will depend on the original inventory of that material, its release factor, its mode of dispersion in the environment, local background levels
621
Environmental Monitoringfor Safeguards
and upon the analytical instrument used to detect it. Limits can be placed on some of these variables but there is a large number of unknown quantities. It is likely, for example, that 100% of the inert gas fission products will be released, however other fission products, activation products, actinides and other processing chemicals can be contained more effectively. Some material will inevitably escape but the fraction will be critically dependent on the measures taken to limit discharges. Measures will undoubtedly be taken to limit discharges, primarily to limit the radiation dose to the plant operators, possibly to limit environmental impact but most obviously to avoid detection. Recovery of Pu from irradiated fuel involves three main stages: 9 fuel dismantling and dissolution; 9 separation and purification, and 9 conversion (from nitrate to metal). The first stages where a signification release of radioactive isotopes could be expected are those of fuel dismantling and, especially, dissolution. Fuel dissolution usually takes place in nitric acid and results in the complete release (i.e. from the fuel and dissolver solution) of most of the volatile isotopes~e.g., 129I, 131I, laC, 85Kr, 133Xe and partial release of 3H and l~176 Non-volatiles may also be released to a much lesser degree as an aerosol of fuel solution droplets. Important isotopes include 95Zr, 95Nb, 134/137Cs,144Ce and 239/24~ Non-active NO~ gases may also be released. These releases are potentially able to escape into the environment unless trapping measures are used. Non-aqueous dissolutions are also possible using a molten chloride system as part of a pyrochemical separation process. In a high containment plant, the release of fuel solution would probably be below current detection limits. However, if minimal off-gas treatment is applied (i.e. caustic scrubbing and filtration is omitted) then there would be potential for the release of fine droplets of fuel solution. The release fractions of volatile radionuclides will depend on the type of off-gas treatment process used. An indication of the activity release for a 50 kg dissolution of 1000 MWd/t fuel, involving the use of different off-gas treatment system (consider, scrubber), is presented in Table 12.7 (after McMahon et al., 1993). Table 12.7 Estimated release fractions from the dissolution of irradiated fuel (after McMahon et al., 1993) Radionuclide released 129I 131I lac 85Kr 133Xe 3H 103nO6Ru
Release fraction for a simple off-gas system 1 1 1 1 1 10-5 10-4
Estimated range at releases from a 50 kg dissolution (Bq) of 1000 MWd/t fuel 1.35•215 4.45•215176 7.20• 106-7.20x 103 6.00x 1011 1.05x1014 2.55x105-2.55x102 1.55x 109-1.55x106
622
Chapter 12
A wide variety of processes are possible for separating plutonium from an irradiated fuel solution. A proliferator would not be constrained to the most common commercial method using solvent extraction by tributyl phosphate in odourless kerosene. These other methods could include ion-exchange, pyrochemical methods and precipitation. There would, therefore, be a wide range of potential signatures arising from the chemicals used in these methods (e.g. secondary and tertiary amines, organophosphorus compounds, alkanes, aromatic hydrocarbons). It seems possible that less stringent precautions would be taken in the handling and preparation of these chemicals prior to contact with the irradiated fuel solution; so greater releases could be expected at that point. The product and waste streams after separation would of course contain already mentioned isotopes, together with the chemicals used in the processes and any radiolysis products of those chemicals. The different methods would also produce different release patterns. Pyrochemical processes are likely to generate considerable quantities of particulate, but because of the formation of stable salts in these conditions, releases of l~176 are likely to be much lower than in aqueous dissolution systems. The plutonium nitrate product of reprocessing needs to be turned into plutonium metal before it can be used in a weapon. This process may involve a precipitation step, using H20 2, HI and H2C204, dehydration and calcination prior to high temperature fluorination with HF, and is likely to give rise to particulate formation. Direct fluorination with HF is also possible and may also give rise to environmental signatures from the high temperature dehydration process prior to calcium reduction. The alternative calcium hexaplutonate process does not require high temperature calcination or dehydration and may, therefore, result in a reduced release to the environment. Calcium reduction of PuF 3 and PuF 4 to Pu metal is carried out in a closed bomb and, therefore, will result in minimal releases. US experience at the Hanford Site has shown that krypton and xenon are present from reprocessing as a result of activities associated with leaching of the chopped fuel materials. Furthermore, it has been observed that 129I emissions are always present, although at fluctuating concentrations, during and after reprocessing. At US facilities efforts were made to effectively reduce emissions but detection was possible kilometres away from the operating facilities. Therefore, it appears that continuous or periodic air sampling and/or monitoring would be effective for facility monitoring.
12.1.4 Discharges from Sellafield reprocessing plant In order to be able to estimate the releases resulting from undeclared activities one should consider releases from the declared sites. As an example we consider here Sellafield reprocessing plant in the UK. British Nuclear Fuels' Sellafield reprocessing plant publishes annual figures of measured discharges and local environmental measurements, in order to conform with environmental legislation (British Nuclear Fuels, 1992). The local environmental burden of anthropogenic radionuclides will be due to discharges from nuclear activities
Environmental Monitoringfor Safeguards
623
Table 12.8 Major source terms from the Sellafield plant (Reprocessing activity, 1991)
Output (TBq) With liquid effluent
3H 14C 35S 9~ 95Zr 95Nb 99Tc l~ l l~ 125Sb 129I 134Cs 137Cs 14ace 238'239'24~ Pu(t~) 241pu 241Am As Gases 3H 14C 85Kr 1291 1311
1803 2.4 1.0 4.1 7.4 5.0 3.9 18.7 0.3 11.6 0.2 0.8 15.6 1.7 1.1 29.5 0.7
619,000 5,550 44,600,000 12 1.3
With particulates
9~ l~ 137Cs 238'239'24~ Pu(o0 241pu 241Am]242Cm
0.4 1.6 1.4 0.2 1.1 0.1
at the site which date from the 1950s. Current annual discharges are much lower than those reported in the 1970s. The environmental fingerprint is therefore rather complex. The plant reprocesses between 1200 and 1600 t fuel per year and the published discharges for 1991 are shown in Table 12.8 and in Figs. 12.1 and 12.2. The discharges are mainly due to the reprocessing of magnox-type fuel with an average burn-up of
624
Chapter 12
Fig. 12.1. Radionuclides released from the Sellafield site as liquid discharges, in 1991. Figures are in TBq and are plotted on a logarithmic scale.
5000 M W d and a 200 day cooling time. It should be emphasized that these are approximate figures, that other types of fuel are reprocessed and that discharges do occur from other nuclear facilities at the site. There are also storage facilities on site which give rise to discharges which relate to fuel reprocessed in previous years. If the activity of each isotope discharged within a given period is divided by the activity in the reprocessed fuel, a factor is obtained which crudely represents the relative ease with which a species is discharged. This release factor will be specific to the Sellafield site, which passes the effluent through a variety of decontamination procedures prior to discharge including storage whilst short-lived radionuclides decay.
Environmental Monitoringfor Safeguards
625
Fig. 12.2. Radionuclides released from the Sellafield site as aerial discharges, in 1991. Figures are in TBq and are plotted on a logarithmic scale.
Nonetheless, beating this in mind, this release factor will give a rough indication which species are most likely to find their way into the environment. Figures 12.3 and 12.4 show released fractions for a number of radionuclides in both liquid effluents and in aerial discharges from the Sellafield reprocessing facility.
12.2 P A T H W A Y S OF THE R E L E A S E S
Those species which are not removed by trapping measures, or are accidentally discharged, are released into the environment. The release may be airborne, usually through a chimney stack; or liquid via a pipeline or drains. The fate of the release in the environment is a very complex issue. An airborne release may consist of gases or aerosols. Inert gases such as Xe and Kr could be expected to travel long distances, others may interact rapidly with the atmosphere. Particulates will be deposited at varying ranges depending on a large number of factors (particle size, density, release height, weather conditions etc.), liquids may contain soluble or suspended species; the solubility may change under different conditions (e.g. pH, salinity etc.). The transfer of radionuclides through environmental pathways is strongly dependent on the nature of the radionuclide and the type of discharge. The discharge can be either atmospheric or aquatic, as shown in Fig. 12.5. The large range of radionuclides, in terms of chemicals and physical characteristics, makes generalisations about environmental behaviour extremely difficult. Indeed, for many radionuclides, little is known about their behaviour and transport in the environment. Nevertheless, material released into the environment can be categorised according to the release pathway. Atmospheric discharges can be categorised as either particulate or gaseous, a
626
Chapter 12
Fig. 12.3. Release fractions for a number of radionuclides in liquid effluents from the Sellafield reprocessing facility.
distinction which itself is a simplification, since iodine exhibits semi-volatile characteristics. Aquatic discharges can be, in simple terms, considered as either soluble or insoluble. The fate of atmospheric effluent emissions from a facility is governed by a combination of physical and chemical properties associated with the current atmospheric conditions as well as with the emitted material. The controlling atmospheric properties are associated with: 1. the advective processes which determine the rate of downwind transport from the emission point, and 2. the diffusive processes due to atmospheric turbulence, which govern the rate of dilution during the downwind transport process.
Environmental Monitoringfor Safeguards
627
Fig. 12.5. Schematic presentation of the discharges from a nuclear facility. Transport pathways in the environment.
628
Chapter 12
The controlling physical and chemical properties of the emitted particle or gaseous material that govern their eventual distribution in the environment include the particle size distribution and the chemical stability in the atmosphere. These properties determine the rate of removal of the material from the atmosphere due to gravitational settling of particles, dry and wet deposition, photo-dissociation due to exposure to sunlight, hydrolysis, or chemical reactions with other atmospheric constituents. In order to predict the relative importance of these processes in defining the spatial and temporal evolution of the emitted material in the environment for a given situation, it is necessary to resort to the use of atmospheric dispersion modelling in conjunction with environmental measurements. Deposition from the atmosphere to the ground, vegetation or water surface can occur via three deposition processes. The direct interaction of a material with the surface is called dry deposition and occurs continuously. The second process is called wet deposition and includes the removal of material from the atmosphere in any falling hydrometeors, such as raindrops, snowflakes or hailstones. The third process includes deposition in fog droplets which do not readily fall under gravity due to size limitations. Fog deposition is a process which falls midway between dry and wet deposition and includes mechanisms which are important in both the wet and dry removal processes. The net result is that the airborne species are deposited onto surfaces at varying ranges from the clandestine plant, including into rivers or lakes whereby they may be transported to much greater distances, or remain airborne. Certain organisms (known as "bio-accumulators") may concentrate some of these species. Liquid effluents may be put directly, or indirectly via the drains, into rivers, lakes or seas. The species may be transported over long distances by rivers or sea currents, but some may be removed more quickly, e.g. by sedimentation processes or by bio-accumulators. The range of environmental media and the number of radionuclides that are of interest makes any generalisation on transport in the terrestrial environment impossible. In addition, the limited information available from experimental work means that our understanding of the transfer processes is far from complete (McMahon et al., 1993). There is a disadvantage with the use of environmental samples to assess the radionuclide concentrations in the environment. This arises since there is the potential for contamination of such samples from previously deposited material. This material may or may not originate from the source of interest and there may be significant contributions from Chernobyl and weapons fallout. Such contamination pathways include soil splash, the deposition of resuspended material and root uptake. Radionuclides may be released into one or more of the four sectors of the hydrosphere from a reprocessing facility. These sectors are: rivers, estuaries; local marine zones (close to the source of the effluent) and regional marine zones (distant from the source of the effluent). A discharge into a river may involve the movement of radionuclides through all four sectors. If the discharge occurs into the sea, it is only necessary, in general, to consider the local and distant marine zones. For many radionuclides, the sediments in the water column form an important sink in the environment. Radionuclides can adhere to the surface of mineral grains and can
Environmental Monitoringfor Safeguards
629
also be scavenged from the water column in the process of floc formation. A key parameter relating to the mobility and therefore the dispersion of nuclides is the distribution coefficient, of Ka. This is simply defined as the ratio of concentration or activity of a nuclide in sediment to the concentration or activity of the same radionuclide in the overlying water. As the adsorption of certain radionuclides to sediments depends on the surface area of the sediment available, the finer sediment fractions tend to have higher specific activities than the coarser ones. To some degree, the level of adsorption also depends on the depth of water column and sediment beds. Typical Kd values of selected actinides are in the order 1• tO 1• (McMahon et al., 1993).
12.3 ANALYTICAL M E T H O D O L O G Y One of the important lessons learned from the preparation for the Chemical Weapons Convention (CWC) was the necessity of at least two different spectrometric techniques to absolutely determine compound identification. Additionally, expert interpretation of the results is necessary. Analytical techniques need to be applied to a variety of sample types. Methods should be chosen which are applicable to individual particles, gases, liquid solution, and bulk materials such as soils and other solids. Many such techniques have been developed. Listed in Table 12.9 is a summary of those methods which have been utilized effectively for environmental sample analysis. Those techniques which are more highly recommended are so marked. Most, although not all, of the equipment is commercially available. It is important to emphasize that the open literature shows that scientists have the ability to detect atomic, nanogram, and/or milligram levels of target species. A combination of several techniques for both bulk and individual particle analysis that will yield the level of information is necessary. The analytical requirements can be summarized as follows: 9 focus on ultra-trace analytical methods (low ppb, ppt, levels) 9 no one technique will be key: combinations essential 9 requires bulk and individual particle-analysis techniques 9 requires multi-component analyses for key elemental species, isotopes, and/or organics (more than one laboratory necessary) 9 chain of custody and good laboratory practice are required. In principle, radiometric techniques are most suitable for the shorter-lived radio-nuclides, whereas at longer half-lives (--100 years or more) techniques based on mass spectrometry or neutron activation analysis start to become more sensitive. This is shown in Table 12.10.
12.3.1 Sampling techniques The initial step in the design of any sampling programme is the definition of the objective of the study. The broad objective of this study is to determine whether
630
Chapter 12
Table 12.9 Recommended analytical techniques Methods
Application
*Accelerator Mass Spectrometry, AMS *Ion Microtomography, IMT
36C1, tritium, reprocessing (14C, 1291)
Particle-Induced X-ray Emission, PIXE Nuclear Reaction Analysis, NRA *Hydrogen Mass Spectrometry, HMS *Noble Gas Mass Spectrometry, NGMS *Thermal Ionization Mass Spectrometry, TIMS *Alpha Spectrometer Systems, ASS *Beta Counting Systems, BCS *Gamma Spectrometer Systems, GSS Liquid Scintillation Systems, LSS *Neutron Counting Systems, NCS Direct-Current Plasma Optical Emission Spectrometry, DCP-OES *IR Spectroscopy (e.g. FTIR) Raman Spectroscopy, RS
Particulates; Reprocessing: transuranics; Enrichment: U/Pu isotopes Ca ~ U (spatial resolution of 10-15) Aged tritium bearing samples D/H, HTO, HT Reprocessing all noble gas isotopics (i.e., 85Kr, 131Xe) reprocessing Isotopics abundances for Pb, Sr, Nd, U, Pu, actinides Radioactive isotopes (e.g. actinides, Ra, Rn, 21~ Radioactive isotopes (e.g. 32p, 35S, 90Sr) Radioactive isotopes (e.g. 6~ 137Cs, 237Np, U, Pu,
141/144Ce)
Radioactive isotopes (e.g. 3H, 14C) Radioactive isotopes (e.g. 24~ 244Cm) Impurity metals in Pu and Pu alloys
Explosives, mock explosives, resins, dyes Trace concentrations; organic & inorganic compounds, CBN Anions (i.e. F-, CI-, 504 =, NO2-, NO3-) Ion Chromatography, IC Trace ppt levels of all elements including isotopics *Inductively Coupled Plasma Mass for masses >40 (Reprocessing: Cs, Ce, Sr, U, Pu, Np, Spectrometry, ICP-MS Am isotopes; Enrichment: U isotopes) F-, Br-, CI-, perchlorate, sodide, NO3-, NO*Capillary Electrophoresis, CE *High Pressure Liquid Chromatography, HPLC High explosives HMX, RDX, TNT Trace analysis technique; organic solvents, unknown Gas Chromatography-Mass Spectrometry organics GC-MS Trace organics in mixtures (Enrichment: PTFE, *Ion Trap Mass Spectrometry, ITMS Freon; Reprocessing: TBP, Kerosene; HE Fab/Test: HMX, RDX, TNT, TATB) Ultra-trace organics without liquid sample Triple Quadrupole Mass Spectrometry, TQMS preparation/confirmation Nondestructive, high Z-elements in low Z matrix *X-ray Fluorescence, XRF Specific actinide compounds, rare earths Laser Induced Fluorescence, LIF Particle topography and elemental screening *Scanning Electron Microscopy w/energy dispersive Elemental analysis & mapping of discrete particles Electron Microprobe Trace elemental levels & isotopic-ratio *Ion Microscope (SIMs) determinations (i.e. 6Li) Nondestructive high Z for spot sizes of 0.1 mm (i.e. *XPS Photoelectron Spectroscopy/ESCA boron nitride, uranyl compounds) Structural and multi-phase analysis X-ray Diffractometry *Key analytical techniques.
