Landolt-Börnstein Numerical Data and Functional Relationships in Science and Technology New Series / Editor in Chief: W. Martienssen
Group VIII: Advanced Materials and Technologies Volume 4
Radiological Protection
Editors: A. Kaul, D. Becker Authors: D. Becker, G. Brix, A. Dalheimer, G. Dietze, H.R. Doerfel, K.F. Eckerman, H. Graffunder, Y. Harima, K. Hayashi, N. Ishigure, A. Kaul, H. Klewe-Nebenius, M. Lasch, H. Paretzke, N. Petoussi-Henss, A. Phipps, H. Smith, J.W. Stather, G.N. Stradling, D.M. Taylor, H.-G. Vogt, W. Weiss
ISSN 1619-4802 (Advanced Materials and Technologies) ISBN 3-540-20207-2 Springer Berlin Heidelberg New York Library of Congress Cataloging in Publication Data Zahlenwerte und Funktionen aus Naturwissenschaften und Technik, Neue Serie Editor in Chief: W. Martienssen Vol. VIII/4: Editor: A. Kaul, D. Becker At head of title: Landolt-Börnstein. Added t.p.: Numerical data and functional relationships in science and technology. Tables chiefly in English. Intended to supersede the Physikalisch-chemische Tabellen by H. Landolt and R. Börnstein of which the 6th ed. began publication in 1950 under title: Zahlenwerte und Funktionen aus Physik, Chemie, Astronomie, Geophysik und Technik. Vols. published after v. 1 of group I have imprint: Berlin, New York, Springer-Verlag Includes bibliographies. 1. Physics--Tables. 2. Chemistry--Tables. 3. Engineering--Tables. I. Börnstein, R. (Richard), 1852-1913. II. Landolt, H. (Hans), 1831-1910. III. Physikalisch-chemische Tabellen. IV. Title: Numerical data and functional relationships in science and technology. QC61.23 502'.12 62-53136 This work is subject to copyright. All rights are reserved, whether the whole or part of the material is concerned, specifically the rights of translation, reprinting, reuse of illustrations, recitation, broadcasting, reproduction on microfilm or in other ways, and storage in data banks. Duplication of this publication or parts thereof is permitted only under the provisions of the German Copyright Law of September 9, 1965, in its current version, and permission for use must always be obtained from Springer-Verlag. Violations are liable for prosecution act under German Copyright Law. Springer-Verlag Berlin Heidelberg New York a member of BertelsmannSpringer Science+Business Media GmbH © Springer-Verlag Berlin Heidelberg 2005 Printed in Germany The use of general descriptive names, registered names, trademarks, etc. in this publication does not imply, even in the absence of a specific statement, that such names are exempt from the relevant protective laws and regulations and therefore free for general use. Product Liability: The data and other information in this handbook have been carefully extracted and evaluated by experts from the original literature. Furthermore, they have been checked for correctness by authors and the editorial staff before printing. Nevertheless, the publisher can give no guarantee for the correctness of the data and information provided. In any individual case of application, the respective user must check the correctness by consulting other relevant sources of information. Cover layout: Erich Kirchner, Heidelberg Typesetting: Authors and Redaktion Landolt-Börnstein, Darmstadt Printing and Binding: AZ Druck, Kempten/Allgäu SPIN: 10723325
63/3020 - 5 4 3 2 1 0 – Printed on acid-free paper
Preface About 10 years before discovery of X-rays and natural radioactivity by W. C. Röntgen and H. Becquerel, more precisely in 1883, Hans Landolt, Richard Börnstein and Julius Springer have started a series of selected and easily retrievable physical data, which became a successful tool for natural scientists working or practising a profession in fields of chemistry, physics or technology. Now, i.e. about 120 years after start of this unique data collection and consequently about 100 years after introduction of ionizing radiations and radionuclides in natural sciences, medicine and technology, the Landolt-Börnstein New Series is submitting to the reader a full volume on protection of man against ionizing radiations and radionuclides, i.e. “Radiological Protection”. A comparison with the 6th edition of Landolt-Börnstein containing merely a six pages chapter on “Strahlenschutz” shows the rapid development of the field within the last five decades. Compared to many of the volumes in the Landolt-Börnstein Series published in the past, the present publication in the group Advanced Materials and Technologies is not only a compilation of numerical data and functional relationships for practical purposes. Rather a comprehensive accompanying text is intended to impart to the scientific or professional user of “Radiological Protection” both data, the concepts and scientific bases of the discipline devoted to prevention of health risks to man from exposure to ionizing radiations and radionuclides. Conceptually, radiological protection is based on the principles of justification of any use of ionizing radiation, of optimization of the application of radiation, and limitation of the radiation risk to man and his environment by acceptable doses, so that use of radiation and radionuclides in scientific research, medicine, technique and daily life is always of net benefit to man. Since findings of various scientific disciplines such as medicine, biology, biophysics, nuclear physics and techniques are the basis for radiological protection, multidisciplinary knowledge of fundamentals of these disciplines is necessary for an effective protection of man against health effects of ionizing radiations. Consequently, the present volume contains contributions of experts internationally qualified in scientific disciplines or subjects such as radiation physics, biology and medicine, external and internal dosimetry of ionizing radiation and radionuclides, decontamination and decorporation of radionuclides, or physical and biological measuring techniques. Although a previous volume in the Landolt-Börnstein Series has already considered shielding against high energy radiation such as of accelerators or of cosmic origin the specific item of assessment of radiation shielding was treated, too, however restricted to an extent being necessary for completion of tasks of practical radiological protection, specifically in the field of lower energies. The present volume addresses to • those already working in radiological protection, under the aspect of making available to them numerical data and functional relationships e.g. on assessment of radiation doses from external and internal sources, or with the aim of further education and impartation of most recent knowledge in radiological protection and scientific disciplines behind; • those participating in post-graduate education programmes in radiological protection with the aim to get a qualified expert e.g. in medical radiation physics, or as an employee in a competent national authority for health protection; • newcomers in the field of radiological protection to submit necessary knowledge on bases and practices of this discipline; • advanced students of physics, techniques or medicine with special interest in a later professional occupation as health physicists, engineers or technicians; • physicians practising in X-ray diagnostics, radiation oncology and nuclear medicine with special interest in medical radiological protection.
In the hardcopy of the present volume a CD-ROM is included containing: • the full text in the multi-platform Adobe-Acrobat(pdf)-format with searchable fulltext index and • additional information and data, which would be beyond the scope of the printed version, within the interactive programme SISy (for MS-Windows only). These refer e.g. to decay data of radionuclides or normalized excretion functions for monitoring workers by quantitative assessment of intakes of radionuclides and calculation of resulting doses. For further numerical data such as dose coefficients for intake of radionuclides by workers or members of the public that are available from publications e.g. of the International Commission on Radiological Protection ICRP or of the International Commission on Radiation Units and Measurements ICRU either as hardcopies or in the Internet are not contained on the CD-ROM. The reader is referred to the relevant original sources. The editors of the present volume want to thank the authors of the contributions for their careful work, the Editor in Chief of the Landolt-Börnstein Series, Prof. Dr. W. Martienssen, for having put "Radiological Protection" on the list of volumes to be prepared for the New Series, and the Publisher, especially Drs. Ch. Meier and R. Poerschke from the editorial office for their permanent and very active engagement in realizing the present opus.
The Editors
Braunschweig/Salzgitter, 2004
Contributors
Editors D. Becker Bundesamt für Strahlenschutz Fachbereich KT 2 Willy-Brandt-Straße 5 38226 Salzgitter-Lebenstedt GERMANY
A. Kaul Physikalisch-Technische Bundesanstalt Bundesallee 100 38116 Braunschweig GERMANY
Authors D. Becker Bundesamt für Strahlenschutz Fachbereich KT 2 Willy-Brandt-Straße 5 38226 Salzgitter-Lebenstedt GERMANY 1 Intoduction
G. Brix Bundesamt für Strahlenschutz Institut für Strahlenhygiene Ingolstädter Landstraße 1 Neuherberg 85764 Oberschleißheim GERMANY 10 Measuring techniques
A. Dalheimer Bundesamt für Strahlenschutz Fachbereich Strahlenschutz und Gesundheit Köpenicker Allee 120-130 10318 Berlin GERMANY 10 Measuring techniques
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Contributors
G. Dietze Physikalisch-Technische Bundesanstalt Abt. 6 Bundesallee 100 38116 Braunschweig GERMANY 4 Radiological quantities and units 10 Measuring techniques
H.R. Doerfel Forschungszentrum Karlsruhe Hauptabteilung Sicherheit Postfach 3640 76021 Karlsruhe GERMANY 10 Measuring techniques
K.F. Eckerman Health Sciences Research Division Oak Ridge National Laboratory 1060 Commerce Park Oak Ridge Tennessee 37831-6480 USA 7 Internal dosimetry of radionuclides
H. Graffunder Ingenieurbüro Graffunder Friedrichstraße 28 76297 Stutensee GERMANY Radiation Protection Information System (SISy)
Y. Harima Tokyo Institute of Tecnology Research Laboratory for Nuclear Reactors 7-3-4-307 Hikarigaoka Nerima-ku Tokyo 179-0072 JAPAN 5 Shielding against ionizing radiation
K. Hayashi Hitachi, Ltd. Nuclear Plant Engineering Department Saiwai-cho, 3-1-1, Hitachi, Ibaraki, 317-8511 JAPAN 5 Shielding against ionizing radiation
Contributors
N. Ishigure Research Center for Radiation Safety National Institute of Radiological Science 4-9-1, Anagawa, Inage, Chiba 263-8555 JAPAN 7 Internal dosimetry of radionuclides
A. Kaul Physikalisch-Technische Bundesanstalt Bundesallee 100 38116 Braunschweig GERMANY 1 Introduction 8 Decontamination 11 Exposures from natural and man-made radiation sources
H. Klewe-Nebenius Forschungszentrum Karlsruhe Institut für Instrumentelle Analytik Postfach 3640 76021 Karlsruhe GERMANY 3 Physical fundamentals
M. Lasch Kernkraftwerke Grundremmingen Betriebsgesellschaft mbH Postfach 89355 Grundremmingen GERMANY 8 Decontamination
H.G. Paretzke Institut für Strahlenschutz GSF-Forschungszentrum für Umwelt und Gesundheit, GmbH Neuherberg, Postfach 11 29 85758 Oberschleißheim GERMANY 6 External dosimetry
N. Petoussi-Henss Institut für Strahlenschutz GSF-Forschungszentrum für Umwelt und Gesundheit, GmbH Neuherberg, Postfach 11 29 85758 Oberschleißheim GERMANY 6 External dosimetry
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Contributors
A. Phipps National Radiological Protection Board Chilton Didcot Oxfordshire OX11 ORQ UNITED KINGDOM 7 Internal dosimetry of radionuclides
H. Smith † National Radiological Protection Board Chilton Didcot Oxfordshire OX11 ORQ UNITED KINGDOM 2 Biological effects of ionising radiation
J.W. Stather National Radiological Protection Board Chilton Didcot Oxfordshire OX11 ORQ UNITED KINGDOM 2 Biological effects of ionising radiation 7 Internal dosimetry of radionuclides
G.N. Stradling National Radiological Protection Board Chilton Didcot Oxfordshire OX11 ORQ UNITED KINGDOM 9 Decorporation of radionuclides
D.M. Taylor 5, Pan Poeth Pen-y-bont, CF31 5BD Wales UNITED KINGDOM 9 Decorporation of radionuclides
H.-G. Vogt Zentrum für Strahlenschutz und Radioökologie Universität Hannover Am Kleinen Felde 30 30167 Hannover GERMANY 5 Shielding against ionizing radiation
Contributors
W. Weiss Bundesamt für Strahlenschutz Fachbereich Strahlenhygiene Institut für Strahlenhygiene Ingolstädter Landstr. 1 85764 Oberschleißheim GERMANY 10 Measuring techniques
Landolt-Börnstein Editorial Office Gagernstr. 8, D-64283 Darmstadt, Germany fax: +49 (6151) 171760 e-mail:
[email protected] Internet http://www.landolt-boernstein.com
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VIII/4 Radiological Protection Contents 1
The development of the organizational and the conceptual basis of radiological protection ........................................................................................................................................................................................................................ 1-1
2 2.1 2.2 2.2.1 2.2.2 2.2.3 2.2.4 2.3 2.3.1 2.3.2 2.3.3 2.3.4 2.4 2.4.1 2.4.2 2.4.3 2.4.4 2.4.5 2.4.6 2.4.7 2.5 2.5.1 2.5.2 2.6 2.6.1 2.6.2 2.6.3 2.6.4 2.7 2.8 2.9
Biological effects of ionising radiation .......................................................................................................................................... 2-1 Introduction .................................................................................................................................................................................................................... 2-1 Cellular effects ............................................................................................................................................................................................................ 2-2 Primary events following exposure to ionising radiation ................................................................................................ 2-2 Cellular damage and repair following the primary radiation events .................................................................... 2-3 Classification of radiation-induced damage .................................................................................................................................. 2-5 Implications of cellular damage for whole or partial body exposure .................................................................. 2-9 Deterministic effects.............................................................................................................................................................................................. 2-9 Tissue and organ development.................................................................................................................................................................... 2-9 Dose-response relationships for radiation damage ............................................................................................................. 2-10 Deterministic effects in humans following acute whole-body irradiation.................................................. 2-11 Deterministic effects following partial body irradiation............................................................................................... 2-12 Radiation-induced cancer ............................................................................................................................................................................. 2-15 Cancer development .......................................................................................................................................................................................... 2-15 Dose-response relationships ...................................................................................................................................................................... 2-16 Exposures to external radiation .............................................................................................................................................................. 2-18 Exposure to internally incorporated radionuclides ............................................................................................................. 2-20 Dose and dose rate effectiveness factors (DDREFs)........................................................................................................ 2-23 Risk coefficients for protection .............................................................................................................................................................. 2-23 Low dose studies ................................................................................................................................................................................................... 2-25 Hereditary disease ................................................................................................................................................................................................ 2-27 Categories of genetic damage .................................................................................................................................................................. 2-27 Risk coefficients for hereditary disease ......................................................................................................................................... 2-28 Irradiation in utero ............................................................................................................................................................................................... 2-29 Deterministic effects.......................................................................................................................................................................................... 2-30 Brain function........................................................................................................................................................................................................... 2-30 Risk coefficients for cancer........................................................................................................................................................................ 2-31 Hereditary disease ................................................................................................................................................................................................ 2-31 Summary of risk factors for cancer and hereditary disease ....................................................................................... 2-32 Conclusions ................................................................................................................................................................................................................ 2-32 References .................................................................................................................................................................................................................... 2-34
3 3.1 3.2 3.2.1 3.2.2 3.3 3.3.1 3.3.2 3.4 3.4.1 3.4.2 3.4.3 3.5