Environmental Monitoringfor Safeguards
631
Table 12.10 Recommended field- and laboratory-based analytical techniques for environmental monitoring Field-based methods
Laboratory-based Methods
Portable NaI, Ge detectors (also vehicle-based systems)
Radiochemistry with: 9 ~/~t-spectrometry 9 scintillation, proportional and GM-counting TIMS, AMS, ICPMS, NAA
Inorganic materials
Portable XRF
XRF Ion Chromatography
Organics
Ion Mobility Spectrometer
High Resolution GC-MS Infrared Spectroscopy GC with: 9 ECD 9 NPD 9 FPD 9 LC-MS
Radionuclides: 9 "Short" tl/2 (<100 yrs)
9 "Long" tl/2 (>100 yrs)
GCMS: 9 Purge and trap 9 Thermal desorption Atmospheric pressure ionisation mass spectrometry Long-path-length IR spectroscopy
undeclared nuclear operations are being, or have been, carried out in an area by detecting contaminants released into the aquatic environment. In addressing this objective some basic information on the area to be investigated is required. Having identified a possible location of suspected nuclear operation, basic information on the topography, geology, hydrology and climate should be obtainable from maps. The locations for which samples should be taken are obviously related to the inspection scenario. However, there are several generalizations which can be made (McGuire, 1992). In the case of an undeclared suspect facility, the best location to obtain samples would be from inside the facility in question. Grab samples from a storage container or production line are primary collections. Lastly, wipe samples from hardware in the suspect area have also been useful. Polymeric material; e.g., from flange gaskets, closure gaskets, pressure fittings, washers, etc., absorb chemicals with which they have been in contact and so provide a "memory" of previous production runs. The same is true of vegetation (e.g. grasses) and soil (taken to a few centimetres in depth) in the near vicinity of suspect process-related buildings that may have come in contact with product from spills, fugitive emissions and/or personnel traffic. Soil from drainage areas is also most useful as well as from natural depressions and sumps. Liquid samples should be obtained from waste storage and waste treatment areas including samples of liquid from standing water in the vicinity and any nearby streams to which runoff may have occurred. Figure 12.6 shows these possible locations from which environmental samples can be collected (Raber, 1993).
Chapter 12
632
Fig. 12.6. Environmental sampling around potential target facilities.
In order to develop a good sampling strategy the following factors must be considered. 1. Selection of appropriate sample sites. Once the general location of the effluent discharge is known, and the type of water body identified, the most suitable sample stations can be assessed. Following a preliminary survey of the area the choice of appropriate media and equipment can be made. If the discharge is likely to be to a river, samples should be taken above and below the suspected point source if possible, to enable the background levels to be well characterised. In estuaries or coastal waters, transects away from the suspected source may be appropriate for the same reason. 2. Selection of appropriate media to sample. In the aquatic environment, the three main media to consider are water (including suspended sediment), deposited sediment and biota. The advantages and disadvantages of these environmental compartments are discussed in detail in the following sections. If a preliminary site visit is not possible, it may be most appropriate to select more than one medium to sample, to allow for unforeseen difficulties once the site has been reached. 3. Selection of appropriate sampling devices. The choice of appropriate sampling equipment can be made using information on the sample site and the range of media to be sampled. The sampling technologies to be used for this purpose all following" 1. Water: it is necessary to use large volume water samplers which would selectively remove anionic, cationic, non-ionic and particulate radionuclides in-situ from about 103-105 1 of fresh or brackish waters at flow rate of 50 l/min. 2. Air: it is necessary to use large volume air samplers to remove particulate radionuclides and radioiodines from air at flow rates of up to 10 m3/min.
Environmental Monitoringfor Safeguards
633
The large volume sampling is required in conjunction with extremely sensitive analytical techniques if one wants to detect signatures of activities several kilometres away. During the sampling use of portable instruments can significantly reduce the number of samples which need to be taken. Elevated concentrations of radionuclides in the environment close to a suspect location may provide strong evidence of an undeclared nuclear operation involving radioactive material at that location. Proof of the operation will require a combination of evidence of various kinds. Increased levels of a small number of radionuclides are unlikely to be conclusive, unless combined with other evidence for a number of reasons: many radionuclides have legitimate industrial uses and accidental or planned dispersion from numerous locations has occurred in the past; widespread dispersal from nuclear weapon tests and from the Chernobyl accident has led to very variable distribution in some areas, due to effects of terrain and weather; disposal of radioactive waste from various industries may conceivably give rise to local concentrations of some isotopes in soil and vegetation; variability in the distribution of natural radionuclides is large in some areas. However, combinations of radioisotopes characteristic of reprocessing or uranium enrichment, in appropriate ratios and distributed in the expected manner geographically and between different sample types, might constitute irrefutable evidence if requisite quality procedures have been applied. The most appropriate sampling techniques are summarized in Table 12.11. Let us mention, as an example, what needs to be done for uranium sampling. The major transport mechanisms for uranium, on-site and in the immediate proximity, will be: 9 air transport and deposition of particulate matter, and 9 surface water transport. Deposition surfaces that should be considered for sampling of air-transported particles, and corresponding sampling techniques, are given in Table 12.12. Soluble and suspended uranium may be transported off-site by process waste water streams, natural surface drainage, and surface streams. Sample types that should be collected include: 9 bottom sediment; 9 suspended sediment, and 9 water. 12.3.2 Use of bioaccumulators A potentially useful method for determining the presence of any reprocessing activity will be the use of a biological accumulator. Such accumulators might include thyroid glands (for iodine), herbivore faeces (for actinides deposited on grass) and fungi (for species such a s ~~ The processes of accumulation are not always well understood and a more extensive search might reveal alternative accumulators. In addition to these methods, natural materials such as mosses and lichens collected close to a site might be useful for collecting materials over lengthy periods. As an example Table 12.13 shows the range of concentration factors for 137Csin various terrestrial plants. While a wide
Chapter 12
634
Table 12.11 Sampling techniques On-site
Mobile Lab. High Vol. Air: 9 Filter, 9 Charcoal (Iodine isotopes); 9 3H Liq. Scint.; 9 Noble gases - compression/counting Sample Screening: 9 at-counting; 9 ),-spectrometry; 9 Portable GC. Sample Collection: 9 wipes; 9 effluents; filters; waters (with preconcentration); 9 protective clothing.
Short Range (10 km)
Mobile Lab. High Vol. Air: 9 Filter; 9 Charcoal (Iodine isotopes); 9 3H Liq. Scint.; 9 Noble gases - compression/counting Sample Screening: 9 High Vol. Air: 9 Filter; 9 Charcoal (Iodine isotopes); 9 Vegetation; 9 Soil; 9 Stream water: above site below site - Tritium, grab samples Meteorology
Long Range
Sample Collection: 9 Stream water Meteorology
Table 12.12 Surfaces to be considered for uranium sampling Deposition Surface
Sample type/method
Interior surfaces where particles accumulate
Wipe sample
Pine needles, grasses, hairy leaves
Vegetation or faecal matter, grab sample
Ground surface
Surface soil or road surface sample
Environmental Monitoringfor Safeguards
635
Table 12.13 Range of concentration factors for
137Csin various terrestrial plant species (After Nicholson et al.,
Plant
Range of values of concentration factor
Grass (pasture)
0.037-0.47 0.16-1.6
Grass (upland species)
0.3-2.0 (peat soil)
Cereal
0.011-0.016 0.04-0.21
Green Vegetables
0.048-0.098 0.19-0.68
Root Vegetables
0.01-0.02 0.13-2.72
Heather
0.6-7.0 0.7-6.37
Ericacaea (Vaccinium species) Moss
0.9-4.0 0.41-23.5
Fungi
0.2-36.6 (typ.) 21.6-92.7 (peat soil)
Fungi (Cortinariaceae species)
20-50
1994)
range of pollutants may be accumulated by a given species, it has been observed also that some species concentrate only specific elements or radionuclides. Whilst there are relatively few data on bioaccumulation according to species type, the efficiency of bioaccumulation varies markedly, even for different species within a given family type. Terrestrial bioindicators like mosses, lichens and fungi show a marked ability to accumulate certain radionuclides. Heather and some other upland species have also shown this potential. The higher plants, grasses, cereals, fruiting trees, food crops, forest trees etc. can possess good interception characteristics but are generally poor accumulators due partly to seasonal growth patterns. The most common bioindicators are lichens and mosses. There is a long history of lichens and mosses being used as indicators of air pollution and there is an extensive body of literature pertinent to the subject (Hawksworth, 1971). However, much of what has been written concerns the use of these plants an indicators of industrial pollution in urban areas, for example in the monitoring of sulphur dioxide levels and heavy metal deposition (Goodman and Roberts, 1971; Rose, 1970). Mosses and lichens have a proven ability to concentrate various heavy metals in large quantities, usually greatly surpassing vascular plants. It is noteworthy that among the lichens, most species are exceptionally sensitive to man-made atmospheric pollution, e.g. SO2, NOx and so are almost absent in densely inhabited or industrialised areas. Among the mosses, however, several varieties may develop and survive in this type of environment with no visible deficiency (Nicholson et al., 1994).
636
Chapter 12
Interest in the uptake of radionuclides by lichens and mosses originated from a concern in Scandinavian and Arctic countries about the transfer of radioactivity through the lichen-reindeer-man food chain. Here, the long-lived radionuclides ~37Cs, 9~ and 239pu, originating from atmospheric testing of nuclear weapons between 1952 and 1963, were found to contribute significantly to the radiation burden of people living on reindeer (caribou) meat. Many studies have been undertaken in order to estimate the possible resulting dose to man (Lid6n and Gustafsson, 1967; Holm and Persson, 1978; Hanson, 1980). More recently, mosses have been used as indicators of airborne radionuclides near major nuclear installations (Sumerling, 1984) and for the regional mapping of 137Cs fallout from the Chernobyl accident (Steinnes and Njastad, 1993). Mosses and lichens have thus been established as very useful bioindicators of radioactivity in the environment. Lichens are adapted to accumulate all the elements necessary for their life from the atmosphere. They have no root system and absorb very little from the substrate on which they grow. Atmospheric materials, including trace metals and radionuclides, can be concentrated by particulate entrapment, ion exchange, electrolytic sorption and processes mediated by metabolic energy (Cr~te et al., 1992). Passive particulate trapping is, however, thought to be the dominant uptake mechanism. This is also true for mosses which absorb nutrients directly through leaf and stem surfaces. The special characteristics of lichens which enhance the uptake of airborne radionuclides include the following. The aerial parts of the plant are persistent and are exposed to deposition all year round. Lichens have a slow growth rate and a long life span with mature plants reaching 45 years of age. This means that the turnover of biomass is small and there is great potential for the accumulation of elements in the plant tissue. Finally, the physical structure of most lichens is adapted to intercept atmospheric debris. They characteristically have a high surface-to-mass ratio. When compared against the use of, for example, sediment or glacial samples to study atmospheric fallout, several advantages are apparent. Lichens are abundant worldwide and can be found in many different types of environment and climate. Samples are generally easy to collect and analysis is straightforward. Lastly, their uptake of radionuclides from air and rainwater is comparatively rapid and, unlike some other bioaccumulators, non-selective. However, if quantitative information is required, rather than simply an indication that deposition has occurred, then the interpretation of lichen radionuclide concentration data is complicated by several factors. For example, the fractional amount of the radionuclide that is intercepted and retained by the plant will vary according to habitat, rainfall, growth rate and species (Hanson, 1967). Once the radionuclide is stored within the plant, similar environmental parameters will act to cause variations in the elimination rates from the lichen. Rainfall, growth rate and the availability of nutrients and metals at the lichen epidermal surfaces will be important (Ellis and Smith, 1987). In addition to lichens and mosses being used as bioindicators there is also a number of publications reporting the use of fungi and mushrooms.
Environmental Monitoringfor Safeguards
637
Interest in the accumulation characteristics of fungi arose from the discovery that they can sometimes contain large amounts of heavy metals, for example Hg, Cd and Pb and other trace elements such as Se. Measured concentrations are not only high relative to vascular plants but can be large in absolute terms, for example in the Agaricus species up to 100 ~tg g-1 (d.w.) of Cd has been found (Seeger et al., 1978). Silver has also been observed to accumulate very well in this species (Byrne et al., 1979). Following the Chernobyl accident, investigations focused on the uptake of radionuclides in fungi, especially ~3VCs and ~~ Many studies have arisen from concern about elevated radioactive contamination levels in edible mushrooms and the subsequent transfer to grazing animals and man. From the collection of data available, there is good evidence of the ability of fungi species to accumulate radionuclides. For organisms like mushrooms, which grow in the upper soil layer and obtain minerals from the soil solution, long-lived fission products have time to migrate down through the surface soil and be taken up along with other nutrients, but the short-lived nuclides may decay before being incorporated. The transfer of ~37Csfrom soil to fungi is correlated to the acidity of the soil, because in common with many other mineral elements, its solubility and mobility increases with a decrease in pH value. This is due to the fact that caesium ions bound to clay minerals in the soil can be exchanged for hydrogen ions. When the soil has a high pH value, less ion exchange is possible, the caesium remains bound and is not, therefore, available for uptake by the fungus. The type of soil is also important for the following reason. Cations that are washed into mineral soils are normally removed rapidly from solution by bonding to exchange sites which occur on both clay minerals and organic molecules. Cations such as potassium and caesium can also be held in the interlayer space of micaceous clay minerals because they have a small ionic radius. They cannot be released until the mineral is weathered and thus, in mineral soils, radiocaesium can rapidly become immobilised. Organic soils typically have a high cation exchange capacity with a low and variable clay content. Bonding of cations is therefore more variable and it is more likely that they will be available for uptake by microorganisms and plant roots (Nicholson et al., 1994). While it is true to say that amongst terrestrial systems the most striking accumulation of radionuclides occurs in fungi and lichens, there are a small number of other plants in which lower but significant concentrations can be found. After the Chernobyl accident, high levels of activity were observed on a great number of species of vegetation, including deciduous plants, trees, conifers, fruit trees, vegetable crops and grasses (Sawidis, 1988; Barci et al., 1988). As well as depending upon the degree of deposition in a given area, the variation in the recorded levels on different plants was the direct result of their efficiency at intercepting and retaining airborne particulates. An example of a highly efficient interception system is the canopy of a coniferous forest. The large specific surface area of pine and spruce trees exceeds that of broadleaved deciduous trees, giving them a high scavenging efficiency for radioactive aerosol particles. Pine needles have been acknowledged as useful monitors of atmospheric pollution (Eriksson et al., 1989). However, radionuclides will be redistributed in the forest ecosystem due to various removal mechanisms such as rain
638
Chapter 12
washing, wind erosion, foliar penetration, root uptake and needle fall. After the Chernobyl accident, Bunzl et al. (1989) quoted an environmental half-life for 134'137Csin a spruce stand of 230 days. This will be due primarily to retention on tree (bark) and foliar surface rather than absorption and subsequent accumulation of radionuclides. The principal removal mechanism is thought to be needle fall with annual losses of 134'137Cs from the canopy in the order of 15% being reported (Bonnett and Anderson, 1993). It is therefore noteworthy that the forest floor in this type of ecosystem should be a very effective sink of airborne radionuclides and plants such as bryophytes and fungi, as well as the forest litter itself, will be able to accumulate higher than average levels of contamination (Nicholson et al., 1994). Analysis of herbage (of forage) has sometimes been used to detect and identify radionuclides deposited from the atmosphere (Jackson et al., 1981). However, the problem arises that when the deposition rate is low, large areas of vegetation need to be sampled for detection. In the case of plutonium, an alternative is to collect the faeces of grazing animals such as cows, sheep and rabbits. Plutonium is very poorly absorbed by the mammalian gut and so virtually all that is ingested by an animal will appear in its faeces. Also, if the species selected obtains its food entirely by grazing, then the isotopic ratio 238pu" 239+24~ will be the same in the faeces as deposited on the vegetation. Thus, the method may be used to detect fallout in a particular area.
REFERENCES Annual Report on Radioactive Discharges and Monitoring of the Environment, Volume I: Report on discharges and environmental monitoring British Nuclear Fuels plc. Health and Safety Directorate, Risley, Warrington, Cheshire, UK (1992). Barci, G., Dalmasso, J. and Ardisson, G., Chernobyl fallout measurements in some Mediterranean biotas. Sci. Total Environ., 70 (1988) 373-387. Bonnett, P.J.P. and Anderson, M.A., Radiocaesium dynamics in a coniferous forest canopy: a mid-Wales case study. Sci. Total Environ., 136 (1993) 259-177. Bunzl, K. Schimmack, W., Kreutzer, K. and Schierl, R., Interception and retention of Chernobyl-derived Cs-134, Cs-137 and Ru-106 in a spruce stand. Sci. Total Environ., 78 (1989) 77-87. Byrne, A.R., Radioactivity in fungi in Slovenia, Yugoslavia, following the Chernobyl accident. J. Environ. Radioactivity, 6 (1988) 177-183. Cr~te, M., Lefebvre, M.A. and Zikovsky, L., Cadmium, lead, mercury and ~37Csin fruticose lichens of northern Qurbec. Sci. Total Environ., 121 (1992) 217-230. Ellis, K.M. and Smith, J.N., Dynamic model for radionuclide uptake in lichen. J. Environ. Radioactivity, 5 (1987) 185-208. Eriksson, G., Jenson, S., Kylin, H. and Strachan, W., The pine needle as a monitor of atmospheric pollution. Nature, 341 (1989) 42-44. Goodman, G.T. and Roberts, T.M., Plants and soils as indicators of metals in the air. Nature, 231 (1971) 287-292. Hanson, W.C., Radioecological concentration processes characterising arctic ecosystems. In" B. Aberg and F.P. Hungate (Eds.), Radioecological Concentration Processes. Pergamon Press, Oxford, UK, 1971, pp. 183-191. Hanson, W.C., Transuranic elements in arctic tundra ecosystems. In: W.C. Hanson (Ed.), Transuranic Elements in the Environment pp. 441-458. U.S. Dept. of Energy, DOE/TIC-22800, Technical Information Centre, Springfield, VA, USA, 1980.
Environmental Monitoringfor Safeguards
639
Hawksworth, D.L., Lichens as litmus for air pollution: a historical review. Int. J. Environ. Studies, 1 (1971) 281-296. Holm, E. and Persson, R.B.R., Fall-out plutonium in Swedish reindeer lichens. Health Phys., 29 (1975) 43-51. Jackson, W.M., Noakes, J.E. and Spaulding, J.D., Forage: A sensitive indicator of airborne radioactivity. Health Phys., 40 (1981 ) 84-91. LidEn, K. and Gustafsson, M., Relationships and seasonal variation of '37Cs in lichen, reindeer and man in northern Sweden 1961-1965. In: B. Aberg and F.P. Hungate (Eds.), Radioecological Concentration Processes. Pergamon Press, Oxford, UK, 1967, pp. 193-208. McGuire, R.R., The role of chemical analysis in support of inspections to verify compliance to chemical weapons treaties, UCRL-ID- 110816, May 1992, Lawrence Livermore National Laboratory. McMahon, A., Roberts, P., Nicholson, K., Toole, J., Wickenden, D., Watterson, J. and Adsley, I., The detection of environmental signatures from covert nuclear fuel reprocessing activity. AEA Technology Report, March 1993. Nicholson, K.W., Rose, C.L., Garland, J.A., MxKay, W.A. and Pomeroy, I.R., Environmental sampling for the detection of undeclared nuclear activities, AEA Technology Report: AEA FS 024(H), February 1994. Raber, E., Potential applications of environmental sampling and analysis for the IAEA, Paper presented at IAEA Consultant's Meeting, Vienna, Austria, March 30-April 2, 1993. Rose, F., Lichens as pollution indicators. Your Environment, 1 (5), (1970) 185-189. Sawadis, T., Uptake of radionuclides by plants after the Chernobyl accident. Environ. Pollut., 50 (1988) 317-324. Seeger, R., Ntitzel, R. and Dill, V., Z. Lebensm, Unters.-Forsch., 166 (1978) 23. Steinnes, E. and Njastad, O., Use of mosses and lichens for regional mapping of '37Cs fallout from the Chernobyl accident. J. Environ. Radioactivity, 21 (1993) 65-73. Sumerling, T.J., The use of mosses as indicators of airborne radionuclides near a major nuclear installation. Sci. Total Environ., 35 (1984) 251-265.
641
CHAPTER 13
Comprehensive Test Ban Treaty (CTBT)
In April 1954, almost 10 years after the first nuclear weapon test was conducted in July 1945, Prime Minister Jawaharlal Nehru of India proposed that nuclear weapon testing be suspended. His proposal was the first initiative of its kind. The arms control negotiations between the then USSR and the USA started in the late 1950s. Much of the negotiation time was spent on the issues of monitoring a nuclear test ban, especially on the technical details of the detection of nuclear weapons tests. The USSR and the US did not agree in 1959 about methods to detect small underground nuclear explosions and therefore settled for the Limited Test Ban Treaty of 1963, which excluded underground tests from consideration for the time being. The Threshold Test Ban Treaty of 1974 limited the yield of underground nuclear weapon tests to 150 kilotons (the equivalent of the explosive force of approximately 150,000 tonnes of trinitrotoluene (TNT). Over 2000 nuclear weapon test explosions were registered during the 51 years between the conduct of the first nuclear test and the opening for signature of the Treaty in September 1996. Science and politics in early nuclear test ban negotiations have been presented recently in an interesting article by Barth (1998). The CTBT was hammered out in the Conference on Disarmament, a 61-member international forum, beginning in January 1994. In that committee, the US, Russia, the UK, France and China (members of the "nuclear club") and a number of the non-nuclear weapons states worked together. Negotiations were greatly aided by behind-the-scenes efforts from a number of states, especially Australia, as well as by the diplomatic skills of the committee chairs. At the last moment, the Conference on Disarmament failed to reach a consensus on the final text of the treaty, primarily due to opposition from India, which prevented the committee from forwarding the treaty to the UN. Belgium, along with 127 cosponsors, kept the treaty alive by introducing the text to the UN General Assembly as a resolution, a process that had never before been attempted. Discussion in the UN was brief, and on 10 September 1996 the General Assembly voted 158 to 3 in support of the treaty (Sullivan, 1998).
642
Chapter 13
The CTBT serves two distinct and important roles. It prevents the declared nuclear weapons states and all other parties to the treaty from developing with confidence new or advanced weapon designs, including "mini-nukes", and it establishes a strong international norm against nuclear proliferation by any state. The CTBT complements and strengthens the regime established by the 1970 Non-Proliferation Treaty, which was renewed for an indefinite period in May 1995 in another hard-won victory. These two arms control measures are linked by the preamble of the Non-Proliferation Treaty, which cites the preamble of the Limited Test Ban Treaty, in which the nuclear signatories expressed their intent to "achieve the discontinuance of all test explosions of nuclear weapons for all time". At a meeting of State Signatories on 19 November 1996, a Preparatory Commission for the Comprehensive Nuclear-Test-Ban Treaty Organization was established. The Preparatory Commission is an international organization financed by the State Signatories, which has been set up to establish the global verification regime of the Treaty and to prepare for its entry into force. The Preparatory Commission consists of two organs: a plenary body composed of all the State Signatories and the Provisional Technical Secretariat. The Preparatory Commission appointed Wolfgang Hoffmann of Germany as its Executive Secretary on 3 March 1997. The Executive Secretary is head of the Provisional Technical Secretarial which started work at its offices in the Vienna International Centre on 17 March 1997. The Preparatory Commission has two subsidiary bodies: Working Group A on administrative and budgetary matters, and Working Group B on verification issues. Both Working Groups make proposals and recommendations for consideration and adoption by the Preparatory Commission at its plenary sessions. Concerns have been raised about the possibility of the continuation of development of new types of nuclear weapons by nuclear weapon states. This could in principle, be realized through "virtual testing", using advanced computers and data provided by experimental facilities such as the billion-dollar laser-fusion National Ignition Facility (NIF), at Lawrence Livermore National Laboratory, USA, and in similar facilities elsewhere. Of particular concern is that the stockpile stewardship program could open up a route to pure-fusion weapons in which thermonuclear explosions would be ignited by means other than fission explosions (Jones et al., 1998). Such designs would make it possible to bypass the techniques used today to verify nuclear nonproliferation (and, in the future, reductions), namely, international controls on highly enriched uranium, plutonium and other artificial fissile materials. The USA has openly stated the official view that "inertial confinement fusion (ICF) and similar experiments" are not banned by the treaty. But the analysis did not define "similar experiments", because of a stalemate over how far to expand the exemption for pure-fusion explosions beyond the one that now exists for ICF. In laser- and particle-beam-driven ICF, a millimetre-scale capsule of deuterium and tritium (D-T) would be imploded to create a sufficiently high density (--1025 g cm -3) and temperature (--10 keV) at the centre to ignite the thermonuclear reaction D+T 4He+n+l 7.6 MeV. The fusion "burn" would then propagate through the surrounding
Comprehensive Test Ban Treat),
643
fuel in less than 10-1~seconds that the pellet remained compressed. An objective being the energy releases equivalent to the explosion of up to tens of kilograms of TNT will be achieved in this way. The Russian weapons laboratories have also attempted to generate pure-fusion explosions with implosions driven directly by chemical explosives. These efforts became known in early 1992, when one of Russia's weapons labs hosted an international conference at which several papers were presented on the systematic efforts made in the Soviet Union since the early 1950s to ignite fusion through chemical implosions of D-T gas. In theory, it should be possible to create a shock pressure high enough to achieve ignition through the cumulation of multiple shock-waves in a spherical system consisting of alternating layers of dense and light materials. In practice, however, deviations from perfect symmetry and mixing between the D-T fuel and the liner compressing it were reported to have limited the temperatures achieved to an order of magnitude below the ignition threshold and neutron yields to less than 10 ~4 neutrons (see Anisimov et al., 1992 as referenced in Jones et al., 1998). A clear line needs to be drawn between permitted and forbidden experiments with pure-fusion explosions. The Comprehensive Test Ban Treaty does not define matter, and several governments have found it too difficult to anticipate future technological developments to make their own definitive interpretation.
13.1 THE TREATY 13.1.1 Preamble to the Treaty "The State Parties to this Treaty (hereinafter referred to as "the State Parties"), Welcoming the international agreements and other positive measures of recent years in the field of nuclear disarmament, including reductions in arsenals of nuclear weapons, as well as in the field of the prevention of nuclear proliferation in all its aspects, Underlining the importance of the full and prompt implementation of such agreements and measures, Convinced that the present international situation provides an opportunity to take further effective measures towards nuclear disarmament and against the proliferation of nuclear weapons in all its aspects, and declaring their intention to take such measures, Stressing therefore the need for continued systematic and progressive efforts to reduce nuclear weapons globally, with the ultimate goal of eliminating those weapons, and of general and complete disarmament under strict and effective international control, Recognizing that the cessation of all nuclear weapon test explosions and all other nuclear explosions, by constraining the development and qualitative improvement of nuclear weapons and ending the development of advanced new types of nuclear
644
Chapter 13
weapons, constitutes an effective measure of nuclear disarmament and non-proliferation in all its aspects, Further recognizing that an end to all such nuclear explosions will thus constitute a meaningful step in the realization of a systematic process to achieve nuclear disarmament, Convinced that the most effective way to achieve an end to nuclear testing is through the conclusion of a universal and internationally and effectively verifiable comprehensive nuclear-test-ban treaty, which has long been one of the highest priority objectives of the international community in the field of disarmament and nonproliferation, Noting the aspirations expressed by the Parties to the 1963 Treaty Banning Nuclear Weapon Tests in the Atmosphere, in Outer Space and Under Water to seek to achieve the discontinuance of all test explosions of nuclear weapons for all time, Noting also the views expressed that this Treaty could contribute to the protection of the environment, Affirming the purpose of attracting the adherence of all States to this Treaty and its objective to contribute effectively to the prevention of the proliferation of nuclear weapons in all its aspects, to the process of nuclear disarmament and therefore to the enhancement of international peace and security, Have agreed as follows..."
13.1.2 Summary of the Treaty The Comprehensive Nuclear-Test-Ban Treaty consists of a preamble, 17 articles, two annexes and a Protocol. The Protocol describes verification procedures and contains two annexes. One annex lists the 337 facilities comprising the International Monitoring System (IMS) and the other annex describes parameters for standard event screening by the International Data Centre (IDC). The preamble stresses the need for "continued systematic and progressive efforts to reduce nuclear weapons globally" with the ultimate goal of their elimination and of "general and complete disarmament under strict and effective international control". It recognizes "the cessation of all nuclear disarmament and non-proliferation in all its aspects". Under Article I (Basic Obligations): 1. Each State party undertakes not to carry out any nuclear weapon test explosion or any other nuclear explosion, and to prohibit and prevent any such nuclear explosion at any place under its jurisdiction or control. 2. Each State party undertakes, furthermore, to refrain from causing, encouraging, or in any way participating in the carrying out of any nuclear weapon test explosion or any other nuclear explosion." Article II (The Organization) establishes the Comprehensive Nuclear-Test-Ban Treaty Organization to ensure the Treaty's implementation and provide a forum for consultation and cooperation. With its seat in Vienna, it will comprise three organs. The Conference of the State Parties will oversee the Treaty's implementation and the
Comprehensive Test Ban Treaty
645
activities of the other two organs. The Executive Council, with a membership of 51 State Parties, will be the principal decision-making body of the Organization and responsible for supervising its activities. The Technical Secretariat, headed by a Director-General, will assist State Parties to implement the Treaty and carry out verification and other functions. It will supervise and coordinate the operation of the International Monitoring System (IMS) and operate the International Data Centre (IDC) at Vienna. Article III (National implementation measures) requires each State party to take any necessary measures to implement its obligations under the Treaty, including the establishment of a National authority for liaison with the Organization and other State Parties. Article IV (Verification) and the Protocol establish the verification regime. Such a regime--consisting of IMS, IDC, consultation and clarification, on-site inspections and confidence-building measures--"shall be capable of meeting the verification requirements of the Treaty" at its entry into force. Verification activities should be based on objective information, limited to the subject matter of the Treaty, and carried out on the basis of full respect for the sovereignty of State Parties and in the least intrusive manner possible consistent with the effective and timely accomplishment of their objectives. Each State party, however, "shall refrain from any abuse of the right of verification". International Monitoring System. The purpose of IMS is to detect and identify nuclear explosions prohibited under article I. As set out in annex 1 to the Protocol, IMS will consist of 50 primary and 120 auxiliary seismological stations equipped to detect seismic activity and distinguish between natural events--such as earthquakes--and nuclear explosions. It will also include 80 radionuclide stations--40 of them capable of detecting noble gases--designed to identify radioactive particles released during a nuclear explosion. The radionuclide stations will be supported by 16 laboratories. In addition, 60 infrasound and 11 hydroacoustic stations will be designed to pick up the sound of a nuclear explosion in the atmosphere or under water, respectively. International Data Centre. The monitoring stations will transmit data to the International Data Centre (IDC) at Vienna. As set out in part I of the Protocol, IDC will produce integrated lists of all signals detected by IMS, as well as standard event lists and bulletins, and screened event bulletins that filter out events that appear to be of a non-nuclear nature. Both raw and processed information will be available to all State Parties. Consultation and clarification. The consultation and clarification component of the verification regime encourages State Parties to attempt to resolve, either among themselves or through the Organization, ambiguous events before requesting an on-site inspection. A State party must provide clarification of an ambiguous event within 48 hours of receiving such a request from another State party or the Executive Council. On-site inspection. If the matter cannot be resolved through consultation and clarification, each State party can request an on-site inspection. The procedures for on-site inspections, which "shall be carried out in the area where the event that
646
Chapter 13
triggered the on-site inspection request occurred" are established in part II of the Protocol. Confidence-building measures. To reduce the likelihood that verification data may be misinterpreted, each State party will voluntarily notify the Technical Secretariat of any single chemical explosion using 300 tonnes or more of TNT-equivalent blasting material on its territory. In order to calibrate the stations of IMS, each State party may liaise with the Technical Secretariat in carrying out chemical calibration explosions or providing information on chemical explosions planned for other purposes. Article V (Measures to redress a situation and to ensure compliance, including sanctions) empowers the Conference to revoke a State's fights under the Treaty, to recommend to State Parties collective measures in conformity with international law, or, alternatively, if the case is urgent, to bring the issue to the attention of the United Nations. Article VI (Settlement of disputes) describes the mechanisms by which disputes concerning the application or interpretation of the Treaty may be settled. Subject to certain conditions, the International Court of Justice may be requested to give an advisory opinion. Article VII (Amendments) gives each State party the right to propose amendments to the Treaty, the Protocol or the annexes to the Protocol at any time after the Treaty' s entry into force. The proposed amendment requires the approval of a majority of State Parties at an amendment conference with no party casting a negative vote. Article VIII (Review of the Treaty) stipulates that a conference to review the operation and effectiveness of the Treaty will be held 10 years after its entry into force, "unless otherwise decided by a majority of the State Parties". Such review would take into account "any new scientific and technological developments". Further review conferences may be held with the same objective at intervals of 10 years thereafter, or less, if the conference so decides in the preceding year. At the request of any State party, the conference may "consider the possibility of permitting the conduct of underground nuclear explosions for peaceful purposes". If it permits such explosions by consensus, then the review conference "shall commence work without delay, with a view to recommending to State Parties an appropriate amendment to this Treaty that shall preclude any military benefits of such nuclear explosions". Article IX (Duration and withdrawal) states that the Treaty is of unlimited duration. The next four articles (X, XI, XII and XIII) deal with the status of the Protocol and the annexes; signature; ratification; and accession. Under Article XIV (Entry into force), the Treaty will enter into force 180 days after the 44 States listed in annex 2 to the Treaty have deposited their instruments of ratification with the Secretary-General of the United Nations, "but in no case earlier than two years after its opening for signature". This list comprises the States that formally participated in the 1996 session of the conference on Disarmament, and that appear in Table 1 of the December 1995 edition of "Nuclear Research Reactors in the World" and Table 1 of the April 1996 edition of "Nuclear Power Reactors in the World", both compiled by the International Atomic Energy Agency.