Physical fundamentals ..................................................................................................................................................................................... 3-1 Natural radioactivity .............................................................................................................................................................................................. 3-1 Elements, isotopes and radionuclides .................................................................................................................................................. 3-2 Atoms, electrons and the Periodic Table of Elements ........................................................................................................ 3-2 Atomic nuclei, nuclides and the Chart of Nuclides ............................................................................................................... 3-3 The structure of the atomic nucleus ....................................................................................................................................................... 3-4 Elementary particles .............................................................................................................................................................................................. 3-4 Nuclear transformations..................................................................................................................................................................................... 3-6 Radioactive decay ................................................................................................................................................................................................ 3-10 Basic properties ...................................................................................................................................................................................................... 3-10 Decay modes ............................................................................................................................................................................................................. 3-15 The natural radioactive decay families ........................................................................................................................................... 3-18 Radioactive radiation ........................................................................................................................................................................................ 3-24
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3.5.1 3.5.2 3.6 3.6.1 3.6.2 3.6.3 3.6.4 3.6.5 3.6.6 3.7
Types of radiation ................................................................................................................................................................................................ 3-24 Physical properties of radiation.............................................................................................................................................................. 3-26 Nuclear fission and fission products ................................................................................................................................................. 3-30 Particle induced nuclear fission ............................................................................................................................................................. 3-30 Fission products ..................................................................................................................................................................................................... 3-30 Nuclear reactors ..................................................................................................................................................................................................... 3-32 Nuclear explosives .............................................................................................................................................................................................. 3-32 Radioactive inventory and nuclear waste..................................................................................................................................... 3-33 Release of radionuclides from the radioactive inventory of a nuclear reactor ....................................... 3-38 References .................................................................................................................................................................................................................... 3-39
4 4.1 4.2 4.2.1 4.2.2 4.3 4.3.1 4.3.2 4.3.3 4.3.4 4.4 4.4.1 4.4.2 4.5 4.5.1 4.5.2 4.5.3 4.6 4.6.1 4.6.2 4.7 4.8 4.9
Radiological quantities and units........................................................................................................................................................ 4-1 Introduction .................................................................................................................................................................................................................... 4-1 Radiation field quantities .................................................................................................................................................................................. 4-2 Scalar radiation field quantities .................................................................................................................................................................. 4-2 Vectorial radiation field quantities.......................................................................................................................................................... 4-4 Interaction coefficients and quantities................................................................................................................................................. 4-5 Cross section.................................................................................................................................................................................................................. 4-5 Mass attenuation coefficient and mass energy transfer coefficient ....................................................................... 4-5 Mass stopping power and linear energy transfer (LET).................................................................................................... 4-6 Mean energy expended in a gas per ion pair formed ........................................................................................................... 4-7 Quantities related to energy transfer ..................................................................................................................................................... 4-7 Stochastic quantities .............................................................................................................................................................................................. 4-7 Non-stochastic quantities.................................................................................................................................................................................. 4-8 Dose quantities in radiation protection ........................................................................................................................................... 4-10 Concept of radiation protection quantities .................................................................................................................................. 4-10 Protection quantities .......................................................................................................................................................................................... 4-11 Operational quantities ...................................................................................................................................................................................... 4-14 Radioactivity quantities .................................................................................................................................................................................. 4-18 Activity, specific activity, activity concentration, activity per area .................................................................. 4-19 Specific quantities for radon, thoron and their progeny................................................................................................ 4-19 Quantities for internal dosimetry .......................................................................................................................................................... 4-22 Limits, constraints, action levels ........................................................................................................................................................... 4-23 References .................................................................................................................................................................................................................... 4-27
5 5.1 5.2 5.3 5.3.1 5.3.2 5.4 5.4.1 5.4.2 5.4.3 5.5 5.5.1 5.5.2 5.5.3 5.5.4 5.6 5.7
Shielding against ionizing radiation ................................................................................................................................................ 5-1 Introduction .................................................................................................................................................................................................................... 5-1 Stopping power and range ............................................................................................................................................................................... 5-1 Penetration depths of charged particles ............................................................................................................................................. 5-2 Heavy charged particles ..................................................................................................................................................................................... 5-2 Electrons and positrons....................................................................................................................................................................................... 5-4 Photons ................................................................................................................................................................................................................................ 5-5 Basic shielding concept ...................................................................................................................................................................................... 5-5 Attenuation data of radioactive sources in shielding materials.................................................................................. 5-6 An example of the calculation of an ambient dose equivalent rate ....................................................................... 5-9 Neutrons ......................................................................................................................................................................................................................... 5-16 Basic shielding concepts ............................................................................................................................................................................... 5-16 Attenuation data of various neutron sources in shielding materials ................................................................. 5-16 Sample shield calculation ............................................................................................................................................................................. 5-18 Induced activity ...................................................................................................................................................................................................... 5-19 Computer codes and online nuclear data services............................................................................................................... 5-28 References .................................................................................................................................................................................................................... 5-32
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6 6.1 6.1.1 6.1.2 6.2 6.2.1 6.2.2 6.2.3 6.2.4 6.3 6.3.1 6.3.2 6.4 6.5 6.5.1 6.5.2 6.6 6.7
External dosimetry .............................................................................................................................................................................................. 6-1 Protection and operational quantities ................................................................................................................................................... 6-1 Protection quantities .............................................................................................................................................................................................. 6-1 Operational Quantities ......................................................................................................................................................................................... 6-1 Dosimetric models ................................................................................................................................................................................................... 6-2 Models and phantoms of the human body ...................................................................................................................................... 6-2 Idealized geometries representing occupational exposures ........................................................................................... 6-4 Environmental source geometries............................................................................................................................................................ 6-4 Methods of calculating protection quantities in computational models............................................................ 6-5 Conversion coefficients for photons ..................................................................................................................................................... 6-6 Occupational .................................................................................................................................................................................................................. 6-6 Conversion coefficients for environmental gamma ray fields ................................................................................ 6-13 Conversion coefficients for neutrons ............................................................................................................................................... 6-20 Conversion coefficients for electrons .............................................................................................................................................. 6-21 Occupational exposure .................................................................................................................................................................................... 6-21 Environmental exposure ................................................................................................................................................................................ 6-23 Doses from external exposure of radionuclides in the environment ................................................................ 6-23 References .................................................................................................................................................................................................................... 6-42
7 7.1 7.2 7.2.1 7.2.2 7.2.3 7.2.4 7.2.5 7.2.6 7.2.7 7.2.8 7.3 7.3.1 7.3.2 7.3.3 7.3.4 7.3.5 7.3.6 7.4 7.4.1 7.4.2 7.4.3 7.5 7.5.1 7.6 7.6.1 7.6.2 7.6.3 7.6.4 7.6.5 7.7 7.7.1 7.7.2 7.8
Internal dosimetry of radionuclides ................................................................................................................................................ 7-1 Introduction .................................................................................................................................................................................................................... 7-1 Biokinetics of radionuclides in the body .......................................................................................................................................... 7-2 Inhalation .......................................................................................................................................................................................................................... 7-3 Ingestion......................................................................................................................................................................................................................... 7-10 Cuts and wounds ................................................................................................................................................................................................... 7-14 Absorption through intact skin ............................................................................................................................................................... 7-15 Systemic behaviour of radionuclides................................................................................................................................................ 7-15 Excretion ....................................................................................................................................................................................................................... 7-21 Embryo and foetus............................................................................................................................................................................................... 7-21 Transfer in maternal milk ............................................................................................................................................................................. 7-24 Dosimetric models ............................................................................................................................................................................................... 7-25 Introduction ................................................................................................................................................................................................................ 7-25 Absorbed fraction and specific absorbed fraction ............................................................................................................... 7-26 Computational models of the human anatomy ....................................................................................................................... 7-28 Dose rate per unit activity, S-factor ................................................................................................................................................... 