Comprehensive Test Ban Treaty
647
If the Treaty has not entered into force "three years after the date of the anniversary of its opening for signature", the Secretary-General of the United Nations, as Depositary of the Treaty, could, at the request of a majority of States that had ratified it, convene a conference to examine the situation and to "decide by consensus what measures consistent with international law may be undertaken to accelerate the ratification process" in order to facilitate the Treaty's early entry into force. Article XV (Reservations) states that the Treaty's provisions are not subject to reservations. Article XVI (Depositary) establishes the Secretary-General of the United Nations as the Treaty's Depositary. Under Article XVII (Authentic texts), the Treaty texts in Arabic, Chinese, English, French, Russian and Spanish are equally authentic. Protocol: Part I describes the International Monitoring System (IMS) and outlines the functions of the International Data Centre (IDC). Part II sets up the procedures for on-site inspections. It specifies the process of designation of inspectors and inspection assistants, their privileges and immunities, points of entry, arrangements for use of non-scheduled aircraft, approved inspection equipment, on-site inspection requests, inspection mandate and notification of inspection. Pre-inspection activities and the conduct of inspections are described in detail. Part III deals with confidence-building measures under article IV (Verification) of the Treaty.
13.2 INTERNATIONAL MONITORING SYSTEM The Treaty has a Protocol under which an International Monitoring System (IMS) and an International Data Centre (IDC) are being established as part of the global verification regime foreseen under article IV (Verification). IMS will consist of a global network of 321 monitoring stations, as well as 16 laboratories, capable of detecting nuclear explosions worldwide. This network of 170 seismic, 80 radionuclide, 60 infrasound and 11 hydroacoustic stations, as well as 16 radionuclide laboratories--comprising a total of 337 facilities--will supply data for processing and analysis to IDC. Both the raw and processed data will be available to all the State Parties. If a suspicious occurrence cannot be resolved through consultation and clarification, each State party has the fight to request an on-site inspection.
13.2.1 The Seismic Monitoring Network The Seismic Monitoring Network serves primarily to detect underground explosions. One part is the Primary Seismic Network of 50 stations that will send data to the International Data Center continuously and in near real time. Approximately 60% of
648
Chapter 13
these stations are or will be seismometer arrays that provide improved signal-to-noise ratios and directional information. For events detected by the primary network that are deemed worthy of closer inspection, data will be requested from the auxiliary Seismic Network of 120 stations. The Auxiliary sites are mostly three-component stations (not arrays), but those closest to events detected by the primary network will be invaluable in improving localization accuracy and the discrimination of earthquakes from underground explosions through the exploitation of regional seismic signals (ranges of less than 2000 km). The combined primary and auxiliary networks are expected to have a detection threshold corresponding to an underground nuclear explosion yield of less than 1 kiloton TNT equivalent fully coupled, independent of location, with event localization in the range of 100-1000 km 2, depending on how well the source-to-receiver paths are characterized. Seismologists have been involved in technical discussion from the 1950s~both the USA and the USSR instructed scientists to improve the detection capabilities of a monitoring system and to find better criteria for distinguishing between earthquakes and underground explosions. A major debate developed at the 1959 talks around the capabilities of seismic instruments. In their new measurements, the USA scientists had used standard shortperiod seismographs of the "Benioff" type, named after Caltech seismologist and instrument specialist Hugo Benioff. Challenging the performance of the Benioff offered the soviets an opportunity to attack the validity of the new seismic data. Although the USA seismologists regarded the Benioff as the best available instrument for the detection of short-period seismic waves, both delegations had agreed on a somewhat different and ultimately superior device. Furthermore, the Soviet instrumentation expert Pasechnik tried to prove that Benioff's instrument considerably distorted seismic signals. Fisk rejected Pasechnik's attack as unfounded and emphasized that no existing instrument, not even any Soviet seismograph, fulfilled the 1958 Geneva specifications. In any case, it was argued, that the performance of the Soviet standard short-period instrument known as the SVKM could not compete with that of the Benioff. The controversy about capabilities shaped nuclear arms control negotiations from the late 1950s until 1997, when an agreement on a comprehensive test ban treaty was reached. And the controversy shaped seismology; in the early 1960s the Department of Defense increased government funding for R&D in seismology by more than a factor of 30, transforming the small academic discipline into a major academic-industrialmilitary endeavour (Bolt, 1976). During an underground nuclear explosion, almost all the energy released is trapped in the first few tens of metres surrounding the explosion point. The energy is mainly absorbed by the vaporization of the rock in the immediate vicinity of the explosion point, then by the fracturing of the material. Beyond a certain distance, known as the "elastic radius", the shock wave is sufficiently attenuated for the movements to be reversible: after the wave has passed, the material returns to its initial state. All that
Comprehensive Test Ban Treaty
649
remains is a vibration which propagates through the Earth in the form of seismic waves. The percentage of the total energy which thus escapes from the explosion zone is very low, from 0.01% to a maximum of 5%. This percentage, known as the "coupling", varies mainly as a function of the mechanical characteristic of the medium around the explosion. The coupling thus differs from one test site to another, but it also varies within a given test site and, in particular, with the explosion depth. The elastic radius delimits a zone inside which the displacements are irreversible. For a given medium, the size of this zone increases with explosion yield and decreases with depth. Observed from long distances compared with this radius, that is, beyond a few kilometres, an underground nuclear explosion can be represented, in seismic terms, by a single point emitting an isotropic seismic wave whose amplitude and frequency content are governed by the explosion yield, the depth of the zero point and the mechanical characteristics of the material surrounding this point. The seismic wave generated by an explosion is a compression wave: the vibration is parallel to the direction of propagation of the wave. On the surface, the initial movement of this wave corresponds to a lifting of the ground. However, this is not the only propagation mode for seismic waves. Three other types are possible: shear wave, Love wave and Rayleigh wave. For use as a monitoring tool, a seismic network must be able to both detect and identify the source of the seismic signal. Detection consists of recognizing that a seismic event has occurred and locating the source of the seismic signal. Identification involves discriminating whether the source was an earthquake, or an explosion. In addition, the monitoring system must be able to perform these tasks despite any plausible attempts to evade the monitoring system. The most problematic evasion scenario is decoupling; that is, muffling the seismic signal by detonating the explosion in a large underground cavity. Decoupling theory was developed in 1958, at Teller' s suggestion by scientists at the RAND Corp. The theory suggested that the amplitude of seismic signals from an underground nuclear explosion could be reduced by as much as a factor of 300, if the nuclear device were exploded in a large underground Cavity. The implications for a test ban treaty were obvious: an explosion 20 times the yield of the Hiroshima bomb (15 kt) would not be detected by the monitoring system. If the theory turned out to be valid and if such large underground cavities could be constructed secretly, it would strike a serious blow to the CTBT monitoring. For the four main vibration modes propagation velocity in a given medium is different. The fastest wave, the P (primary) wave, is the compression wave; the medium is alternately compressed and expanded in the direction of propagation of the seismic wave. For the rocks most frequently encountered at the surface of the globe, the velocity, Vp of this wave varies from 2 km/s (sediments) to more than 6 km/s (granite); in water its velocity is 1.5 km/s. The maximum velocity is 13.7 kilometres/second at a depth approaching 3000 km inside the Earth. The shear wave or S (secondary) wave, has a vibration perpendicular to the direction of propagation. This wave does not propagate in water. For the vast majority
650
Chapter 13
of rocks, the velocity of the S wave is little more than half that of the P wave. In a homogeneous medium, an explosion does not generate an S wave. However, at each discontinuity encountered, part of the P wave is converted into S waves and vice versa. As underground nuclear tests are carried out at a depth where the density of discontinuities in the Earth is highest, that is, close to the surface, a non-negligible part of the emitted P wave is converted into S wave. The Rayleigh and Love waves propagate at the surface of the Earth. Because of their mechanism, which is essentially symmetrical, and in contrast to earthquakes, the explosions generate very few surface waves. The amplitude of the surface waves enables the magnitude of an earthquake on the Richter scale to be calculated. An earthquake of M s --- 7.0 (approximately 30 earthquakes of magnitude greater than this value occur per year in the world) causes ground displacements of approximately 100 ~tm at 10,000 km. This is not the case for explosions, for which the surface waves are almost undetectable: their amplitude is only a few micrometres at 1000 kilometres from a 100-kt explosion. The propagation of the P and S waves, which are known as volume waves because they propagate inside the ground or the ocean, obeys the laws of classical optics. At each change of medium, the waves are reflected and refracted according to Descartes' law. By analogy with optics, their path is often represented by a seismic ray, it being understood that, instead of a single propagating optical ray, two seismic waves (P and S) follow the same path with different vibration modes and velocities. The recording at long distances of the seismic waves generated by underground nuclear explosions has four steps: 9 detection of the seismic waves; 9 identification of the event type (discrimination between nuclear explosion and earthquake); 9 location of the source; 9 estimation of the total energy released by the explosion. Because of the continuous presence of seismic background noise, only explosions whose yield exceeds a certain value are detected. This detectability limit varies with the distance from the explosion point and increases with the amplitude of the background noise. In the case of the Pacific Test Centre, the station at Rarotonga (Cook Islands), located at approximately 2000 km on a direct uninterrupted line from Mururoa, receives the T waves emitted by weak explosions and thus has a detectability close to one kiloton. The other stations are much further away and, apart from a few particularly sensitive stations such as the Yellowknife network in Canada, they have a much higher detectability limit. The discrimination between explosion and earthquake is generally based on the shape difference between the observed seismic signals. The seismic sources associated with explosions and with earthquakes have very different characteristics. The explosions are point sources in space (a few hundred metres) and in time (a few tenths of a second), whereas earthquakes result from shifting of faults which can be as long as several tens of kilometres. The explosions are essentially superficial (depths of less
Comprehensive Test Ban Trea~
651
than 2 km), whereas the depths of earthquakes range from a few kilometres to more than 600 km. These differences affect the seismic signals: near-complete absence of surface waves for explosions, whereas they are strong for superficial earthquakes; raising of the ground for the compression waves generated by an explosion, while for an earthquake the initial movement depends on the position of the station with respect to the fault that shifted; duration of the compression waves much shorter for explosions than for earthquakes, etc. However, this recognition is not always very simple, particularly for events for which the signal-to-noise ratio is low. The location of a seismic event, explosion or earthquake, is calculated from the chronometry of the seismic waves received by many stations surrounding the epicentre as well as possible. The primary network station is made up of 50 stations (see Table 13.1). It transmits data continuously to the International Data Center in Vienna and it is made up of three-component stations and array stations. Three-component stations are seismic stations with sensors that measure the three components (one vertical and two horizontal) of the arriving waves from seismic events, including earthquakes and explosions. These data enable the location and the size of the seismic event to be determined. They are not as sensitive as array stations but are less expensive. Table 13.1 contains the designations of countries, places, latitudes and longitudes of those contained in the Comprehensive Nuclear-Test-Ban Treaty, which was adopted by the General Assembly of the United Nations on 10 September 1996. Array stations are sets of 9-25 geometrically arranged seismic sensors distributed over an area of up to 500 km 2. These arrays have an enhanced detection capability and accurately measure the direction of the source of an event and its distance. In addition there is an auxiliary network of stations (listed in Table 13.2) made up of 120 stations which provide information to the International Data Centre upon request to supplement the data from the primary network.