7-31 Specific absorbed fractions for various radiations.............................................................................................................. 7-32 Calculation of doses to soft tissues and the skeleton........................................................................................................ 7-36 Dose coefficients ................................................................................................................................................................................................... 7-37 Method of calculation ...................................................................................................................................................................................... 7-38 Sources of dose coefficients ...................................................................................................................................................................... 7-41 Dose coefficients for selected radionuclides ............................................................................................................................ 7-43 Internal monitoring ............................................................................................................................................................................................. 7-46 Methods of individual monitoring....................................................................................................................................................... 7-47 Monitoring programme................................................................................................................................................................................... 7-51 Need for a monitoring programme ..................................................................................................................................................... 7-51 Routine monitoring ............................................................................................................................................................................................. 7-51 Special or task-related monitoring ...................................................................................................................................................... 7-52 Confirmatory monitoring .............................................................................................................................................................................. 7-52 Wound monitoring .............................................................................................................................................................................................. 7-52 Dose Assessment................................................................................................................................................................................................... 7-53 Estimation of intake and dose .................................................................................................................................................................. 7-53 Control of worker doses ................................................................................................................................................................................ 7-54 Monitoring data for radionuclides (H-3, Co-60, Sr-90, Ru-106, I-131, Cs-134, Cs-137, Ce-144 U-234, Pu-239, Am-241).............................................................................................................................................................................. 7-55 References .................................................................................................................................................................................................................... 7-68
7.9
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8 8.1 8.1.1 8.1.1.1 8.1.1.2 8.1.1.3 8.1.1.4 8.1.1.5 8.1.1.6 8.1.1.7 8.1.1.8 8.1.1.9 8.1.2 8.1.2.1 8.1.2.2 8.1.2.3 8.1.2.3 8.1.3 8.1.3.1 8.1.3.2 8.1.3.3 8.1.3.4 8.1.3.5 8.1.3.6 8.1.3.7 8.1.3.8 8.1.3.9 8.1.3.10 8.1.4 8.2 8.2.1 8.2.2 8.2.2.1 8.2.2.2 8.2.3 8.2.3.1 8.2.3.2 8.2.4 8.2.4.1 8.2.4.2 8.2.4.3 8.2.4.4 8.2.5 8.2.5.1 8.2.5.2 8.3
Decontamination .................................................................................................................................................................................................... 8-1 Decontamination of materials ...................................................................................................................................................................... 8-1 General approaches to decontamination ........................................................................................................................................... 8-2 Contamination .............................................................................................................................................................................................................. 8-2 Characteristics of oxide layer in BWRs and PWRs .............................................................................................................. 8-2 Other types of contamination ....................................................................................................................................................................... 8-3 Decontamination ....................................................................................................................................................................................................... 8-4 The use of decontamination in decommissioning ................................................................................................................... 8-4 Identification of decontaminable components ............................................................................................................................ 8-5 Effectiveness of decontamination, decontamination factor ........................................................................................... 8-6 Decontamination techniques (processes) ......................................................................................................................................... 8-7 Decontamination and secondary waste generation................................................................................................................. 8-8 Decontamination techniques for large volume closed systems ................................................................................. 8-8 Reactor decontamination in BWRs and PWRs .......................................................................................................................... 8-8 Fuel assemblies and decontamination.............................................................................................................................................. 8-12 Decontamination of sodium cooled systems............................................................................................................................. 8-12 Gas cooled reactors (WAGR) .................................................................................................................................................................. 8-13 Decontamination techniques for segmented parts ............................................................................................................... 8-13 Chemical decontamination .......................................................................................................................................................................... 8-13 Electrochemical decontamination ........................................................................................................................................................ 8-15 Jetting decontamination techniques ................................................................................................................................................... 8-18 Ultrasonic decontamination ....................................................................................................................................................................... 8-19 Decontamination by foams ......................................................................................................................................................................... 8-21 Decontamination by gels............................................................................................................................................................................... 8-21 Decontamination by pastes ......................................................................................................................................................................... 8-21 Mechanical decontamination techniques ...................................................................................................................................... 8-21 Decontamination by strippable coatings ....................................................................................................................................... 8-22 Melting ............................................................................................................................................................................................................................ 8-22 Decontamination techniques for building surfaces ............................................................................................................ 8-22 Decontamination of skin ............................................................................................................................................................................... 8-24 Introduction ................................................................................................................................................................................................................ 8-24 Transport of radioactive substances via the skin .................................................................................................................. 8-24 Anatomy of the skin........................................................................................................................................................................................... 8-24 Transport procedure ........................................................................................................................................................................................... 8-25 Skin dose at contamination ......................................................................................................................................................................... 8-25 Calculation of the equivalent dose to the skin......................................................................................................................... 8-25 Equivalent dose rate conversion coefficients ........................................................................................................................... 8-26 Decontamination measures ......................................................................................................................................................................... 8-30 Organisational and preliminary measures ................................................................................................................................... 8-30 First aid measures of skin decontamination............................................................................................................................... 8-30 Specific decontamination procedures .............................................................................................................................................. 8-31 Decontamination of specific body regions and organs .................................................................................................. 8-32 Procedure at residual contamination and fixing a reference value..................................................................... 8-32 Frequency of decontamination steps................................................................................................................................................. 8-32 Derivation of the reference value for residual contamination ................................................................................. 8-32 References .................................................................................................................................................................................................................... 8-34
9 9.1 9.2 9.2.1 9.2.2 9.2.3
Decorporation of radionuclides ............................................................................................................................................................ 9-1 Introduction .................................................................................................................................................................................................................... 9-2 General considerations ........................................................................................................................................................................................ 9-2 Factors affecting the efficacy of treatment ..................................................................................................................................... 9-2 Factors influencing treatment decisions ............................................................................................................................................ 9-3 Decision levels ............................................................................................................................................................................................................ 9-3
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9.4.4 9.4.5 9.4.6 9.4.7 9.4.8 9.4.9 9.4.10 9.4.11 9.4.12 9.4.13 9.4.14 9.5 9.5.1 9.5.2 9.5.3 9.6 9.6.1 9.6.2 9.6.3 9.6.4 9.6.5 9.6.6 9.6.7 9.7 9.8
Perception of risk and its implications ................................................................................................................................................ 9-4 Approaches to treatment.................................................................................................................................................................................... 9-5 Methods of treatment ............................................................................................................................................................................................ 9-6 Non-specific procedures .................................................................................................................................................................................... 9-6 Procedures to enhance systemic radionuclide excretion................................................................................................... 9-7 General comments on the efficiacy of chelating agents for the actinides ....................................................... 9-8 What are the factors that govern the efficacy of chelating agents ? ..................................................................... 9-9 Can the efficacy of treatment be predicted from animal studies ? .......................................................................... 9-9 Are chelating agents always most effective when the radionuclides are present in circulating blood ? ................................................................................................................................................................................................................................. 9-9 Is DTPA effective for all actinides ? ................................................................................................................................................ 9-10 Will the administration of chelating agents result in enhanced tissue deposition ? .......................... 9-10 Is the administration of sodium carbonate effective for uranium ? ................................................................... 9-10 Must chelating agents be administered promptly to be effective ? .................................................................... 9-10 Is intravenous injection the best mode of administration ?........................................................................................ 9-11 How can judgements on efficacy be made ?............................................................................................................................. 9-11 When should treatment start ? ................................................................................................................................................................. 9-11 When should treatment stop ? ................................................................................................................................................................. 9-11 For which materials are chelating agents likely to be effective ? ........................................................................ 9-11 For which materials are chelating agents unlikely to be effective ? ................................................................. 9-12 Is lung lavage more effective than chelation treatment for inhaled materials ?.................................... 9-12 Recent developments ........................................................................................................................................................................................ 9-13 Plutonium and americium ............................................................................................................................................................................ 9-13 Thorium .......................................................................................................................................................................................................................... 9-13 Uranium.......................................................................................................................................................................................................................... 9-13 Optimum treatment protocols................................................................................................................................................................... 9-14 Tritium ............................................................................................................................................................................................................................. 9-14 The alkaline earth elements, strontium, barium and radium..................................................................................... 9-14 Iodine ................................................................................................................................................................................................................................ 9-16 Caesium .......................................................................................................................................................................................................................... 9-16 Plutonium and americium ............................................................................................................................................................................ 9-18 Thorium .......................................................................................................................................................................................................................... 9-26 Uranium.......................................................................................................................................................................................................................... 9-27 Future research needs ....................................................................................................................................................................................... 9-29 References .................................................................................................................................................................................................................... 9-31