13.2.2 The Radionuclide Monitoring Network The Radionuclide Monitoring Network will be distributed over 80 sites to detect atmospheric nuclear tests and underground tests when venting occurs or when radioactive xenon escapes along natural ground fractures. Radionuclide aerosol sampler analyzers will be installed at all sites, and automated radioxenon sample analyzers will be installed at 40 of the sites, capable of detecting four xenon isotopes by the radioxenon detectors. The discrimination against reactor-produced radioxenon must be achieved. Both types of sensors will automatically analyze samples and report the results to the International Data Center. In addition, certified labs will be identified to which samples can be sent for further analysis. The particulate sensors will be highly sensitive to atmospheric tests (threshold lower than 1 kt) and will provide unambiguous evidence of a nuclear explosion, although sensor response time is slow and localization accuracy limited. For
Chapter 13
652
Table 13.1 Primary network of seismological stations State responsible
Location
Latitude
Longitude
Type of station
Argentina
Paso Flores
40.7 S
70.6W
three component
Australia
Alice Springs, NT Mawson, Antarctica Stephens Creek, SA Warramunga, NT
23.7 67.6 31.9 19.9
133.9 62.9 141.6 134.3
array three component three component array
Bolivia
La Paz
16.3 S
68.1 W
Brazil
Brasilia
15.6 S
48.0 W
three component
Canada
Lac du Bonnet, Man. Schefferville, Que. Yellowknife, NWT
50.2 N 54.8 N 62.5 N
95.9 W 66.8 W 114.6 W
three component three component array
Central African Republic
Bangui
05.2 N
18.4 E
three component
China
Hailar Lanzhou
49.3 N 36.1 N
119.7 E 103.8 E
Colombia
El Rosal
04.9 N
74.3 W
three component
C6te d'Ivoire
Dimbroko
06.7 N
04.9 W
three component
Egypt
Luxor
26.0 N
33.0 E
array
Finland
Lahti
61.4 N
26.1 E
array
France
Tahiti
17.6 S
149.6 W
Germany
Freyung
48.9 N
13.7 E
array
Iran (Islamic Republic of)
Tehran
35.8 N
51.4 E
three component
Japan
Matsushiro
36.5 N
138.2 E
array
Kazakhstan
M akanchi
46.8 N
82.0 E
array
Kenya
Kilimambogo
01.1 S
37.2 E
three component
Mongolia
Javhlant
48.0 N
106.8 E
Niger
(new site)
(to be determined)
array
Norway
Hamar Karasjok
60.8 N 69.5 N
10.8 E 25.5 E
array array
Pakistan
Pari
33.7 N
73.3 E
array
Paraguay
Villa Florida
26.3 S
57.3 W
three component
Republic of Korea
Wonju
37.5 N
127.9 E
array
Russian Federation
Khabaz Norilsk Peleduy Petropavlovsk-Kamchats kiy Ussuriysk Zalesovo
43.7 69.0 59.6 53.1
42.9 88.0 112.6 157.8
three component three component array array
S S S S
N N N N
44.2 N 53.9 N
E E E E
E E E E
132.0 E 84.8 E
three component
array array
three component
array
array array
Comprehensive Test Ban Treaty
State responsible
Location
653
Latitude
Longitude
Type of station
Saudi Arabia
(new site)
(to be determined)
array
South Africa
Boshof
28.6 S
25.6 E
three component
Spain
Sonseca
39.7 N
04.0 W
array
Thailand
Chiang Mai
18.8 N
99.0 E
array
Tunisia
Thala
35.6 N
08.7 E
three component
Turkey
Belbashi (subject to relocation at Keskin)
39.9 N
32.8 E
array
Turkmenistan
Alibeck
37.9 N
58.1 E
array
Ukraine
Malin
50.4 N
29.1 E
array
United States of America
Eielson, AK Lajitas, TX Mina, NV Pinedale, WY Vanda, Antarctica
64.8 29.3 38.4 42.8 77.5
N N N N S
146.9 W 103.7W 118.2 W 109.6 W 161.9 E
array array array array three component
well-continued underground nuclear tests, the radioxenon detectors will be the key technology, assuming sufficient xenon escapes into the atmosphere. The particle size of a fission aerosol, and the distribution of fission products between particulate and vapour phases, depends on the mechanism of release to the atmosphere. In a weapons explosion, some physicochemical fractionation of radionuclides may occur, particularly if the explosion is near the ground. Everything in the vicinity is vaporised by the heat of the explosion, but within less than a minute the fireball cools to a temperature in the range 1000-2000~ and refractory materials such as metal oxides and silicates condense to form particles. Refractory fission products, and plutonium, are incorporated in these particles. Less-refractory fission products condense later onto the surface of the particles. Those with gaseous precursor, for example 9~ ad 137Cs, condense as they are formed by decay of their parent nuclides. Let us mention that at distances of about 100 km from ground zero of the Nevada tests, the particle size of fallout ranged from a few ~tm to a few 100 ~tm. Fission products with volatile precursors were enhanced by about a factor two compared with refractory fission products. The fractionation was greater, the smaller the particles. When H-bombs are exploded at high altitude no ground-based material is incorporated, so the condensation aerosol particles are very small, with 90% of the activity in particles less than 0.3 ~tm in diameter. Coagulation takes place with the natural stratospheric aerosol, which is thus labelled with fission products. In the troposphere there is further coagulation, and near ground most activity is in the 0.3-1 ~tm size range.
Chapter 13
654
Table 13.2 Auxiliary network of seismological stations State responsible
Location
Latitude
Longitude
Type of station
Argentina
Coronel Fontana Ushuaia
31.6 S 55.0 S
68.2W 68.0 W
three component three component
Armenia
Garni
40.1 N
44.7 E
three component
Australia
Charters Towers, QLD Fitzroy Crossing, WA Narrogin, WA
20.1 S 18.1 S 32.9 S
146.3 E 125.6 E 117.2 E
three component three component three component
Bangladesh
Chittagong
22.4 N
91.8 E
three component
Bolivia
San Ignacio
16.0 S
61.1 W
three component
Botswana
Lobatse
25.0 S
25.6 E
three component
Brazil
Pitinga Rio Grande do Norte
0.7 S 6.9 S
60.0 W 37.0 W
three component three component
Canada
Bella Bella, BC Dease Lake, BC Inuvik, NWT Iqaluit, NWT Mould Bay, NWT Sadowa, Ontario
52.2 58.4 68.3 63.7 76.2 44.8
N N N N N N
128.1 W 130.0 W 133.5 W 68.5 W 119.4 W 79.1 W
three three three three three three
Chile
Easter Island Lemon Verde
27.2 S 22.6 S
109.4 W 68.9 W
three component three component
China
Baijiatuan Kunming Sheshan Xi'an
40.0 25.2 31.1 34.0
116.2 102.8 121.2 108.9
three three three three
Cook Islands
Rarotonga
21.2 S
159.8 W
three component
Costa Rica
Las Juntas de Abangares
10.3 N
85.0 W
three component
Czech Republic
Vranov
49.3 N
16.6 E
three component
N N N N
E E E E
component component component component component component
component component component component
Denmark
S~ndre Str~mfjord, Greenland 67.0 N
50.6 N
three component
Djibouti
Arta Tunnel
11.5 N
42.9 E
three component
Egypt
Kottamya
29.9 N
31.8 E
three component
Ethiopia
Furi
8.9 N
38.7 E
three component
Fiji
Monasavu, Viti Levu
17.8 S
178.1 E
three component
France
Kourou, French Guiana 5.2 N Port Laguerre, New Caledonia 22.1 S
52.7 W 166.3 E
three component three component
Gabon
Bambay
1.7 S
13.6 E
three component
Germany/S. Africa SANAE Station, Antarctica
71.7 S
2.9W
three component
Greece
Anogia, Crete
35.3 N
24.9 E
three component
Guatemala
Rabir
15.0 N
90.5 W
three component
Iceland
Borgarnes
64.8 N
21.3 W
three component
Comprehensive Test Ban Treaty
655
State responsible
Location
Latitude
Longitude
Type of station
Indonesia
Cibinong, Jawa Barat Jayapura, Irian Jaya Kappang, Sulawesi Selatan Kupang, Nusatenggara Timur Parapat, Sumatera Sorong, Irian Jaya
6.5 S 2.5 S 5.0 S 10.2 S 2.7 N 0.9 S
107.0 E 140.7 E 119.8 E 123.6 E 98.9 E 131.3 E
three component three component three component three component three component three component
Iran (Islamic Republic of)
Kerman Masjed-e-Soleyman
30.3 N 31.9 N
57.1 E 49.3 E
three component three component
Israel
Eilath Parod
29.8 N 32.6 N
34.9 E 35.3 E
three component array
Italy
Enna, Sicily
37.5 N
14.3 E
three component
Japan
Chichijima, Ogasawara Hachijojima, Izu Islands Kamikawa-asahi, Hokkaido Kunigami, Okinawa Ohita, Kyushu
27.1 33.1 44.1 26.8 33.1
142.2 139.8 142.6 128.3 130.9
three component three component three component three component three component
Jordan
Ashzof
32.5 N
37.6
three component
Kazakhstan
Aktyubinsk Borovoye Kurchatov
50.4 N 53.1 N 50.7 N
58.0 E 70.3 E 78.6 E
three component array array
N N N N N
E E E E E
Kyrgyzstan
Ala-Archa
42.6 N
74.5 E
three component
Madagascar
Antananarivo
18.9 S
47.6 E
three component
Mali
Kowa
14.5 N
4.0 W
three component
Mexico
La Paz, Baja California Sur Tepich, Yucatan Tuzandepeti, Veracruz
24.2 N 20.2 N 18.0 N
110.2 W 88.3 W 94.4 W
three component three component three component
Morocco
Midelt
32.8 N
4.6 W
three component
Namibia
Tsumeb
19.1 S
17.4 E
three component
Nepal
Everest
28.0 N
86.8 E
three component
New Zealand
Erewhon, South Island Raoul Island Urewera, North Island
43.5 S 29.2 S 38.3 S
170.9 E 177.9 W 177.1 E
three component three component three component
Norway
Jan Mayen Spitsbergen
70.9 N 78.2 N
8.7 W 16.4 E
three component array
Oman
Wadi Sarin
23.0 N
Papua New Guinea Bialla Port Moresby
58.0 E
three component
5.3 S 9.4 S
151.1 E 147.2 E
three component three component
Peru
Cajamarca Nana
7.0 S 12.0 S
78.0 W 76.8 W
three component three component
Philippines
Davao, Mindanao Tagaytay, Luzon
7.1 N 14.1 N
125.6 W 120.9 E
three component three component continued
Chapter 13
656
Table 13.2 (continuation) State responsible
Location
Latitude
Longitude
Type of station
Romania
Muntele Rosu
45.5 N
25.9 E
three component
56.4 68.0 58.6 44.0 59.6 55.1 62.9 51.7 71.6 51.1 62.0 47.0 53.9
58.6 E 166.4 E 49.4 E 42.7 E 150.8 E 36.6 E 152.4 E 103.6 E 128.9 E 132.3 E 129.7 E 142.8 E 57.0 E
three component three component three component array three component three component three component three component three component three component three component three component three component
Russian Federation Arti Bilibino Kirov Kislovodsk Magadan Obninsk Seymachan Talaya Tiksi Urgal Yakutsk Yuzhno-Sakhalinsk Zilim
N N N N N N N N N N N N N
Afiamalu
13.9 S
171.8 W
three component
Saudi Arabia
Ar Rayn
23.6 N
45.6 E
three component
Senegal
Mbour
14.4 N
17.0 W
three component
Solomon Islands
Honiara, Guadalcanal
9.4 S
160.0 E
three component
South Africa
Sutherland
32.4 S
20.8 E
three component
Sri Lanka
Colombo
6.9 N
79.9 W
three component
Sweden
Hagfors
60.1 N
13.7 E
array
Switzerland
Davos
46.8 N
9.8 E
three component
Uganda
Mbarara
0.4 S
30.4 E
three component
United Kingdom
Eskdalemuir
55.3 N
3.2 W
array
United States of America
Albuquerque, NM Attu Island, AK Elko, NV Guam, Marianas Islands Kodiak Island, AK Newport, WA Palmer Station, Antarctica Pifion Flat, CA San Juan, PR South Pole, Antarctica Tuckaleechee Caverns, TN Yreka, CA
35.0 52.8 40.7 13.6 57.8 48.3 64.8 33.6 18.1 90.0 35.7 41.7
106.5 W 172.7 E 115.2 W 144.9 E 152.5 W 117.1 W 64.1 W 116.5 W 66.2 W J 83.8 W 122.7 W
three component three component three component three component three component three component three component three component three component three component three component three component
Venezuela
Puerto la Cruz Santo Domingo
10.2 N 8.9 N
64.6 W 70.6 W
three component three component
Zambia
Lusaka
15.3 S
28.2 E
three component
Zimbabwe
Bulawayo
(to be advised)
Samoa
N N N N N N S N N N N N
three component
Comprehensive Test Ban Trea~
657
Underground nuclear explosions produce trace amounts of distinctive but ephemeral radionuclide gases. In the context of monitoring a comprehensive test ban treaty, the detection of these gases within the territory of a signatory, during a challenge inspection, may indicate the occurrence of a clandestine nuclear event (Hannon, 1985; Zuckerman, 1993; Drell and Peurifoy, 1994; Zucca, 1994; Carrigan, 1994). In the paper by Carrigan et al. (1994) the results of an experiment simulating a well-contained underground nuclear explosion, undertaken to test the ability of natural gas-transport processes to move highly dilute and rapidly decaying radionuclides to the surface have been reported. They found that trace gases are transported to the surface within periods of weeks to a year, by flow along faults and fractures driven by barometric pressure variations. Both their observations and related simulations exhibited a chromatographic behaviour, with gases of higher atomic mass and lower diffusivity reaching the surface more rapidly. For a 1-kilotonne nuclear test under conditions identical to those of the experiment, they predicted that short-lived 133Xeand 3VArwould be detectable, respectively, about 50 and 80 days after the detonation. Their results indicated that radionuclides sampling along natural faults and fractures, as a forensic tool, could be an extremely sensitive way to detect nearby underground nuclear explosions that do not fracture the surface. If full reliance is to be placed on the seismic and hydroacoustic network for detecting underground and underwater tests, so that the radionuclide detection system is targeted only at atmospheric or surface tests, then it is possible to specify fairly precisely the type of detection capacity needed. Very high volume air samplers could remove particulate radionuclides from the air and check for short-lived radionuclides and their ratios (signatures) with local high sensitivity gamma spectrometers. This technology is relatively well-proven and development of a network is unlikely to produce major problems. The minimum detection limits for airborne particulate radioactivity are indicated in Table 13.3. For greater reliance on the radionuclide monitoring system for the detection of underground and underwater tests, more sophisticated systems are required. If the test is poorly conducted with substantial venting to atmosphere, then the task is much the same as for a surface test but with a probably somewhat lower detection. If the test is so controlled that essentially all the particulates are retained in the ground or in water, then it will be necessary to monitor for those gases, in particular noble (inert) gases, that are likely to escape. This may also be necessary if an atmospheric test is carried out under conditions of high precipitation such that atmospheric particulates are essentially washed out. Suitable noble gas nuclides for detection would be those with half-lives of a few days up to a few weeks (e.g. argon-37: half-life 35 days; xenon-133: half-life 5.2 days; xenon-135: half-life 9.1 hours), which would indicate recent fission. Table 13.4 shows, as an example, the sources of different radioisotopes of xenon. Noble gases with longer half-lives also might be of some use. However, krypton-85 (half-life 10.7 years), for example, would have to be counted against a worldwide background caused by its routine release in the reprocessing of spent reactor fuel. In any case, the sophisticated techniques available for measuring noble gases (perhaps
Chapter 13
658 Table 13.3 Minimum detection limits for airborne particulate radioactivity Nuclide
Half-life
Detection limit (~tBq/m3)
24Na
15.020 h 83.83 d 27.704 d 312.5 d 78.76 d 270.9 d 70.80 d 44.529 d 5.271 y 243.9 d 14.10 h 119.77 d 26.32 h 35.30 h 106.60 d 58.51 d 63.98 d 86.6 h 35.15 d 16.90 h 66.0 h 39.35 d 35.36 h 371.63 d 249.9 d 7.45 d 49.51 d 44.6 d 53.46 h 16.78 d 2.70 d 119.7 d 60.20 d 9.64 d 2.73 y 12.4 d 3.85 d 33.6 d 30. h 8.04 d 11.8 d 78.2 h 20.8 h 10.74 y
280 0.03 0.68 0.03 0.03 0.05 0.03 0.06 0.03 0.06 460 0.10 15. 1.6 0.03 13. 0.05 1.1 0.03 80. 0.40 0.03 16 0.27 0.04 1.7 0.65 1.5 1.5 0.04 0.33 0.05 0.03 0.64 0.15 0.09 0.32 1.2 14 0.12 0.10 0.24 27 0.08
465c 51Cr 54Mn 56Co 57Co 58Co 59Fe 6~ 65Zn 72Ga 75Se 76As 82Br 88y 91y 95Zr 95mNb 95Nb 97Zr 99Mo l~ l~ l~ 11~ 11lAg l lamln 115mCd ll5cd 121Te 122Sb 123mTe 124Sb 125Sn 125Sb 1265b 127Sb 129mTe 131mTe 1311 131Ba 132Te 1331 133Ba
659
Comprehensive Test Ban TreaO;
Nuclide
Half-life
Detection limit (~tBq/m3)
134Cs
2.062 y 13.16 d 30.0y 137.66 d 12.746 d 40.272 h 32.50 d 1.9.13 h 33.0 h 284.9 d 10.98 d 13.33 y 46.7 h 241.6 d 8.8 y 4.96 y 15.19 d 18.56 h 144.4 d 72.3 d 26.80 h 93.1 d 32.022 d 128.6 d 4.19 d 70. 6.71 d 160.9 d 4.4 d 115.0 d 93.6 d 90.64 h 23.9 h 16.98 h 15.4 d 2.9 d 74.02 d 30.5 h 171. d 6.183 d 64.1 h 23.8 h 2.696 d 46.60 d 6.75 d 2.355 d 432.2 y
0.03 0.07 0.03 0.06 0.16 0.75 0.10 980 11 0.38 0.33 0.18 3.4 0.17 0.07 0.21 0.42 790 8.5 0.12 190 0.05 0.18 3.1 2.8 0.06 1.2 0.27 0.13 0.09 0.03 2.4 30.4 960 0.24 1.5 0.07 180 0.41 0.14 3.6 52. 0.43 0.09 0.59 5.4 0.44
136Cs 137Cs 139Ce 14~ 14~ 141Ce 142pr 143Ce 144Ce 147Nd 152Eu 153Sm 153Gd 154Eu 155Eu 156Eu 159Gd 159Dy 16~ 166Ho 168Tm 169yb 17~ 175yb 175Hf 177Lu 177mLu 181Hf 182Ta 185Os 186Re 187W 188Re 191Os 191pt 192Ir 193Os 194mlr 196Au 197Hg 197mHg 198Au 2~ 237U 239Np 241Am
660
Chapter 13
Table 13.4 Primary potential sources of radioxenons 131mxe Weapons tests
133mxe
133Xe
135Xe
X
X
X
Reactor operations Fuel reprocessing Medical isotopes
X
X
using some sort of cryogenic gas chromatography) entail greater complexity and cost, and careful feasibility and cost-benefit studies would be necessary. Access to data that are likely to be restricted would also be necessary. Such data would include quantification of the venting of specific gases released from tests of various types in different host rocks as well as information on advanced procedures for gas collection and analysis. The surface area of the Earth is some 500 million km 2, implying that each station should cover on average some 6 million km 2. Distances from a test could thus be up to a few thousand kilometres, over which there is a dilution of up to some 4 or 5 orders of magnitude (i.e. ten to a hundred thousand times) in air concentration of radionuclides. Together with the greater dilution, there is also a corresponding dispersion, which would increase the probability of radionuclides being detected at more than one site: a desirable checking feature to increase confidence in the system. In theory, using more sophisticated monitors to achieve lower detection limits could reduce the theoretical number of stations needed to detect a test. However, since the mean distance between the stations would become larger, there would be a loss in the ability to identify the region within which a test might have occurred. Thus any detailed modelling studies on a monitoring network would need guidance as to the requirement of the system to identify the potential source as opposed to merely detecting a transgression. Since most of the Earth's surface is water, in developing a global network, all else being equal, the majority of the monitoring stations ought to be based in ocean locations. Consideration needs to be given to the feasibility of this and to whether these should be fixed or mobile. In order to cover adequately some areas of the ocean, ships mounted with detection capabilities would be desirable. Their mobility could have advantages in increasing the likelihood of detection in the event of a test, with potential savings in the number of detectors needed. Radionuclide stations sample large volumes of air to detect radioactive particles and noble gases released from atmospheric nuclear explosions and radioactive gases vented from underground and underwater nuclear explosions. These radionuclides can be carried great distances by winds. The presence and ratios of specific radionuclides provide evidence of a nuclear explosion. The major components of an air sampling system for detecting radionuclides on particles include at least the following items on-site:
Comprehensive Test Ban Treat 3,
661
1. a collecting system with a filter holder and a high capacity blower. The blower should be capable of drawing at least 500 cubic meters of air per hour (m3/h) through a filter paper. The filter should collect submicron particles with high efficiency, very near 100% and optimized for 0.1-1.0 ~tm particles. The filters should be approximately one quarter of a square meter in area; 2. equipment, like a small press, to prepare the filter samples for analysis; 3. a well shielded coaxial germanium detector for gamma counting of the filter samples. The detector should have an energy resolution of less than 2.2 keV FWHM (Full Width at Half Maximum) at 1332 keV gamma energy, should cover the energy range of at least 40 to 2000 keV, be very efficient (>80% as defined for germanium detectors by IEEE), and have very low background such as provided by a 10 cm thick lead shield appropriately lined. The detector should be cooled and use only ultra low background component; 4. appropriate electronics to process the signals from the detectors. The data processing electronics should provide stable, digitized data in at least 4096 channels. It should be possible to check and adjust these units remotely over a telephone or a satellite link from the IDC; 5. a computer, desktop size or smaller, should be used to run the counting procedures, store raw data, handle the data transfer to the IDC and also to oversee and communicate important parameters in the system like weather data, indoor temperature, the condition of the electric power system and data from sensors suggested under (14) and (15) below; 6. a satellite communication system to transmit the data electronically and reliably from the site to the IDC in less than one hour. In normal operation there should be at least one message per day consisting of less than 100 kB of data each. On request the shipment of the filter samples to the IDC should be accomplished within three weeks. On request the shipment of a sample to a designated laboratory for a comprehensive analysis should be expedited immediately; 7. a small meteorological set including sensors for wind direction, wind speed, humidity and temperature. The xenon sampling system would share the main computer and communication gears with the filter system but the following items are needed specifically: 8. columns and cold traps to take water and carbon dioxide away from the air stream going into the sampler; 9. a low temperature freezer (-77~ where the noble gases are collected in a copper coil filled with activated charcoal (air flow 0.5-1 m3/h); 10. an oven (275~ to drive the xenon out from the sampling coil for further processing; 11. a system to significantly reduce the volume of the gas sample; 12. a gas chromatograph to separate the xenon (and possibly in the future krypton) from disturbing radon gas. The chromatogram is also used to quantify the
662
Chapter 13
radioactive xenon isotopes as natural stable xenon, with its known atmospheric concentration, is dominating the pulse; 13. a well shielded planar germanium detector with associated electronics to measure the xenon isotopes in a small glass coil filled with activated charcoal. The detector should have a sensitive area of at least 20 cm 2 and a sensitive depth around 10 mm. Alternatively, the 13-7 coincidence system could be used as a detector. The system on-site could preferably also include: 14. a gamma dose rate meter capable of measuring ambient radiation levels and higher levels in an accident situation; 15. a separate radon gas monitor. The station should be capable of near automatic, autonomous operation, and require a minimum of local support. Work envisioned at a manual station is to change samples daily, prepare the sample for analysis (like pressing a filter into a disk) and then put it on the detector and push a button. This can be done in less than an hour each time. In areas where line power is unreliable and of bad quality a diesel generator to provide power would, however, probably require more hours for service, fuel filling and maintenance. Full automatic stations have also been designed. As already mentioned above, each station should collect samples daily. This gives a good chance of picking up fairly short-lived radionuclides and gives also short enough sampling periods that in most cases (depending on how complex weather systems are prevailing) allow a fairly good analysis of where to look for a source. As an illustration of the performance of an atmospheric radioactivity station let us mention the so-called a priori detection limits for the Grindsj6n station in the Swedish National Surveillance Network for airborne particulate radioactivity, shown in Table 13.3. These numbers express the smallest concentrations of single man-made radionuclides that one can expect to detect at Grindsj6n. The station to be used in the CTBT should have similar performance. Table 13.3 includes fusion products and neutron (and a few proton) activation products with half-lives no less than 12 hours. High detection limits, in the 10-1000 ~tBq/m 3 range, are mostly that high because the limits apply to the middle of the sampling period which in these cases is many half-lives before the counting process. 3 How much is one ~tBq/m3 in simpler words? One can translate it into atoms per m by multiplying with 0.5 (=24 x 3600 x 106/1n2) and the half-life expressed in days. For, e.g., 14~ (barium-140) this gives a detection limit close to one atom per cubic meter of air, which is one, in our case interesting, atom among almost 1026others. The network is made up of a total of 80 stations as shown in Table 13.5. In addition the system includes 16 radionuclide laboratories, as shown in Table 13.6, which provide on request the more detailed analyses to supplement the data from radionuclide stations. Some consideration must be given to protecting the monitoring system against breaches of security with intent to deceive. These considerations ought probably to
Comprehensive Test Ban Treat3'
663
Table 13.5 Radionuclide stations State responsible Location
Latitude
Longitude
Argentina
Bariloche Buenos Aires Salta
41.1 S 34.0 S 24.0S
71.3W 58.0 W 65.0 W
Australia
Cocos Islands Darwin, NT Macquarie Island Mawson, Antarctica Melbourne, VIC Perth, WA Townsville, QLD
12.0 12.4 54.0 67.6 37.5 31.9 19.2
S S S S S S S
97.0 E 130.7 E 159.0 E 62.5 E 144.6 E 116.0 E 146.8 E
Brazil
Recife Rio de Janeiro
8.0S 22.5 S
35.0 W 43.1W
Cameroon
Douala
4.2 N
9.9E
Canada
Resolute, NWT St. John's, NL Vancouver, BC Yellowknife, NWT
74.7 47.0 49.3 62.5
N N N N
94.9 W 53.0 W 123.2 W 114.5 W
Chile
Hanga Roa, Easter Island Punta Arenas
27.1 S 53.1 S
108.4 W 70.6W
China
Beijing Guangzhou Lanzhou
39.8 N 23.0 N 35.8 N
116.2 E 113.3 E 103.3 E
Cook Islands
Rarotonga
21.2 S
159.8 W
Ecuador
lsla San Crist6bal, Galapagos Islands
1.0 S
89.2 W
Ethiopia
Filtu
5.5 N
42.7 E
Fiji
Nadi
18.0S
177.5 E
France
Cayenne, French Guiana Dumont d'Urville, Antarctica Papeete, Tahiti Pointe-?a-Pitre, Guadeloupe Port-aux-Franqais, Kerguelen R6union
5.0N 66.0 S 17.0S 17.0 N 49.0 S 21.1 S
52.0 W 140.0 E 150.0 W 62.0 W 70.0 E 55.6 E
Germany
Schauinsland/Freiburg
47.9 N
7.9E
Iceland
Reykjavik
64.4 N
21.9 W
Iran (Islamic Republic of)
Tehran
35.0 N
52.0 E
Japan
Okinawa Takasaki, Gunma
26.5 N 36.3 N
127.9 E 139.0 E
Kiribati
Kiritimati
2.0N
157.0 W
Kuwait City
29.0 N
48.0 E
Kuwait
continued
Chapter 13
664
Table 13.5 (continuation) State responsible Location
Latitude
Longitude
Libyan Arab Jamahiriya
Misratah
32.5 N
15.0 E
Malaysia
Kuala Lumpur
2.6 N
101.5 E
Mauritania
Nouakchott
18.0 N
17.0 W
Mexico
Baja California
28.0 N
113.0 W
Mongolia
Ulaanbaatar
47.5 N
107.0 E
New Zealand
Chatham Island Kaitaia
44.0 S 35.1 S
176.5 W 173.3 E
Niger
Bilma
18.0 N
13.0 E
Norway
Spitsbergen
78.2 N
16.4 E
Panama
Panama City
8.9 N
79.6 W
Papua New Guinea
New Hanover
3.0 S
150.0 E
Philippines
Quezon City
14.5 N
121.0 E
Portugal
Ponta Delgada, Sao Miguel, Azores
37.4 N
25.4 W
Russian Federation
Bilibino Dubna Kirov Norilsk Peleduy Petropavlovsk-Kamchatskiy Ussuriysk Zalesovo
68.0 56.7 58.6 69.0 59.6 53.1 43.7 53.9
166.4 E 37.3 E 49.4 E 88.0 E 112.6 E 158.8 E 131.9 E 84.8 E
South Africa
Marion Island
46.5 S
N N N N N N N N
37.0 E
Sweden
Stockholm
59.4 N
18.0 E
Tanzania
Dares Salaam
6.0 S
39.0E
Thailand
Bangkok
13.8 N
100.5 E
United Kingdom
British Indian Ocean Territory/Chagos Archipelago Halley, Antarctica St. Helena Tristan da Cunha
7.0 S 76.0 S 16.0 S 37.0 S
72.0 E 28.0 W 6.0 W 12.3 W
United States of America
Ashland, KS Charlottesville, VA Melbourne, FL Midway Islands Oahu, HI Palmer Station, Antarctica Sacramento, CA Salchaket, AK Sand Point, AK Upi, Guam Wake Island
37.2 N 38.0 N 28.3 N 28.0 N 21.5 N 64.5 S 38.7 N 64.4 N 55.0 N 13.7 N 19.3 N
99.8 W 78.0 W 80.6 W 177.0 W 158.0 W 64.0 W 121.4 W 147.1 W 160.0 W 144.9 E 166.6 E
Comprehensive Test Ban Treaty
665
Table 13.6 Radionuclide laboratories No
State responsible
Place
1 2
Argentina Australia
Buenos Aires Melbourne,VIC
3
Austria
Seibersdorf
4 5
Brazil Canada
Rio de Janeiro Ottawa,Ontario
6
China
Beijing
7 8
Finland France
Helsinki Montlhrry
9
Israel
Yavne
10 11
Italy Japan
Rome Tokai,Ibaraki
12
New Zealand
Christchurch
13 14
Russian Federation South Africa
Moscow Pelindaba
15 16
United Kingdom United States of America
Chilton Sacramento,CA
include: (1) interruptions to power supplies of monitoring stations; (2) interference affecting information transmitted from stations to IDC; (3) decoy releases of radioactive materials from nuclear installations in order to mask releases from a test; (4) potential corruption among personnel operating the monitoring stations; (5) degradation of equipment and supplies. Any system should be designed in consideration of these and other issues. In particular, consideration should be given to the need for "independent" monitoring stations, particularly in the territory of continental-sized countries and/or allied countries, and in international waters.
13.2.3 The Hydroacoustic Monitoring Network The Hydroacoustic Monitoring Network will consist of six underwater hydrophone stations and five island-based stations (T-phase stations). The hydrophones will detect signals from underwater explosions that readily propagate to great distances in the ocean sound channel. The less sensitive but much less expensive island stations will detect seismic signals that are generated by a hydroacoustic wave when it strikes an island. Sensitivity in the broad ocean areas (a likely place for a clandestine test) is expected to be far less than 1 kt, with localization to within 1000 km 2 and with a source depth estimate. These capabilities are for events detected by three or more stations. Hydrophone stations use underwater microphones (hydrophones) attached to cables anchored to the seabed 50-100 km offshore and connected by cable to a land
666
Chapter 13
Table 13.7 Hydroacoustic stations State responsible
Location
Latitude
Longitude
Type of station
Australia Canada Chile France
Cape Leeuwin, WA Queen Charlotte Islands, BC Juan Fernandez Island Crozet Islands Guadeloupe Clarion Island Flores BritishIndian Ocean Territory/Chagos Archipelago Tristan da Cunha Ascension Wake Island
34.4 S 53.3 N 33.7 S 46.5 S 16.3 N 18.2 N 39.3 N
115.1 E 132.5 W 78.8 W 52.2 E 61.1 W 114.6 W 31.3 W
hydrophone T-phase hydrophone hydrophone T-phase T-phase T-phase
7.3 S 37.2 S 8.0 S 19.3 N
72.4 E 12.5 W 14.4 W 166.6 E
hydrophone T-phase hydrophone hydrophone
Mexico Portugal United Kingdom
United States of America
facility for data transmission. They are extremely sensitive and pick up acoustic waves from underwater events, including explosions, vast distances away. T-phase stations are located on oceanic islands and use seismometers to detect the acoustic waves generated by explosions in the oceans. These stations are not as sensitive as hydrophone stations but are less expensive. Altogether there are 11 hydroacoustics stations, their location being shown in Table 13.7.
13.2.4 The Infrasound Monitoring Network The Infrasound Monitoring Network will consist of 60 stations, each equipped with an array of four microbaragraphs--three located at the vertices of a 1 km equilateral triangle and the fourth at the center of the triangle. The operation frequency range will be 0.1-16 Hz, with digital recording, sophisticated signal processing and automatic reporting. The detected signals are the low-frequency components of the strong shock wave produced by an atmospheric burst and can be detected up to a few thousand kilometres away by virtue of the "wavequide" created by reflection from Earth and refraction from the atmosphere for infrasound. The infrasound network is expected to be able to detect atmospheric explosions in the 1 kt range worldwide, with regional and seasonal atmospheric conditions determining. Infrasound stations use very sensitive microbarometers to detect low frequency sound waves from atmospheric explosions. These stations are arrays and are able to determine accurately the location and size of an atmospheric explosion. Altogether there are 60 stations in the network as shown in Table 13.8.