10 10.1 10.1.1 10.1.2 10.1.2.1 10.1.2.2 10.1.2.3 10.1.2.4 10.1.3 10.1.4 10.1.5 10.1.6 10.1.7 10.1.8 10.1.9 10.2 10.2.1
Measuring techniques .................................................................................................................................................................................. 10-1 Detectors for radiation protection ........................................................................................................................................................ 10-1 Overview and general characteristics of radiation detectors .................................................................................... 10-1 Gas-filled ionization detectors................................................................................................................................................................. 10-3 Ionization and gas amplification ........................................................................................................................................................... 10-3 Ionization chambers ........................................................................................................................................................................................... 10-4 Proportional counters........................................................................................................................................................................................ 10-8 Geiger-Müller counters ............................................................................................................................................................................... 10-10 Scintillation detectors .................................................................................................................................................................................... 10-11 Semiconductor detectors ............................................................................................................................................................................ 10-14 Thermoluminescence and radiophotoluminescence detectors ............................................................................. 10-18 Photographic films ........................................................................................................................................................................................... 10-20 Detectors for neutrons................................................................................................................................................................................... 10-21 Biological dosimetry ...................................................................................................................................................................................... 10-23 References for 10.1 .......................................................................................................................................................................................... 10-25 Radiological protection measurements: external exposure ..................................................................................... 10-27 Operational quantities ................................................................................................................................................................................... 10-27
9.2.4 9.2.5 9.3 9.3.1 9.3.2 9.4 9.4.1 9.4.2 9.4.3
Contents
XVII
10.2.2 10.2.3 10.2.3.1 10.2.3.2 10.2.4 10.2.4.1 10.2.4.2 10.2.4.3 10.2.5 10.2.5.1 10.2.5.2 10.2.6 10.2.6.1 10.2.6.2 10.2.6.3 10.2.6.4 10.2.7 10.3 10.3.1 10.3.1.1 10.3.1.2 10.3.1.3 10.3.1.4 10.3.2 10.3.2.1 10.3.2.2 10.3.2.3 10.3.2.4 10.3.2.5 10.3.2.6 10.3.2.7 10.3.2.8 10.3.2.9 10.3.2.10 10.3.2.11 10.3.3 10.3.3.1 10.3.3.2 10.3.3.3 10.3.3.4 10.3.3.5 10.3.3.6 10.3.3.7 10.3.3.8 10.3.3.9 10.3.3.10
Reference levels .................................................................................................................................................................................................. 10-29 Types of exposure ............................................................................................................................................................................................. 10-29 Occupational exposure ................................................................................................................................................................................. 10-30 Public exposure ................................................................................................................................................................................................... 10-30 Types of monitoring programs............................................................................................................................................................. 10-31 Individual monitoring for external exposure ......................................................................................................................... 10-31 Area monitoring for external exposure ....................................................................................................................................... 10-31 Calibration ................................................................................................................................................................................................................ 10-32 Requirements for individual monitoring of external exposure ........................................................................... 10-33 Operational requirements .......................................................................................................................................................................... 10-34 Accuracy requirement ................................................................................................................................................................................... 10-35 Personal dosimeters for individual monitoring in different radiation fields.......................................... 10-36 Photon dosimetry ............................................................................................................................................................................................... 10-36 Beta dosimetry ...................................................................................................................................................................................................... 10-37 Neutron dosimetry ............................................................................................................................................................................................ 10-37 Dosimetry in mixed field situations (photons and neutrons) ................................................................................ 10-38 References for 10.2 .......................................................................................................................................................................................... 10-39 Radiological protection measurements: internal exposure ...................................................................................... 10-41 Measurement of radon and its progeny ....................................................................................................................................... 10-41 Measurement of radon in air .................................................................................................................................................................. 10-44 Measurement of radon progeny in air........................................................................................................................................... 10-47 Measurement of radon in the ground and in water .......................................................................................................... 10-50 References for 10.3.1 ..................................................................................................................................................................................... 10-51 In vivo measurements ................................................................................................................................................................................... 10-52 Introduction ............................................................................................................................................................................................................. 10-52 Requirements ......................................................................................................................................................................................................... 10-53 Principles of γ spectrometry ................................................................................................................................................................... 10-56 Equipment ................................................................................................................................................................................................................. 10-58 Spectrum evaluation ....................................................................................................................................................................................... 10-66 Measuring geometries ................................................................................................................................................................................... 10-67 Calibration ................................................................................................................................................................................................................ 10-71 Uncertainties and detection limits..................................................................................................................................................... 10-77 Measurement procedure.............................................................................................................................................................................. 10-79 Quality assurance and control............................................................................................................................................................... 10-79 References for 10.3.2 ..................................................................................................................................................................................... 10-81 In vitro measurements: excretion analyses .............................................................................................................................. 10-83 Introduction ............................................................................................................................................................................................................. 10-83 Urine samples ........................................................................................................................................................................................................ 10-83 Faeces samples ..................................................................................................................................................................................................... 10-84 Exhalation ................................................................................................................................................................................................................. 10-84 Other biological samples............................................................................................................................................................................ 10-85 Radiochemical analyses .............................................................................................................................................................................. 10-85 Measuring techniques.................................................................................................................................................................................... 10-88 Quality assurance............................................................................................................................................................................................... 10-91 Examples for dose estimations from in vitro measurements ................................................................................. 10-93 References for 10.3.3 ..................................................................................................................................................................................... 10-97
11 11.1 11.2 11.2.1 11.2.2
Exposures from natural and man-made radiation sources ............................................................................. 11-1 Introduction ................................................................................................................................................................................................................ 11-1 Exposures by cosmic radiation and cosmogenic radionuclides............................................................................. 11-2 Origin and kinds of cosmic radiation ............................................................................................................................................... 11-2 Exposures by cosmic radiations............................................................................................................................................................. 11-2
XVIII 11.3 11.3.1 11.3.2 11.4 11.5 11.6
Contents Terrestrial radiation ............................................................................................................................................................................................ 11-4 External exposures .............................................................................................................................................................................................. 11-5 Internal exposures ................................................................................................................................................................................................ 11-6 Enhanced exposures form industrial activities ....................................................................................................................... 11-9 Worldwide average exposure from natural and man-made sources .............................................................. 11-10 References ................................................................................................................................................................................................................. 11-12
1 The development of the organizational and conceptual basis of radiological protection
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1 The development of the organizational and conceptual basis of radiological protection1
Within a few weeks of Roentgen's discovery of X-rays in 1895, the potential of X-rays for diagnosing fractures became apparent. However, the occurrence of acute adverse effects such as erythema and skin burns within the next few years made persons applying X-rays in medicine and technique aware of the need to avoid overexposure. Similar undesirable effects were reported after the discovery of natural radioactivity by H. Becquerel in 1896, specifically of radium by M. Curie, and medical application. The first organized coordinated effort for radiation protection came in 1921 when the British X-ray and Radium Protection Committee issued detailed recommendations and instructions. The American Roentgen Ray Society also proposed general recommendations in the early 1920s on the basis of avoiding acute effects. At the Second International Congress of Radiology held in Stockholm in 1928 [28B1], the unit “roentgen” (R) was recognized as the unit for X-ray dose. It was at this congress that the “International X-ray and Radium Protection Commission” was founded, the forerunner of the later (from 1950 onwards) “International Commission on Radioligical Protection (ICRP)”. The primary concern of the 1928 Commission was to elaborate recommendations designed to provide protection to members of the medical profession in their work with X-rays and gamma-rays from radium. In 1934 the Commission recommended 0.2 R per day as the “tolerance dose” [34I1]. Due to the great expansion in radiation protection work consequent upon nuclear energy developments in the period from 1940 to 1950, the International Congress of Radiology in 1950 [51I1] extended the scope of the Commission - now ICRP - and broadened its area of responsibility beyond the protection of the medical profession only. It was in 1950 that the ICRP spelt out the first time the various effects which were to be considered in making its recommendations. These recommendations were to deal primarily with the basic principles of radiation protection and to leave to the various international and regional agencies such as IAEA, EURATOM and national regulatory bodies the responsibility of introducing detailed technical regulations, codes of pratice or laws suited to the needs of their member countries or specific countries. The present Commission of ICRP is assisted by 4 Committees working in the following specialized fields [99I1]: • Committee 1 (Radiation Effects) considers the risk of induction of cancer and heritable disease together with the underlying mechanisms of radiation action; also, the risks, severity, and mechanism of induction of tissue/organ damage and developmental defects. • Committee 2 (Doses from Radiation Exposures) is concerned with the development of dose coefficients for the assessment of internal and external radiation exposure, development of reference biokinetic and dosimetric models, and reference data for workers and members of the public.