Comprehensive Test Ban TreaO,
667
Table 13.8 Infrasound stations State responsible
Location
Latitude
Argentina
Paso Flores Ushuaia
40.7 S 55.0 S
Australia
Cocos Islands Davis Base, Antarctica Hobart, TAS Narrogin, WA Warramunga, NT
12.3 68.4 42.1 32.9 19.9
Bolivia
La Paz
16.3 S
68.1 W
Brazil
Brasilia
15.6 S
48.0 W
Canada
Lac du Bonnet, Man.
50.2 N
95.9 W
Cape Verde
Cape Verde Islands
16.0 N
24.0 W
Central African Republic
Bangui
05.2 N
18.4 E
Chile
Easter Island Juan Fernandez Island
27.0 S 33.8 S
109.2 W 80.7 W
China
Beijing Kunming
40.0 N 25.0 N
116.0 E 102.8 E
C6te d'Ivoire
Dimb6kr6
06.7 N
04.9 W
Denmark
Dundas, Greenland
76.5 N
68.7 W
Djibouti
Djibouti
11.3 N
43.5 E
Ecuador
Galapagos Islands
0.0 N
91.7 W
France
Kerguelen Kourou, French Guiana Marquesans Islands Port LaGuerre, New Caledonia Tahiti
49.2 5.2 10.0 22.1 17.6
Germany
Freyung Georg von Neumayer, Antarctica
48.9 N 70.6 S
13.7 E 8.4 W
Iran (Islamic Republic of)
Tehran
35.7 N
51.4 E
Japan
Tsukuba
36.0 N
140.1 E
Kazakhstan
Aktyubinsk
50.4 N
58.0 E
Kenya
Kilimanbogo
01.3 S
36.8 E
Madagascar
Antananarivo
18.8 S
47.5 E
Mongolia
Javhlant
48.0 N
106.8 E
Namibia
Tsumeb
19.1 S
17.4 E
S S S S S
S N S S S
Longitude 70.6 W 68.0 W 97.0 77.6 147.2 117.2 134.3
69.1 52.7 140.0 166.3 149.6
E E E E E
E W W E W
continued
Chapter 13
668
Table 13.8 (continuation)
State responsible
Location
Latitude
Longitude
New Zealand
Chatham Island
44.0 S
176.5 E
Norway
Karasjok
69.5 N
25.5 E
Pakistan
Rahimyar Khan
28.2 N
70.3 E
Palau
Palau
7.5 N
134.5 E
Papua New Guinea
Rabaul
4.1 S
152.1 E
Paraguay
Villa Florida
26.3 S
57.3 W
Portugal
Azores
37.8 N
25.5 W
Russian Federation
Dubna Petropavlovsk-Kamchatskiy Ussuriysk Zalesovo
56.7 53.1 43.7 53.9
N N N N
37.3 158.8 131.9 84.8
E E E E
South Africa
Boshof
28.6 S
25.4 E
Tunisia
Thala
35.6 N
08.7 E
United Kingdom
Ascension Bermuda British Indian Ocean Territory/ Chagos Archipelago Tristan da Cunha
8.0 S 32.0 N 5.0 S
14.3 W 64.5 W 72.0 E
37.0 S
12.3 W
Eielson, AK Hawaii, HI Midway Islands Newport, WA Pifion Flat, CA Siple Station, Antarctica Wake Island Windless Bight, Antarctica
64.8 N 19.6 N 28.1 N 48.3 N 33.6 N 75.5 S 19.3 N 77.5 S
146.9 W 155.3 W 177.2 W 117.1 W 116.5 W 83.6 W 166.6 E 161.8 E
United States of America
13.3 INTERNATIONAL DATA CENTER (IDC) The International Data Center (IDC) of the CTBT organization functions as the nerve center of the CTBT's verification system. If an anomaly is detected in the results of a routine air sample and transmitted to the International Data Centre from an air monitoring station, then several steps would follow in a response phase. These steps are in many ways analogous to the procedures followed by the IAEA in operating its Emergency Response Unit (ERU) under the convention on Early Notification of a Nuclear Accident and the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency.
Comprehensive Test Ban TreaO'
669
The data received would be rapidly confirmed and authenticated with the monitoring station itself, to rule out an illicit attempt to trigger a response. The information would be cross-checked against data as available from the seismic and/or hydroacoustic systems and the ratios of radionuclides present in the spectrum ("signatures") would be examined in order to establish whether the information were evidence of a test or indicated some other nuclear source. This requires skill and experience as well as current knowledge of nuclear plant management. Back trajectories of the plume would be rapidly evaluated in order to bound an area within which the source would be likely to lie. This could be co-ordinated through the World Meteorological Organization (WMO) and specialized national meteorological services. Moreover, it would also be advantageous to calculate forward trajectories from known nuclear installations in the region and to seek confirmation directly with the installations, i.e. to ask whether indeed there had been an anomaly in their operations which could explain the monitoring results. The IAEA has considerable experience in this field. Its ERU has established and developed an extremely close working relationship with WMO whereby, in the event of notification of an accident or other event, standardized meteorological products based on key parameters of the event are automatically requested. Designated WMO centres rapidly respond to make projections of the future extent of any radioactive plume and to update these as further information becomes available. Seismic, radionuclide, hydroacoustic, and infrasonic are the technologies that are combined into a unified monitoring system. The objectives of this system are to detect, identify, and locate a nuclear weapons test of 1-kt and below from any arbitrary point on the earth. Design of an international radionuclide monitoring system is predicated upon assessing meteorological and radiological influences. Atmospheric conditions which drive the transport of airborne debris include advection, deposition, dry removal, wet scavenging, vertical temperature profile, mixing layer dynamics, and gravitational setting. A critical part for the functioning of IDC is the Global Communications Infrastructure (GCI). The GCI will support the transmission of raw data from the 337 facilities of the International Monitoring System (IMS) to the IDC in Vienna, and the distribution of data and IDC products to State Signatories, primarily through their national data centres (NDCs). The GCI will be the world's first, global VSAT-based satellite system, linking stations and NDCs in populated and quite remote areas of all continents and island sites. On 7 September 1998, the CTBTO PrepCom and the international partnership, HOT (Hughes Olivetti Telecom Ltd.) signed a 10-year contract to establish this historic system. It is expected that the first 30 sites will be providing data to the IDC by the second quarter of 1999. By the time of full CTBTO operations, the network will have to provide effective, error-free transport of over 10 gigabytes of data daily, with delays of just seconds. It will also have to operate for 99.5 per cent of the time and be capable of functioning over a broad range of environmental conditions. The network will deploy the latest communications technology and will itself be continuously monitored and adjusted to ensure continuous high performance and protection from unauthorized access.
670
Chapter 13
There are a number of benefits based on the Treaty which result from the national implementation of verifications measures: 1. State Signatories will have access to data, through the IDC, that will serve many purposes other than monitoring the non-occurrence of a nuclear explosion. In particular, this data may be of great use for natural hazard mitigation; 2. The IDC will provide State Parties with open, equal, timely and convenient access to all IMS data, raw or processed, all IDC products, and all other IMS data in the archive of the IDC or, through the IDC, of IMS facilities. The IDC will also assist State Parties to identify the source of specific events (Protocol Part I, paragraph 20); 3. The IDC shall, where required, provide technical assistance to individual State Parties (Protocol Part I, paragraph 22), in formulating their requirements for selection and screening of data and products and developing the capability to receive, process and analyze IMS data at their national data centres; 4. The IDC will also support individual State Signatories by installing at the IDC, at no cost to a requesting State Party, specific computer algorithms or software provided by the State Party; and 5. The IDC will provide unique scientific and technical training to personnel from State Parties. Upon returning to their country of origin, these personnel would be available to support national missions, e.g. earthquake and environmental monitoring, computer hardware and software industry, etc.
13.4 O N - S I T E INSPECTIONS Article IV of the CTBT not only provides for remote monitoring, but also for on-site inspections (OSI) when other means fail to clarify a suspect activity. Any state that is party to the treaty may request such an inspection and may cite additional evidence, including that collected using its "National Technical Means", in support of its request. "The sole purpose of OSI shall be to clarify whether a nuclear weapon test explosion or any other nuclear explosion has been carried out in violation of article I" and, to the extent possible, to identify any possible violator. The requesting State Party shall keep the request within the scope of this Treaty and shall refrain from unfounded or abusive inspection requests. "The OSI request shall be based on information collected by the IMS, on any relevant technical information obtained by national technical means (NTM) of verification in a manner consistent with generally recognised principles of international law, or on a combination thereof." (Treaty, article IV Verification, paragraphs 34-37). An inspection requires approval from the 51-state Executive Council by at least 30 votes, with the vote taken no later than 96 hours after receipt of the initial request. The inspection party shall then arrive at the inspection site no later than nine days following the initial request for an inspection. The activities of the inspection team are restricted to an area not to exceed 1000 km 2, with a maximum time on site of 130 days.
Comprehensive Test Ban Treat3'
671
The on-site inspection (OSI) is an integral part of the overall Comprehensive Nuclear-Test-Ban Treaty (CTBT) verification regime. The OSI should be conducted with the view of obtaining scientific evidence to clarify whether an ambiguous event is a nuclear explosion. There are a number of phenomena that are associated with nuclear explosions which distinguished these from chemical explosions and/or natural events like earthquakes etc. Table 13.9 shows the list of nuclear explosion phenomena as well as the appropriate techniques which can be used for their detection. Different geophysical, radionuclide and visual observation technologies, allowed by the CTBT, are to be applied to search for signatures of nuclear explosions (NE) depending on the possible test environment: underground, underwater or atmosphere. Figure 13.1 illustrates possible on-site inspection activities, including sampling, search for surface disturbances and underground voids, radioactivity measurements, seismic listening for residual cavity collapse activity and, in the extreme case, drilling down into a suspected test cavity. Inspections will serve as an important supplement to the international data networks. During the initial phase of the inspection the highest priority should be given to techniques for capturing the shortest-lived phenomena, such as aftershocks. The narrowing of the search area has to be conducted with the least intrusive procedures
Fig. 13.1. Many phenomena make underground nuclear tests vulnerable to detection. On-site monitoring technologies include radionuclide sensing, flyovers for visual and multispectral analysis, local seismic recording, environmental sampling, potential field mapping and geophysical sounding. In a typical case, only a subset of these technologies would be used. The most intrusive measure would be drilling at the site of a suspected underground nuclear test.
672
0
0 0
~a
Z
c~
~,,,~
0
e-
.s
m
0
~
9~-~ ~ .~-
~
~- ~
0 ~ .=
t._.
~ -~ ~
~D a~
m
i -~ ~
•
X
XX
•
X
X
~
~
X
X
X
X
;~
X
X
X
X
~X
X
Chapter 13
XX
X
Comprehensive Test Ban Trea~
673
possible. The continuation and extension phases, if applied, have to be conducted to further refine evidence collected in an effort to identify regions to collect additional samples and data and then to obtain indicative information of an ambiguous event. The Provisional Technical Secretariat of the CTBTO PrepCom in co-operation with national OSI experts will develop the OSI methodology with a Concept of Operations and an Operational Manual as basic elements of this development. This will include identification of detection levels and blinding requirements for available technologies (gamma-monitoring, multi-spectral imaging, environmental radioactive and nonradioactive sampling), as well as identification and development of detailed equipment specifications for unique technologies.
REFERENCES Anisimov, A.N. et al., in Third Zababakhin Scientific Readings, Research Institute of Technical Physics, Chelyabinsk, Russia (1992), p. 24. Barth, K.-H., Science and politics in early nuclear test ban negotiations. Physics Today, March (1998) 34. Bolt, B.A., Nuclear Explosions and Earthquakes: The Parted Veil. Freeman, San Francisco, 1976. Carrigan, C.R. in Non-Proliferation Experiment: Results and Implications for Test Treaties (eds. Denny, M.D. and Stull, S.P.) 8-51-8-62 (Lawrence Livermore National Lab., Livermore, 1994. Carrigan, C.R., Heinle, R.A., Hudson, G.B., Nitao, J.J. and Zucca, J.J., Trace gas emissions on geological faults as indicators of underground nuclear testing. Nature, 382 (1996) 528. Drell, S. and Peurifoy, B.A., Rev. Nucl. Part. Sci., 44 (1994) 285-327. Hannon, W.J., Science, 227 (1985) 251-257. Jones, S., Kidder, R. and Frank von Hippel, The question of pure-fusion explosions under the CTBT. Physics Today, Sept. (1998) 57. Sullivan, J.D., The comprehensive test ban treaty. Physics Today, March (1998) 24. Zucca, J.J. in Non-Proliferation Experiment: Results and Implications for Test Treaties (eds. Denny, M.D. and Stull, S.P.) 8-1-8-6 (Lawrence Livermore National Lab., Livermore, 1994). Zuckerman, L., Nature, 361 (1993) 392-396.