1
A concise and consolidated summary is given by A. Nagaratnam [95N1] in his handbook on the salient features of the information given in ICRP Publications on the concept of radiological protection. Landolt-Börnstein New Series VIII/4
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• Committee 3 (Protection in Medicine) is concerned with protection of persons and newborn children when ionizing radiation is used for medical diagnosis, therapy, or for biomedical research; also, assessment of the medical consequences of accidental exposures. • Committee 4 (Application of the Commission's Recommendations) is concerned with providing advice on the application of the recommended system of protection in all its facets for occupational and public exposure; it also acts as the major point of contact with other international organizations and professional societies concerned with protection against ionizing radiation. In its 1950 recommendations [51I1], ICRP replaced the 1934 concept of “tolerance dose” [34I1] by that of the “maximum permissible dose” with the recognition that there could be risk even at these levels: “Whilst the values proposed for maximum permissible exposure are such as to involve a risk which is small compared to other hazards of life, nevertheless, in view of the unsatisfactory nature of much of the evidence on which our judgement must be based, coupled with the knowledge that certain radiation effects are irreversible and cumulative, it is strongly recommended that every effort be made to reduce exposure to all types of radiation to the lowest possible levels ... and that any unnecessary exposure be avoided”. According to the 1958 recommendations [59I1] the basic permissible dose to gonads, bloodforming organs, and lenses of the eyes for persons occupationally exposed at any age over 18 years was 5 rem (50 mSv) per year or weekly 0.1 rem (1 mSv), used for purposes of planning and design. No recommendation was made for exposure of individual members of the public but it was suggested that the per capita dose should not exceed 5 rem (50 mSv) per generation excluding medical exposures and exposures to natural background radiation. A linear non-threshold response was assumed for genetic effects. In 1962 it was recommended by the ICRP that the dose to individual members of the population at large should be limited to 0.5 rem (5 mSv) per year [64I1]. ICRP made it explicit that doses from natural background and from medical exposures were excluded from the maximum permissible doses. However, ICRP “recognizes especially the importance of the gonad doses resulting from medical exposure and the attendant genetic hazard to the population”, and recommended that “the medical profession exercises great care in the use of ionizing radiation in order that the gonad dose received by individuals before the end of their reproductive periods be kept at the minimum value consistent with medical requirements”. In 1977 [77I1] ICRP published epoch-making recommendations giving a new philosophical and conceptual framework of radiological protection. It is characterized by 1. Statement of the aim of radiation protection as being to prevent detrimental non-stochastic effects and to limit the probability of stochastic effects to levels deemed to be acceptable. 2. Formulation of the basic tenets of the system of radiation protection as a: Justification: No practice shall be adopted unless its introduction produces a positive net benefit. b: Optimization: All (necessary) exposures shall be kept as low as reasonably achievable, economic and social factors being taken into account (ALARA principle). c: Dose limitation: The dose equivalents to individuals shall not exceed the limits recommended for the appropriate circumstances by the Commission (limitation of the effective dose equivalent for stochastic effects in workers to 50 mSv per year, for non-stochastic effects in specific organs to 500 mSv annually; limitation of the effective dose equivalent to control the risk from stochastic effects of individual members of the public (critical groups) to 5 mSv in a year, and to 50 mSv annually for non - stochastic effects). Subsequent to the publication of the 1977 recommendations there have been clarifications and amendments, the most important ones at the 1985 meeting of the Commission [85I1]: Considering the effective dose equivalent limits for members of the public, made in its 1977 recommendations, “the Commission's present view is that the principal (stochastic) limit is 1 mSv in a year. However, it is permissible to use a subsidiary dose limit of 5 mSv in a year for some years, provided that the average annual effective dose equivalent over a lifetime does not exceeed the principal limit of 1 mSv in a year”. Landolt-Börnstein New Series VIII/4
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Apart from changes in terminology and definitions: • • • •
“non - stochastic effects” are now called “deterministic effects”, “quality factor” is replaced by “radiation weighting factor”, “dose equivalent” is replaced by “equivalent dose”, “effective dose equivalent” is replaced by “effective dose”
the 1990 recommendations of the ICRP in its Publication 60 [91I1] have brought down significantly the dose limits for occupational exposure from 50 mSv for the annual effective dose to 100 mSv in 5 years corresponding to an average of 20 mSv annually. (Additionally the effective dose should not exceed 50 mSv in a single year). The concept of justification, optimization and individual dose limits has been retained, however, a distinction is made between the systems of protection for proposed and continuing practices, and intervention: While, as in the past, the system of protection in practices is following the general principles of justification of a practice, optimization of protection (the magnitude of individual doses, the number of people exposed, and the likelihood of incurring potential exposures should all be kept as low as reasonably achievable, economic and social factors being taken into account), and limitation of individual dose and risk, an additional system of protection in intervention has been introduced. This is based on the following principles: 1. The proposed intervention should do more good than harm, i.e. the reduction in detriment resulting from the reduction in dose should be sufficient to justify the harm and the costs, including social costs, of the intervention. 2. The form, scale, and duration of the intervention should be optimized so that the net benefit of the reduction of dose and consequently of the detriment should be maximized. Regarding hereditary effects, ICRP Publ.26 [77I1] added the hereditary risk to the first and second generation offspring to the stochastic risk to the exposed individual, the effects in later generations being considered as part of the consequences for society. ICRP Publ.60 [91I1] now attributes the whole hereditary detriment to the detriment suffered by the exposed individual, thus avoiding the need for a two - stage assessment. Under the motto “Radiological Protection at the Start of the 21th Century” ICRP in 2002 has started an initiative which represents a genuine attempt to simplify the system of protection to one that is more coherent and easily explicable [02C1; 02C2]. Since classical cost-benefit analysis based on an utilitarian ethical policy answering the question “how much does it cost to reduce a dose and how many lives are saved?”, is unable to consider the individual, the Commission already modified the principle of optimization by the introduction of the concept of a constraint. Constraint is an individual-related criterion, applied to a single source in order to ensure that the most exposed individuals are not subjected to excessive risk, and to limit the inequity introduced by cost-benefit analysis. Although in the future the process of taking all reasonable action to reduce exposures is still likely to be called the Optimization of Protection, optimization is intended to be replaced by a different requirement. Namely, residual doses, after the application of Constraints, should be kept “as low as reasonably achievable” (ALARA). In this context the emphasis of constraints should provide a basic level of health protection for individuals exposed to a particular controllable source. Since there is likely to be some risk to health even at small doses introduction of a moral requirement is discussed for each controllable source to take all reasonable steps to restrict both the individual doses to levels below the action level and the number of exposed individuals. In this context it should be emphasized that these Constraints are not intended to be applied to justified medical exposures. Under the aspect “common sense would be often more important than formal application of differential equations in optimization” stakeholder involvement is discussed to determine or negotiate for the best level of protection in the circumstances. This means that whilst the dose constraints thus represent a basic standard of individual health protection, stakeholder involvement determines how far Landolt-Börnstein New Series VIII/4
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below the action level is “as low as reasonable practicable”, and will avoid the previous formal costbenefit analyses. Consequently ALARA would represent the optimum level of protection from the source under control or for an uncontrolled source.
References 28B1 Bureau of Standards, Circular No. 374: X-Ray and Radium Protection; Recommendations of the 2nd International Congress of Radiology, 1928; Br. J. Radiol.1, (1928), 359 34I1 International X-ray and Radium Protection Commission: Br. J. Radiol. 7, (1934), 1 51I1 International Commission on Radiological Protection: Radiology 56, (1951), 431; Br. J. Radiol. 24, (1951), 46 59I1 International Commission on Radiological Protection: Publ. 1, Pergamon Press, Oxford (1959) 64I1 Recommendations of the International Commission of Radiological Protection, as amended 1959 and revised 1962. ICRP Publ. 6. Pergamon Press, London (1964). 77I1 International Commission on Radiological Protection: Publ. 26, Annals of the ICRP 1 (3) (1977) 85I1 Statement from the Paris Meeting of the ICRP: Annals of the ICRP 15 (1985) 91I1 1990 Recommendations of the International Commission on Radiological Protection: Publ. 60, Annals of the ICRP 21 (1991) 95N1 Nagaratnam, A.: Defence Research and Development Organisation, Ministry of Defense, New Delhi - 110 011 (1995) 99I1 International Commission on Radiological Protection; Annual Report (1999): 24-06-2000 02C1 Clarke, R. H.: Int. Zeitschr. f. Kernenergie 47,1, (2002), 20 02C2 Clarke, R. H.: Strahlenschutzpraxis 8, 1, (2002), 45
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2 Biological effects of ionising radiation
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2 Biological effects of ionising radiation
This Chapter describes the effects of ionising radiation on the body. It covers both effects at the subcellular and cellular level as well as on the whole body. Acute effects can result from high radiation doses and in extreme cases can cause severe tissue damage and even death. For most exposures of people to radiation it is low doses that are of most concern. These can give rise to radiation-induced cancer in those exposed and hereditary disease in future generations. The chapter discusses the sources of information on radiation damage and includes estimates of risk for these different effects.
2.1 Introduction Within a few weeks of Wilhelm Conrad Roentgen’s discovery of X-rays on 8 November 1895, for which he received the Nobel Prize for physics in 1901, the potential of the technique for diagnosing fractures and other medical problems had become apparent, but acute adverse effects (such as hair loss, erythema and dermatitis) were also found. Similar undesirable effects were reported shortly after the discovery of radium (by Henri Becquerel in 1896) and its subsequent medical applications. In 1904, the first death of a person exposed to X-rays was reported; X-ray burns had developed into cancer. This death was soon followed by a steady stream of ‘martyrs to science through roentgen rays’ to use the title of a book by a radiologist who subsequently died of cancer. The widespread use of X-rays and radium in treating disease in the early 1900s led to the recognition of a cancer risk in many organs and tissues following high radiation doses which caused gross tissue damage. There was, however, a delay of about 40 years before it became clear that there was a risk of radiation-induced cancer from irradiation at lower doses and that there is no apparent threshold dose below which exposure to radiation can be considered safe. This delay can be attributed to the fact that radiation-induced cancers do not differ in any known way from those occurring naturally or caused by other agents. For many cancers there is also a long interval between exposure and the appearance of the tumour. It is now believed that any radiation dose, whether from external radiation or from incorporated radionuclides, is capable of inducing cancer and that the probability of its occurrence, but not its severity, depends on the radiation dose. Animal studies have shown that an increased incidence of certain types of inherited disorders can also occur in the descendants of irradiated parents. For both cancer and inherited disorders the probability of their occurrence, but not their severity, depends on the radiation dose. In radiological protection terminology they are termed stochastic effects. A second type of damage is seen after exposure of the whole or parts of the body to high doses of radiation between a few gray and a few tens of gray. It is a reflection of impairment of the functional capacity of tissues and is referred to as a deterministic effect. Severity of the damage is related to the extent of radiation exposure and it is assumed that there is a threshold below which the clinically detectable damage does not occur. If damage is extensive death may result. Following radiation exposure in utero serious mental retardation has been observed in the children of the atomic bomb survivors in Japan. Current evidence suggests this phenomenon is deterministic with a threshold related to the minimum shift in intelligence quotient (IQ) that can be measured. This Chapter reviews the sources of information available on the response of the body to radiation damage, and considers the extent to which dose-response relationships can be determined and quantified. Landolt-Börnstein New Series VIII/4
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2.2 Cellular effects 2.2.1 Primary events following exposure to ionising radiation Ionising radiations, hereafter abbreviated to radiation(s), can be classified into directly or indirectly ionising. Charged particles such as alpha particles and beta particles emitted from radionuclides are directly ionising if they have sufficient kinetic energy to disrupt atomic structure. Other types of radiation such as X-rays (generated artificially) or gamma rays (from nuclear transitions) are indirectly ionising. When passing through matter, they give up their energy to the atoms with which they collide and high velocity charged electrons are ejected from these atoms leaving behind positive ions. These electrons move randomly along a trajectory and may ionise other atoms in their path. If this occurs, more electrons are ejected, while the incident electrons continue on their trajectory with decreased energy and velocity, having transferred some of their energy to the newly formed electrons and eventually come to rest. Neutrons also lose energy in various ways, an important means being through collisions with hydrogen nuclei, which are single protons. The protons are set in motion and, being charged, they again deposit energy through electrical interactions. The unique feature of ionising radiation, then, is the localised release of energy in sufficient amounts to alter atomic and molecular structure. The particle track is the ensemble of ionisations (and excitations) along the trajectory of the electron or proton. One way of expressing the amount of atomic disruption is to quote the average energy loss along the track. This is referred to as the unrestricted linear energy transfer (LET or L). LET quantities are given in terms of average energy lost per unit path length, expressed in terms of kiloelectronvolts per micron (keV µm−1). This physical quantity has been used extensively in experimental radiobiology as a simplistic approach in order to relate the quality of radiation to cellular damage [91I2]. The rate of energy loss in biological material can vary greatly along the particle track depending upon kinetic energy and charge. In general terms, photons and electrons have LET values in the range of about 0.2 to 10 keV µm−1; for example, 1 MeV, 100 keV, 10 keV and 1 keV electrons have LET values of 0.2, 0.5, 2, and 10 keV µm−1 respectively. Protons, alpha particles and neutrons have LET values between about 10 and 100 keV µm−1; and heavy charged particles (e.g. nuclei of elements such as C, Ne and Si) can have still higher values to about 2000 keV µm−1. LET does not address the magnitude of the individual energy-loss events that occur along the track; nor does it address the amount of energy lost to matter in the volume of interest. This can be expressed as mean lineal energy which, in concept is more meaningful than LET [93I4]. The random nature of the particle track can be simulated by computer analysis using Monte Carlo techniques. A two-dimensional clustering of ionisations is shown in Fig. 2.1. This is only an approximation of the more complex three-dimensional events that involve random clustering of ionisations on a sub-atomic scale. Nevertheless, the figure illustrates the concept that low energy electrons are sparsely ionising because the ionisations are well separated spatially. Alpha particles, in contrast, are densely ionising because the ionisations are closely packed together along the track. It has been calculated that a single particle track of low-LET radiation (e.g. 1 MeV gamma-rays) passing through an 8 µm diameter spherical nucleus delivers an absorbed dose of about 1 mGy [94G2]. The gamma-rays are about one hundred times less damaging than high-LET radiation, for example 1 MeV neutrons which deliver an absorbed dose of a few hundred mGy in the same shape of nucleus. Each ionisation can result in energy being deposited within the atoms of a target molecule in sufficient amounts to disrupt chemical bonds. Alternatively, it may indirectly break the chemical bonds in a nearby molecule. It is the predominant reaction in water molecules in cells after exposure to X-rays. Free hydroxyl and other related radicals are produced and during their short existence of about a microsecond, these highly reactive radicals are capable of diffusing a few micrometres to reach and damage a target molecule such as deoxyribonucleic acid (DNA).
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Fig. 2.1. Simulated low energy electron track (upper: initial energy 500 eV) and simulated short portion of an alpha particle track (lower: 4 MeV). Large circles are ionisations, small circles are excitations. A Section of DNA is shown to give a perspective on dimensions; [94G2].
2 nm
The temporal sequence of ionisations (and excitations) leading to biological effects is illustrated in Fig. 2.2. Physico-chemical events are completed rapidly, the repair of damage may be completed within tens of minutes while effects in cells can arise within hours or days. The biological manifestations in multi-cellular organisms, including man, can be delayed for many years or, as in the case of hereditary disease, only be manifest in future generations. Tissue and whole body effects Cellular effects
Enzymatic actions ( repair/fixation of damage ) Formation of radicals and radical interactions Ionisations and excitations
1 min
1 year
Fig. 2.2. Timescale of events leading to radiation effects following exposure to ionising radiations.
Seconds
2.2.2 Cellular damage and repair following the primary radiation events It is widely accepted that the most important cellular constituent to be damaged by radiation is nuclear DNA. The molecular structure consists of a double helix (Fig. 2.1), formed from two complementary strands of nucleotides. These are purine and pyrimidine bases linked to sugar molecules with phosphate Landolt-Börnstein New Series VIII/4
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molecules joined by ester linkages. The two strands are held together by hydrogen bonds between guanine-cytosine and adenine-thymine base pairs. The cell’s genetic information is carried in a linear sequence of nucleotides that make up the estimated 100,000 genes in the human genome. Each gene controls a discrete characteristic. Just as cells inherit genes, they also inherit a set of instructions that tell the genes when to become active. These gene regulatory proteins recognise short stretches of nucleotide sequences on the double helix and determine which of the genes in the cell will be transcribed. About two-thirds of genes provide instructions for cell division and for the synthesis of tens of thousands of proteins that provide the structural components of cells, as well as numerous enzymes promoting and controlling cellular activity. Ribonucleic acid (RNA) is the molecule that helps to transport, translate and implement the coded instructions from the genes in the nucleus to the body of the cell. All cell types contain the same genes, but encoding sets of genes is cell-specific. This uniqueness ensures that cells in each tissue produce their own proteins. Maintaining stability in the gene is essential for cell survival. This stability requires not only extremely accurate mechanisms for DNA synthesis and replication, but also mechanisms for repairing DNA damage before replication. Observations with proliferating cells in the laboratory indicate that DNA is subjected to only an estimated few tens of base-pair or nucleotide permanent changes per year during normal metabolism, despite the fact that metabolic processes alter thousands of bases and nucleotides every day. DNA single strand breaks, without base involvement, are effectively ligated enzymatically. Base excision repair pathways require different groups of enzymes that identify and excise the damaged base site, make a complementary copy of the information bases on the opposite undamaged strand, and seal the correct sequence of copied bases in the gap. If nucleotide damage occurs, nucleotide excision repair pathways are able to repair the more extensive damage on one strand. Once the lesion is identified along the strand, the damaged nucleotides are removed and repair proceeds thereafter as for base damage. DNA double strand damage with or without base damage, occurs much less frequently than damage to single strands during normal cellular activity. Recombination repair pathways are available, but they are not totally effective, since there is no undamaged strand to act as a template for base or nucleotide replacement. Damage to bases can result in their alteration or loss. When the repair processes fail, the resulting misrepair is referred to as a mutation. DNA damage due to radiation causes similar lesions to those occurring after normal metabolism, but double strand breaks, multiple gene losses and the translocation of gene sequences occur more frequently as a dose-related effect. The probability of misrepair is greater under these circumstances. Estimated yields of damage caused by low-LET radiation are shown in Table 2.1 [88W1]. Table 2.1. Examples of damage in a mammalian cell nucleus from 1 Gy of low-LET radiation. Initial physical damage Ionisations in cell nucleus ~ 100,000 Ionisations directly in DNA ~ 2,000 Excitations directly in DNA ~ 2,000 Selected biochemical damage DNA single-strand breaks ~ 1,000 Base (8-hydroxyadenine) damage ~ 700 Base (thymine) damage ~ 250 DNA double-strand breaks ~ 40 DNA-protein cross links ~ 150 Selected biochemical damage Lethal events ~ 0.2-0.8 Chromosome aberrations ~ 0.4 Hprt(1) gene mutations 0.6 × 10−5 Translocation frequency (2 loci) 1.2 × 10−4 (modified from 88W1) (1) hypoxanthine-phosphoribosyl transferase Landolt-Börnstein New Series VIII/4
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Recent investigations have revealed that DNA repair pathways may work in conjunction with other intracellular activities in order to minimise cell damage. These include delay in cell-cycling (as a means of maximising the chances of repair); and programmed cell death (apoptosis), whereby severely damaged cells are eliminated to stimulate cell proliferation.
2.2.3 Classification of radiation-induced damage Laboratory techniques have been available for many years to observe radiation effects in proliferating cells. These techniques include measuring changes in cell survival, in the frequency of chromosomal aberrations (deletions and translocations), in gene structure (mutations), and in oncogenic transformation (neoplasia). 2.2.3.1 Cell survival Cellular damage can be classified into three arbitrary categories: lethal damage which results in cell death; sublethal damage, which may be repaired; and potentially lethal damage, defined as damage that can be repaired by altering the growth conditions as for cells in culture. Cell lines of fibroblasts from rodent and human tissues have been used extensively to establish doseresponse relationships [93U6]. Expressed graphically as the logarithm of cell survival plotted against absorbed dose on a linear scale, the dose-response is linear for low-LET radiation at low doses, followed by a curvature at higher doses. Expressed mathematically, the relationship can be represented by a linear-quadratic equation: S = e − (α D+ β D
2
)
(1)
where S is the surviving fraction after exposure to dose D and α and β are coefficients representing the linear and quadratic components for cell killing. The initial slope of the relationship is determined by α, while the quadratic component, reflects the curvature in the dose-survival relationship (Fig. 2.3). The dose at which the linear and quadratic components are equal is the ratio of α and β. The response to highLET radiation is also shown in Fig. 2.3 where survival is best expressed as a linear function of dose passing through the origin. 1 aD b D2
Cell survival
10 -1
10 -2 High LET 10 -3 0 Landolt-Börnstein New Series VIII/4
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a/b 4
8 Dose D [Gy]
12
16
Fig. 2.3. Typical survival curves for cultured cells exposed at high dose-rate (>0.1 Gy min−1). The curves illustrate the linear-quadratic relationship for low-LET radiations and linear relationship for high-LET radiations.
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A plausible explanation of the linear component following exposure to low-LET radiation at low doses is that the majority of DNA interactions are single particle track events [94G2]. Under these circumstances, DNA damage can be effectively repaired. As the dose increases, multi-track events reflecting the quadratic component, and which are associated with clustered DNA damage, increasingly predominate with a consequent increase in the probability of misrepair and lethal events. At 1 Gy, for example, lethal events have a frequency of about 0.2 to 0.8 per cell (Table 2.1). Protracted exposure to low-LET radiation results in less damage, per unit of dose, compared with acute exposure [93U6, 00U8]. This is referred to as the dose rate effect and is due to the ability of cells to repair more sublethal damage as the dose rate is reduced. Below about 1 Gy min−1, the slope on the exponential portion of the survival curve typically becomes progressively shallower as more and more sublethal damage is repaired. Below about 0.01 Gy min−1, undamaged cells are able to proliferate at a sufficient rate to offset the reduction in cell numbers while repair is progressing. This response is illustrated in Fig. 2.4. A dose rate effect is not observed after exposure to high-LET radiation, suggesting little repair of damage. 1
Surviving fraction
0.004 Gy min -1 10 -1 Proliferation
0.01 Gy min -1
10 -2
Repair
1 Gy min -1
10 -3 0
4
8 Dose D [Gy]
12
16
Fig. 2.4. Dose-rate effect showing the influence of repair and repopulation on the dose-survival relationship for cells.
The relative biological effectiveness (RBE) of different types of radiation is defined as the ratio of a dose of a reference low-LET radiation to a dose of the test radiation that gives an identical biological endpoint [90N3]. RBE values are influenced by variations in LET, dose and dose rate. RBE values increase to a maximum at about 100 keV µm−1, decreasing thereafter because of an “overkill effect”. The absolute value of the RBE is not unique but depends on the level of biological damage and, therefore, on the absorbed dose [86B1]. For irradiation by alpha particles, for example, the RBE is generally taken to be 20 for stochastic effects (cancer and hereditary disease) but to have a lower value of around 5 for deterministic effects. 2.2.3.2 Damage to viable cells Chromosome aberrations and gene mutations The technique of culturing human lymphocytes in vitro has been available for many years. It provides a means of measuring the frequency of unstable and stable chromosome aberrations at various stages in the cell-cycle. In terms of unstable aberrations, their frequency increases from a background level of about 10−3 to a rate of about 4 × 10−2 Gy−1 after exposure to low-LET radiations. Dose-response relationships for different types of radiation are illustrated in Fig. 2.5 [89E1]. Neutrons are more damaging than X-rays Landolt-Börnstein New Series VIII/4
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or gamma-rays and low energy neutrons are more damaging than high energy neutrons. For low-LET radiations a linear-quadratic relationship is consistent with the data. That is: E = αD +βD2
(2)
where E is the frequency of chromosome aberrations (i.e. a stochastic effect), D is the dose, and α and β are the linear and quadratic coefficients for the induction of the aberrations. 1.0
Dicentrics per cell
Fission neutrons
0.5
250 kVp X - rays
0 0
1
2
3
Fig. 2.5. Dicentric yield in chromosomes per cultured human lymphocyte as a function of dose for selected radiations; [89E1].