675
INDEX
absorbed dose 272 accelerator 88, 114, 335-337 accelerator-driven system (ADS) 335, 337 accelerator mass spectrometry (AMS) 208, 209, 215, 216, 222-224, 227, 630, 631 accident 453 accidental release 377, 383, 385, 406, 453,472 accountancy 582 accountability verification 567,570 actinides 633, 199, 64, 65, 530 Action Team 598,599 activation product 378 aerosol samples 432, 433 Aerosol Sampling Station, ASS-500 425,427, 431 air sampler 424, 426, 437 air sampling 380, 424, 426 aircraft 50, 51,363,467, 495 airborne radioactivity 384, 417-419, 440, 658 algae 632 alpha-particle spectrometry 179, 181,600, 601 ambient dose equivalent 272 americium 201, 513, 55 annual dose limit 272 annual limit of intake (ALl) 272 Association for the Advancement of Medical Instruments (AAMI) 297 atmosphere 491 atmospheric emission 626, 627 ---discharge 329 --release 287 atmospheric test 378,489-491,495,531,534, 542, 546, 651,657, 666 atomic bomb 351,352, 356-359, 362, 365, 372, 373,489, 606 Australian Radiation Laboratory (ARL) 512 average glandular dose 272 avertable dose 272
balloon 495,536 basaltic rock 500 basic safety standards (BSS) 278, 281-284, 389, 550 beryllium 370, 593 beta particle spectrometry 177 Bikini atoll 513-518, 521,523-525 bio-accumulator (bio-indicator) 628, 633,635, 636 boiling-water reactor (BWR) 39, 309, 329, 341, 617 body burden 271 borehole 498 Born' s approximation 120 boron 360, 594 bottom ash 40 branchytherapy 81 Bremsstrahlung 264, 371
CAD 395 calcium 593 calibrated reference source 109, 110 calutron 89, 90, 602-605,615 CANDU reactor 617 capacitors 597 capture of neutrons 134 cardiology 78 cavity 504, 505 centrifuge 589 Cerenkov radiation 194 CERN 213 certified reference material (CRM) 571, 572-575 caesium 196, 288, 383,436, 470, 471,491, 518, 521,524, 545,635 chain reaction 335 Chernobyl 280, 324, 407, 599, 628
676 Chernobyl accident 396, 397, 401,422, 430, 431,441,449, 459-476, 633,637, 638 chronic exposure 272 coal 35-38, 85 coal fired powerplant 34 coconut 518 Codex Alimentarius Commission 450 colorimeter 298, 299 collected effective dose 273 committed effective dose 273 committed equivalent dose 273 Compton edge 130, 131 Compton effect 128, 131,133 computers 594 Concorde 49 consumer products 53, 54 core inventory 379, 466 cosmic rays 15, 49, 51,357, 409 Coulomb interaction 117, 126 counting 161,163-166 counting efficiency 178 counting rate 183 critical group 274, 286 critical mass 355,369 cross section 62 crucibles 593 CTBT 641,642-673 cyclotron 88 derived intervention limit (DIL) 453 destructive analysis (DA) 576 detection limit 169 detectors 497 deterministic effects 260 detonation 373 deuterium 596, 642 diaminotrinitrobenzene (DATB) 373 diet sampling 452 direct current plasma optical emission spectrometry (DCP-OES) 630 directional dose equivalent 274 dose 274, 281,285,286, 329, 406, 407,453, 468,469, 491,523,524, 530, 636 dose constraint 274 dose limitations 278, 279 dose pathways 289 dose rate 23, 50, 262-264, 267, 268, 454, 528, 529 dosimeter 293,297, 299, 454
Index
dosimetry 95,292, 294, 299, 456, 588 --high-dose 295,296 dose limits 284, 285 dust sampler 411, 419 early warning system 402 earthquake 650, 651 eddy diffusion 493 effective dose 274 elastic scattering 133 Electrical Power Research Institute (EPRI) 339 electrodeposition 201 electromagnet 591,604 electron spectroscopy for chemical analysis (ESCA) 618, 619 electron microprobe 630 electron microprobe analysis (EMA) 208-210 electron spin resonance (ESR) 298, 299-301 electron-positron pair 129 electro-thermal vaporization ICP-MS (ETV-ICP-MS) 208, 214 Emergency Response Center (ERC) 324 emergency response system (ERS) 403,404 Emergency Response Unit (ERU) 668, 669 energy calibration 173, 174 energy loss 118, 120-123 Enewetak atoll 516, 521 environmental contamination 468 enrichment 607, 608, 618, 619 Environmental Measurement Laboratory (EML) 422-424, 451,452 environmental monitoring 386, 390-402, 453, 478, 609, 613,618 environmental sampling 609, 613-615 environmental monitoring and sampling (EMS) 613,614 Environmental Protection Agency (EPA) 35, 451,513 equivalent dose 275 European Organization for Research and Treatment of Cancer (EORTC) 292 explosion 356, 525 explosive 373,374, 595,615,642 faeces samples 270, 633 fallout 391,455, 458 Fangataufa atoll 367, 496-498,500, 502, 532-538, 542, 549, 552-554 Fano factor 150
Index
Faraday cup 225 fast breeder reactor 306 fast neutrons 67 fertilizer 524 fissile material 567, 588 fission 62, 134, 357, 497, 537, 541,616, 621, 653 fission products 193, 378 fluorimetry 605 fly ash 3941 Food and Agriculture Organization (FAO) 280, 296, 450, 451,474 Food and Drug Administration (FDA) 451 food chain 551 food irradiation 70-73 food monitoring 450 food preservation 70, 72, 588 fossil fuel 323 fruit fly 70, 71 Fuego volcano 440 fuel fabrication 332 fungi 633,635,636 furnaces 593 fusion 305, 343,344, 346, 352, 354, 370, 497, 642 gamma ray spectrometry 169-171,600, 601, 6O5 gas amplification 142 gas centrifuge 603,608, 615, 618, 619 gas chromatography 660, 661 gas chromatography mass spectrometry (GCMS) 631 gas diffusion 603,606, 608, 618, 619 gas effluent 329 gas enrichment 608 gas flow counter 164 gas-cooled reactor 309, 339 gas-filled detector 138-140 gaseous waste 419 geosphere 551 geothermal energy 53 germanium detector 139, 154, 155, 157, 171 GIS 395, 396 Global Positioning System (GPS) 392-395 glow-discharge mass spectrometry (GDMS) 208, 214 glow-discharge optical emission spectroscopy (GD-OES) 208, 211
677 GM counter 137, 140, 142, 143, 177, 178, 384, 425,428, 429, 437 grass 635 graphite furnace AAS (GF-AAS) 208, 209, 211 graphite-moderated light-water reactor (LGR) 309 ground zero 525, 529, 530, 653
hafnium 594 half-value thickness (HVT) 266, 267 Hao atoll 532 heavy metals 635,637 heavy water 607 heavy-water reactor 309 high explosive 373, 374 high pressure liquid chromatography (HPLC) 630 high-level waste (HLW) 332, 333 high-volume air sampler 411 highly enriched uranium (HEU) 613-615, 618 Hiroshima 259, 363 hot cell 592 hot particles 472-474 hydro power 323 hydroacoustic monitoring network 665,666 hydrogen 597 hydrogen bomb 365,653 hydrogen mass spectrometry (HMS) 630 hydrosphere 628
IDMS 605 illicit trafficking 371 impact parameter 119 inductively coupled plasma (AES) 601 inductively coupled-plasma mass spectrometry (ICP-MS) 208, 214, 216-222, 630, 631 inductively coupled plasma-optical emission spectroscopy (ICP-OES) 208, 211,212 inelastic scattering 134 inertial confinement 345 inertial confinement fusion (ICF) 642 infrasound monitoring network 666-668 insects 69 intercomparison 234-236, 242, 399, 400, 470--472, 523 intermediate-level waste (ILW) 332, 333 internal exposure 269
678 International Atomic Energy Agency (IAEA) 81,234, 240, 244, 280, 283, 293,294, 296, 309, 323,324, 331-333,375,403,406, 474, 477,482, 522, 527,549, 550, 562, 563, 566-568, 581-583,588, 605,606, 613,668,669 International Chernobyl Project 242 International Commission on Radiological Protection (ICRP) 281,286, 289, 290, 327, 403,407, 442, 449, 455 International Data Centre (IDC) 644--647, 651, 668-670 International Dose Assurance Service (IDAS) 296 International Monitoring System (IMS) 644-647, 669, 670 International Nuclear Event Scale (INES) 324 international safeguards 561-566, 570, 571 International Standard Organization (ISO) 297, 298, 421 iodine 288, 383,468, 470, 633 ion chromatography (IC) 619, 630, 631 ion exchange separation 201 ion microtomography (IMT) 630 ion source 603,604 ionization 118, 120 ionization chamber 140, 141, 419 isobar 226, 227 isotope ratio mass spectrometry (IRMS) 228 isotope separator 603 isotope tracer technique 82, 83 kerma 275 Kili island 518 Kurchatov 526 laser 227, 345,590, 591 laser-ablation resonance-ionization spectroscopy (LARIS) 208, 215 laser-excited atomic fluorescence spectroscopy (LEAFS) 208,211 laser-excited resonance ionization spectroscopy (LERIS) 208, 213 laser-induced-breakdown spectroscopy (LIBS) 208, 211 laser-induced photoacoustic spectroscopy (LIPAS) 208, 229-231 laser mass spectroscopy (LMS) 208, 215 Lawrence Livermore National Laboratory (LLNL) 519
Index
leukaemia 329, 474 lichens 635-637 light-water reactor (LWR) 39, 566 liquid effluent 329, 330, 628 liquid-metal-cooled fast breeder reactor (LMFBR) 309 liquid nitrogen 147, 148,527 liquid release 288 liquid scintillation counting (LSC) 184, 185, 187, 190, 195,419, 630 liquid waste 419 lithium 593, 618 Lop Nor 507 low-enriched uranium (LEU) 566 low-level counting 162 low-level radioactivity 169, 188, 191 low-level waste (LLW) 332, 333,456 low-volume air sampler 411 luminescence 186 magnesium 594 magnetic confinement 344 Maralinga test site 508, 509, 513,522 Marine Environmental Laboratory (MEL-IAEA) 455,456 Marshall Islands 513, 514, 518-522 mass spectrometer 590 mass spectrometry 570, 601,604 --thermal ionization 600 --isotope dilution 600 maximum permissible dose 278, 279 medical exposure 284 metabolic pathways 269 minimum detectable activity (MDA) 161 mixed uranium-plutonium oxide (MOX) 334, 566, 580, 581 mobile radiological unit (laboratory) 453,454, 634 multichannel analyzer (MCA) 171 multiple scan average dose 275 Mururoa atoll 367,496--498, 500, 502, 531-538, 542, 549, 552, 553,650 mushrooms 637 mosses 635,636 mutants 68, 69 Nagasaki 259, 363,373 National Physical Laboratory (NPL) 298 neptunium 203
679
Index
neutron activation analysis 160, 338, 601 neutron detection 160 neutron moisture gauge 86 neutron radiography 82 neutrons 353-357 Nevada test site 507, 531 noble gas mass spectrometry (NGMS) 630 noble gases 657 non-destructive assay (NDA) 569, 565,571, 582 non-radiometric methods 207 nuclear accident 207, 378, 380, 386, 460, 477, 478, 668 nuclear bomb (device) 378, 390, 537 nuclear bore-hole logging 86 nuclear disarmament 562 nuclear explosion 366, 369, 643,644, 648, 649, 651, 671,672 nuclear explosive device 537, 561 nuclear imaging 79 nuclear material 614, 584-586 nuclear magnetic resonance (NMR) 591 nuclear medicine 74 nuclear power plant 306, 308, 310-322, 330, 339, 399, 458, 580, 581 nuclear powers 490 nuclear reaction 61-65,357, 502 nuclear reaction analysis (NRA) 630 nuclear reactor 379, 457, 477, 591 Nuclear Regulatory Commission (NRC) 339 nuclear safety 323,325 nuclear submarine 457 nuclear terrorism 371 nuclear track detector 159, 445,446 nuclear warhead 457 nuclear weapon 379, 489, 494, 523,561,562, 586, 588, 605,636, 641,642, 644 nuclear weapon states 642 nutrition 588 occupational exposure 284 on-site inspections (OSI) 670-673 optical emission spectroscopy 600 optical fibre 496 organ dose 276 oscilloscope 595 Pacific Test Center 507 pair production 129
particle induced X-ray emission (PIXE) 208-210, 630 paths of exposure 409, 410 penetrating power 135, 136 personal dose equivalent 276 PETN 595 pharmaceuticals 595, 596 phosphate fertilizers 42, 43, 46, 48, 67 phosphates 42, 45, 47, 605 photomultiplier tube 184, 185,444, 597 pine needles 637 PIPS detector 138, 179, 444 plasma 343,344, 371 plutonium 200-205, 333, 334, 367, 369, 370, 371,374, 388, 436, 438, 509, 512, 513, 521,522, 525,530, 537,541,545,546, 551,553,569, 589, 594, 601,602, 613, 614, 621,622, 638 polygon 525,526, 528 positron emission tomography (PET) 78 potassium 524 pregnancy 283 pressurized light-water reactor (PWR) 39, 309, 329, 341,617 project dose 276 proliferation 561,606, 608, 609 proportional counter 140, 142, 143,442 public exposure 276, 285 pulmonary tract 421 pulse-shape discrimination 162, 167 QA/QC procedures 231 quality assurance 231-234, 582 quality control 232, 233, 523,601 quenching 188 radiation cross-linking 87 radiation dose 183 wassessment 286, 289 radionuclide monitoring network 651,663-665 radiation detectors 137 radiation exposure 260, 278 radiation injury 278 radiation monitoring 391,392 radiation processing 88, 588 radiation safety 259 radiation survey meter 577 radiation weighting factor 276 radioactive decay series 7, 9
680
--thorium-232 10, 14, 22 --uranium-238 10-13, 22 --uranium-235 10 radioactive fallout 449 radioactive plume 526 radioactive waste 331 radioactivity 5, 21, 22, 61 --air 24 --coal 35-37 --rocks and soils 21, 22 --water 24 radioactivity standards 95 radiochemical analysis 192 radiography 87, 588 radioisotope instruments 84 radioisotope gauge 84, 85 radioisotopes 17-21, 65, 88, 90 radiological monitoring 388, 389, 391 radionuclides 5-8, 16, 596, 628, 631-633 radiopharmaceuticals 74-77 radiotherapy 292, 588 radiotracers 588 radon 27-31, 53, 183, 421,422, 4 4 2 4 4 8 Raman spectrometry (RS) 630 range 118, 123-126, 136, 264 rapid methods 205 --plutonium 205 --strontium 194, 206 --transuranic 205 Rayleigh scattering 127 RDX 595 reactor 336, 338, 607, 615,616 reactor fuel 618 reactor meltdown 387 reference air kerma rate 277 reference level 277 reference man 277 reference materials 238, 239, 241 release of effluents 326, 328 Remote Atmospheric Measurement Programme (RAMP) 423,424 reprocessing 581,592, 607, 620, 622, 624, 628, 633 research reactor 338, 567, 568, 581 resonance-ionization mass spectrometry (RIMS) 208, 209, 215,216 resonance-ionization spectroscopy (RIS) 208, 211 respiratory tract 290, 291
Index
robots 597 rocket 495 Rongerik atoll 516 rotor 589 Rutherford scattering 127 Safeguard Agreement 583,584 Safeguard Analytical Laboratory (SAL) 565, 568, 569, 560, 598,599 safeguards 113,579, 582, 582, 585,605,609, 610,613,614 --inspections 564, 566 safety trial (test) 537,542, 546, 553 sample preparation 196, 197 sampling methods (techniques) 413, 414, 415, 420, 629 scintillation detector 138, 144, 145, 166 spallation 64, 337 secondary-ion mass spectrometry (SIMS) 208, 209, 215 secondary standard dosimetry laboratory (SSDL) 293,294 sediment 629, 633 seismic monitoring network 647, 652-656 seismometer 648 self-absorption 164 Sellafield reprocessing plant 622-625 semiconductor detectors 137, 138, 146-151, 428 Semipalatinsk 507, 525,527 shock wave 356 silicon detector 156, 157 --efficiency 157, 158 soil sampling 412, 413,528 spark-source mass spectrometry (SSMS) 208, 214 spent fuel reprocessing 332 sputter-initiated resonance-ionization spectroscopy (SIRIS) 208, 215 standard solutions 101 standard sources 101, 103, 104 stellarator 346 sterile insect technique (SIT) 69, 70 stochastic effects 260 stopping power 122, 180, 181 stratosphere 492-495 strontium 192, 193,206, 213,470, 491,545 surface barrier detectors 138, 149 switching devices 597
681
Index
synchrotron radiation induced X-ray emission (SRIXE) 208-210 tandem accelerator 225 tandetron 225 tantalum 594 teletherapy 81 tenth-value thickness (TVT) 266 thermal ionization mass spectrometry (TIMS) 208, 209, 214, 619, 630 thermal neutrons 63 thermoluminescent detector (TLD) 159, 399 thermonuclear bomb (device) 369, 370, 515 thermonuclear reaction 642 thorium 455,569 Three Mile Island 280, 324, 466 threshold amount (TA) 614 thyroid cancer 474, 476 thyroid gland 80, 468, 470, 476, 633 tissue weighting factors 277 titanium 590 TNT 352, 356, 373,641,643,646, 648 Tokomak 346 tolerance dose 278 toluene 189, 190 toroidal confinement 344, 345 total reflection X-ray fluorescence analysis (TXRF) 208-210 tracer isotopes 568 tracer techniques 568 track detector 182 transmutation 335, 336 triaminotrinitrobenzene (TATB) 373 tritium 191,411,455,491,552, 553, 596, 618, 642 troposphere 492-495 Tuamotu archipelago 532 Tureia atoll 532, 533 undeclared activities 614 underground test 489, 490, 496, 531,534, 536, 538, 542, 546, 549, 552, 554, 648, 657,671
United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) 280, 385,463,475,550 uranium 351,367, 368, 436, 455, 527, 567, 569, 589, 593, 594, 601-605, 618, 619, 633,634 uranium enrichment 589, 591 urine samples 270 Van de Graaff accelerator 225 VYNS film 164, 165 waste 455,593 water samples 182, 195, 412 water-cooled reactor 339 weapon 352, 367-371,375,607, 614, 642 weapon test site 489, 532 weaponization 587, 607, 608 weapons fallout 628 weapons of mass destruction 585 whole body counting 400 Windscale 459, 466 World Energy Council (WEC) 305,343 World Health Organization (WHO) 81,244, 245,280, 284, 293, 294, 296, 403,450, 474, 550 World Meteorological Organization (WMO) 324, 326, 669 xenon 650, 660, 661 X-ray computed tomography 79 X-ray fluorescence (XRF) 48, 85, 108, 208, 209, 210, 560, 577, 599, 630, 631 X-ray generator 594 X-ray spectrometry 497 X-rays 65, 67, 73, 259, 266, 370 yellowcake 584 yttrium 194 zero point 498, 500, 501 ZnS counter 438, 440, 442, 444