Dose D [Gy]
A number of specific-locus mutation test systems using mouse, hamster and human fibroblasts have been developed to measure mutagenesis. One cell line, the human B-lymphoblastoid TK6, illustrates the use of the test [89K4]. Cultured cells were exposed to radiation and the mutation frequency at two loci (hgprt and tk) was measured under different exposure conditions. For acute radiation exposure, 100 kVp X-rays (0-2 Gy) and (Pu, Be) neutrons (0-0.2 Gy) both showed a linear dose-response relationship in terms of induced mutants. The induced mutant frequency per 0.01 Gy per surviving cell was 0.55 × l0−7 (CI 0.09) and 1.92 × l0−7 (CI 0.03) respectively. Protracted exposure to X-rays (0.01-0.1 Gy per day) for 5 to 20 days showed a slight increase in the mutation frequency (0.84 × l0−7 (CI 0.17)); while continuous exposure to neutrons (0-0.4 Gy) resulted in a substantial increase (6.00 × l0−7 (CI 0.7)). These data demonstrate an ‘inverse dose-rate effect’ for neutron-induced mutation in human cells. Syrian hamster embryo cells showed a similar effect, but other cell lines did not. It is concluded that there are a number of difficulties in interpreting the results of somatic cell mutations. Estimated yields of chromosome aberrations and mutation frequency are shown in Table 2.1. Cell transformation An established technique for studying carcinogenic potential is that of culturing cells that can grow indefinitely, provided that they are frequently transferred to fresh media. Under specific conditions, cells that have acquired this ability are said to be immortalised. A characteristic of these immortalised cells is that they stop dividing when they come into contact with similar cells in the culture medium (contact inhibition). They are not classified as malignant cells because they do not cause tumours when injected into immunologically-suppressed animals. Occasionally, an immortalised cell undergoes a spontaneous change, whereby it loses its contact inhibition and continues to proliferate by spreading over adjacent immortalised cells to form a recognised foci of cells. Such cells are said to have undergone Landolt-Börnstein New Series VIII/4
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transformation and when they are injected into animals, they develop into tumours. Spontaneous transformation is a rare event, occurring at a rate of about one in ten thousand to one in a hundred thousand per surviving cell. The mechanism is not fully understood, but it is thought to involve the mutation of two or more genes. Two classes of mutated genes in particular have been identified and characterised. These are ‘gain-of function’ mutations of proto-oncogenes, whereby the mutated genes (oncogenes) stimulate cell proliferation in an uncontrolled manner; and ‘loss-of function’ tumour suppressor genes, whereby cells are no longer prevented from proliferating in defiance of normal controls. The cell types used in transformation studies are mainly derived from fibroblasts. It is generally accepted that the sensitivity of the test is low, the detection limit being about 0.25 Gy. Ideally, human epithelial cells would be a better choice to represent human cancers. Future studies are in hand which aim to use this type of cell. To illustrate the technique, C3H/1OT½ fibroblasts derived from the prostate of the C3H mouse embryo, were irradiated with low-LET radiation [98M2]. Cell survival and transformation frequencies were simultaneously measured. The survival curve was consistent with a linear-quadratic dose-response relationship (Equation 1), while the transformation frequency per surviving cell following exposure to X-rays was consistent with a linear relationship (Fig. 2.6a). However, if the number of transformants per cell at risk was plotted, the relationship to intermediate doses was consistent with a linear-quadratic equation, the transformation frequency reaching a maximum at about 2 Gy (Fig. 2.6b). This doseresponse relationship is consistent with other results reported in the literature, although the maximum transformation frequency was usually in the 3 to 4 Gy range. 6
2.0
Transformation frequency [×10 -4 ]
Transformation frequency [×10 -4 ]
5 4 3 2 1
1.6
1.2
0.8
0.4 Cells at risk
Cells viable for survival 0 0
a
1
2 3 4 Absorbed dose D [Gy]
5
0
6
0
b
1
2 3 4 Absorbed dose D [Gy]
5
6
Fig. 2.6. Transformation frequencies per surviving C3H 10 T½ cell (a) and per cell at risk (b) as a function of absorbed dose after exposure to 250 kVp X-rays at 2 Gy min−1; [98M2].
Exposure to neutrons resulted in a higher transformation frequency than for low-LET radiation, with no evidence of a dose rate effect. One exception was a study of 5.9 MeV or fission neutrons where an inverse dose-rate effect was reported [93U6]. It is concluded that there are difficulties in interpreting data on cell transformation studies. Generalised dose-response relationships The conventional approach to representing the absolute biological effectiveness of a given radiation at low doses is based on the assumption derived from target theory in which the induction I of an effect as a function of dose D can be represented by Landolt-Börnstein New Series VIII/4
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I (D )= α1 D + β1 D 2 e − (α 2 D+ β 2 D
(
)
2
)
2-9
(3)
in which α1 and β1 are single and multihit components for a radiation effect and α2 and β2 represent single and multihit components for cell killing. At low doses the incidence from effect is determined by α1 with the response increasing linearly with dose. It is generally assumed that in this region α1 will be independent of dose rate. With increasing dose the amount of damage due to multihit effects increases, resulting in a quadratic component in the dose-response curve. At doses above a few gray β1 and β2 become significant resulting in a reduction in tumour yield due to the effect of cell killing. For high-LET radiation (α particles, neutrons) the dose-response curve is generally found to be linear up to the point at which cell killing starts to exert an effect and reduces the tumour yield.
2.2.4 Implications of cellular damage for whole or partial body exposure The outcome of cell damage in terms of human radiation detriment can be profoundly different according to the exposure conditions. Cellular studies can provide a sound basis upon which to judge these outcomes. After acute exposure to absorbed doses above a few gray, the cells at greatest risk are selfreplicating stem cells that supply functional cells. They are programmed to divide so that one daughter cell remains as a stem cell (in order to ensure that stem cell numbers in the tissue remain constant), while the other daughter cell proceeds to specialise (differentiate) by clonal expansion. If sufficient numbers of stem cells in a tissue are killed or are prevented from dividing at the appropriate rate, the tissue loses its ability to function effectively. The consequential effects are referred to as deterministic. Studies have established that cell survival is dose and dose-rate dependent for low-LET radiations, and that there is a tissue-specific dose threshold. At high risk are rapidly dividing bone marrow stem cells, and stem cells in the epithelium of the gastrointestinal tract, lungs, thyroid, gonads, skin and lens of the eye. The effects due to the proliferation of mutated cells at low doses are termed stochastic. There is sufficient radiobiological evidence for low-LET radiation to support the general assumption of an increasing risk of an effect with increasing dose at low to intermediate doses, with no threshold. Cellular techniques are providing insight into the way in which radiation can initiate the complex multistage process of carcinogenesis. However, there is still much to be learned about the molecular changes that lead to cells with the potential towards malignancy; and most importantly, any advances in knowledge at the cellular level have to be seen in the context of the living organism.
2.3 Deterministic effects 2.3.1 Tissue and organ development In the space of a few weeks, a single fertilised human egg gives rise to a complex multicellular organism consisting of embryonic cells arranged in a precise pattern, each in its proper place. In the subsequent period of fetal growth, the cells continue to proliferate in the developing tissues and organs. Growth of tissues and organs continues in childhood with increase in cell mass in many tissues, but growth essentially ceases in the adult when cell masses reach a predetermined size. The majority of cells in tissues of the adult are differentiated, that is, they have developed specific morphology and function which is usually irreversible, but these cells are predestined to die. In many tissues of the body, the rate of death of differentiated cells is rapid and, in a healthy state, must be balanced by proliferation from stem cells. These cells, by definition, are cells that have retained Landolt-Börnstein New Series VIII/4
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embryonic characteristics. They are able to divide during the lifetime of the organism, yielding progeny that are destined to differentiate by a process of clonal expansion. Stem cells also retain the ability of selfrenewal. These characteristics are illustrated in Fig. 2.7. The number of stem cells compared to differentiated cells varies according to the tissue, but they usually represent, at most, a few percent of the total cell numbers. Furthermore, only a small fraction of the stem cells are active at any one time under normal circumstances. It is not known how the balance between cell proliferation and cell death is achieved, but it is thought that all cells are genetically programmed to die, by apoptosis. When differentiated cells die, a feed-back mechanism is activated to stimulate the stem cells to divide and replenish the population.
Stem cell
S
Self - renewal
Clonal expansion Maturation without division Functional cells with finite lifetime
Fig. 2.7. Derivation of differentiated cells from a selfrenewing stem cell.
If enough stem cells in a tissue are killed or prevented from undergoing cell division, there will be loss of tissue function; termed deterministic by the International Commission on Radiological Protection (ICRP). The dose-response relationship is characterised by a frequency and severity that increases with dose above a threshold. Most tissues and organs of the body are able to compensate for small reductions in the number of differentiated cells. But if the decrease is large enough, there will be changes seen as loss of tissue or organ function and a consequential response to repair the damage.
2.3.2 Dose-response relationships for radiation damage The probability of detecting loss of tissue or organ function following exposure to radiation increases steeply above a threshold dose to a maximum. Expressed as a generalised dose-response relationship, the plot of the frequency of the effect versus dose expressed on linear axes is sigmoid (Fig. 2.8, upper panel). Above the threshold dose, the severity of the effect also increases with dose reflecting more cell loss and hence damage to tissue function (Fig. 2.8, lower panel). Protracting the dose results in a lower frequency and less severe symptoms at a given dose compared with acute exposure, demonstrating the importance of stem cell repopulation. There is individual variation in radiosensitivity in any exposed population. This variation reflects differences in the ability of individuals to cope with radiation-induced cellular damage. Any response is influenced by the age and state of health of the exposed individual.
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Frequency [%]
100
50
0
Seveity
Variation in population
Threshold
Fig. 2.8. Dose-response relationship for deterministic effects. Variation in frequency and severity; [based on 91I2]. Dose
2.3.3 Deterministic effects in humans following acute whole-body irradiation Evidence of the deterministic effects of radiation comes from several sources. These include retrospective studies on radiotherapy patients, radiologists in the early part of the 20th century, Japanese populations exposed to radiation from atom bombs, and individuals accidentally exposed to high doses following nuclear reactor accidents and radiographic sources. Understanding the effects of acute high doses is important as an aid to prognosis in the treatment of accidental over-exposure, and to ensure that deterministic effects are avoided in normal practices and minimised in accidents. Evidence on deterministic effects also comes from studies with animals. After exposure to doses of a few Gy, the depression in the numbers of circulating white blood cells (granulocytes) and blood platelets may be so severe as to result in death from septicaemia (infection) and haemorrhage. This is referred to as the haematopoietic syndrome. Recovery depends upon the radiation dose and the ability of the remaining stem cells in the marrow to recover. Loss and recovery of granulocytes and blood platelets follows a similar dose- and time-related pattern. Depression of the stem cells providing the protective mucosal cells lining the intestinal tract wall results in a denuding of the gut surface. This gastrointestinal syndrome is seen in individuals who have received doses to the gastrointestinal tract in excess of about 5 Gy. Leakage of blood from damaged blood capillaries results in severe anaemia and ingress of intestinal bacteria through the damaged blood vessels results in septicaemia. The haematopoietic syndrome will manifest itself concurrently at these higher doses. Landolt-Börnstein New Series VIII/4
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Damage to endothelial cells lining the alveolar air sacs may result in acute inflammation of the lungs (pneumonitis) at doses in the range 5 - 15 Gy. This may occur after radiotherapy or after the inhalation of high specific activity radioactive particles. If the individual survives the pneumonitis, lung fibrosis may later develop which can also be life-threatening through loss of lung function. At higher whole-body doses (>15 Gy), generalised shock occurs affecting the brain and the cardiovascular system. Coma and death develop rapidly thereafter. The range of doses associated with death from these syndromes after acute exposure to low-LET radiation is given in Table 2.2. The ranges are based upon human data, supplemented by knowledge of the form of the dose-response relationship derived from animal experiments. No individual would be expected to die after receiving absorbed doses below about 1 Gy. The dose range where half the exposed population would be expected to die without medical treatment is 3 to 5 Gy. Death would be likely at doses between about 6 Gy and 10 Gy, unless they receive treatment to prevent infection and bleeding. Above about 10 Gy death is assumed at present to be inevitable, even after attempts to stimulate the bone marrow or bone marrow transfusion from a suitable donor. These estimates of lethality do not take account of any concurrent radiation-induced damage (e.g. skin burns), or existing debilitating diseases. Table 2.2. Range of doses associated with acute radiation syndromes in adults exposed to low-LET radiation. Whole body absorbed dose 1-6 Gy 5-15 Gy >15 Gy
Principal effect contributing to death Damage to bone marrowa Damage to the gastrointestinal tract and lungsb Damage to nervous system and shock to the cardiovascular system
Time of death after exposure [days] 30-60 10-20 1-5
a) Dose range considered to result in 50 % of an exposed population dying (LD50) 3-5 Gy. b) Damage to vasculature and cell membranes especially at high doses is an important factor in causing death.
2.3.4 Deterministic effects following partial body irradiation 2.3.4.1 Tolerance doses in adults after radiotherapy Extensive experience in the treatment of patients undergoing radiotherapy has provided data upon which to determine the tolerability of healthy tissues and organs to radiation. Called the tolerance dose by clinicians, it is defined as the amount of radiation received during conventional treatment below which unacceptable effects do not occur in more than a few percent of patients within 5 years following treatment. The tolerance doses for some adult tissues are shown in Table 2.3 (children are usually less tolerant to exposure). It is evident that the gonads, lens of the eye and the bone marrow are the most radiosensitive. 2.3.4.2 Threshold doses in radiological protection The limitations of using data on tolerance doses to derive threshold doses for radiological protection purposes need to be recognised. In contrast to the precise exposure conditions of radiotherapy, exposure of workers to high doses of low-LET radiation is most likely to be non-uniform and resulting from mixed radiations. The tolerance dose therefore can at best be used as a cautious approximation to a threshold dose.
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Table 2.3. Tolerance doses for deterministic effects in adults after fractionated radiotherapy treatment Organ Effect Tolerance dose [Gy] Total bone marrow Blood cell depletion 1-2 Ovary Permanent sterilisation 2-6 Testis Permanent sterilisationa 3-4 5-10 Eye Cataractb Kidney Nephrosclerosis 23 Liver Loss of function, ascites 35 Lung Pneumonitisc 40 Heart Pericarditis 40 Lymph nodes Hypoplasia, fibrosis 35-45 Thyroid, pituitary Hypoplasia >45 Other organs Hypoplasia, fibrosis >45 a) A significant but reversible, depression of sperm count occurs after about 0.1 Gy brief exposure. b) About 2 Gy after a brief exposure. c) LD50 after brief exposure is about 10 Gy.
The threshold doses recommended by the ICRP for the most radiosensitive tissues and organs are summarised in Table 2.4. Thus the threshold dose for temporary sterility in the male for a single absorbed dose in the testes is about 0.15 Sv. Under conditions of prolonged exposure, however, the dose rate threshold is about 0.4 Sv y−1. The corresponding values for permanent sterility are about 3.5 Sv and 2 Sv y−1. The threshold dose for permanent sterility in women for a single absorbed dose is in the range from about 2.5 Sv. For protracted exposure, the dose rate threshold is about 0.2 Sv y−1. Clinically significant depression of the blood-forming process occurs above a single bone marrow dose of about 0.5 Gy. The dose rate threshold for protracted exposure is about 0.4 Gy y−1. The tolerance dose for death is in the range of 6 to 7 Gy if the radiation is spread over 30 fractions in a period of 6 weeks. Table 2.5 summarises the principal syndromes associated with whole body exposure. Table 2.4. Estimates of the thresholds for deterministic effects in adults recommended in radiological protection [91I2]. Equivalent dose rate Equivalent dose brief Tissue and effect exposure [Sv] protracted exposure [Sv y−1] Testes Temporary sterility 0.15 0.4a Permanent sterility 3.5-6.0 2.0 Ovaries Sterility 2.5-6.0 >0.2 Lens Detectable opacities 0.5-2.0 >0.1 Visual impairment (cataract) 5.0c >0.15 Bone marrow Blood cell depletion 0.5 >0.4b a)
This dose is higher because differentiating cells are more radiosensitive than the stem cells so the latter can replenish the differentiating cells at an adequate rate. b) Supported by evidence of effects after chronic radiation of Beagle dogs. c) Range 2-10 Sv.
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Table 2.5. Summary of acute radiation syndrome Syndromes Intestinal Cerebral
Bone marrow
Critical organ Latent period Syndrome threshold [Gy] Death threshold [Gy] Death occurring within Cause of death
Bone marrow 2-3 weeks 1 2 3-8 weeks Haemorrhage, infection
Prodromal vomiting Symptoms
Brain 20 min 20 50 2 days Cerebral oedema, heart failure Minutes Tremors, cramps, loss of coordination, lethargy, impaired vision, coma
Treatment
Palliative
Prognosis
Hopeless
Small intestine 3-5 days 3 10 2 weeks Sloughing of gut, shock
1 hour A few hours Loss of appetite, vomiting, Fever, breathlessness, internal diarrhoea with bleeding, bleeding, depletion of bone fever, electrolyte and fluid marrow leading to low blood balance counts Barrier nursing, fluid and electrolyte replacement, transfusions of blood cells, bone marrow transplants Very poor Dose-dependent and influenced by treatment
2.3.4.3 Skin irradiation Based upon extensive experience in the use of fractionated X and gamma radiation in radiotherapy, (typically, 20 to 30 fractions each of 2 to 6 Gy over several weeks), various degrees of skin damage can be observed according to the area and depth of skin involved, the absorbed dose and the duration and frequency of the exposure. The earliest observable change is a transient reddening within a few hours after exposure to doses above about 2 Gy; due to increased capillary permeability. This is followed after moderate doses (about 5 Gy) two to four weeks later by a persistent reddening (the main erythematous reaction) and peeling of skin (dry desquamation). This is due to secondary inflammation resulting from the death of basal (stem) cells of the epidermis. Hair loss also occurs. At higher doses (about 20 Gy), blistering (moist desquamation) occurs after about four to six weeks due to the inability of basal cells in the irradiated area to divide and for viable basal cells to migrate into the area at a sufficiently rapid rate. It is the health effect to be avoided in both radiotherapy and radiation protection practice. The threshold doses for moist desquamation depends upon the area irradiated and the penetrating powers of the radiation. Ulceration is the result of infection following moist desquamation and may occur after about 6 weeks. Necrosis due to irreversible damage to the basal cells of the dermis and the underlying blood vessels occurs within two to three weeks after doses of tens of Gy. Late effects developing months to years later include changes in pigmentation; atrophy of the epidermis, sweat glands and sebaceous glands and hair follicles; and fibrosis. Quantifying the threshold doses for these effects is complicated in practice by the multiplicity of targets at different critical cell depths, which makes it difficult to select a single depth at which to specify the dose to the skin. The depths at which the most serious effects arise are estimated to be in the range of 300 - 500 µm. However, a conservative approach for protection purposes is to use shallower depths (20 100 µm, typically 70 µm) for monitoring specifications. To prevent moist desquamation, the dose must be reduced as the radiation field is increased. To illustrate the importance of field size, the tolerance doses following a single treatment with orthovoltage X-rays was found to be 20 Gy for an area of 6 × 4 cm, and 11 Gy for an area of 15 × 20 cm. Following fractionated treatment, the tolerance doses were estimated to be about 50 Gy and 30 Gy respectively for the two field sizes. From experimental studies, the estimated dose threshold following exposure of large areas of skin is about 20 Gy; and no acute tissue breakdown was observed at a dose rate of 0.4 Gy h−1 with total doses of about 100 Gy. Accidental over-exposure of industrial radiographers is a cause for concern in radiation protection. In normal practices, ICRP recommends a limit on effective dose of 20 mSv per year, averaged over 5 years with the further provision that the effective dose should not exceed 50 mSv in a single year [91I2]. This limitation is on effective dose and is assumed to be adequate to prevent deterministic effects. However, an additional annual limit is recommended for localised exposures in order to prevent deterministic effects to the skin. It is 500 mSv averaged over any 1 cm2 regardless of the area exposed. Landolt-Börnstein New Series VIII/4
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2.4 Radiation-induced cancer 2.4.1 Cancer development The development of cancer is the major late effect resulting from exposure to radiation. Cancer is generally understood to develop in a number of stages. That is, for malignancies to be expressed a series of events must occur in cells and the rate at which they occur is thought to be reflected in the way cancers appear in the population over the course of time. The development of cancer in tissues is a complex, multi-stage process that can be sub-divided into four phases: neoplastic initiation; neoplastic promotion; conversion and progression. The sub-divisions are necessarily simplifications of the overall process which is, in any event, somewhat variable between different tumour types. However, they do provide a basis from which to interpret the cellular and molecular changes involved [93U6, 00U8]. Neoplastic initiation encompasses the essentially irreversible cellular damage, which although not necessarily expressed immediately, provides the potential in cells for the development of cancer. There is good evidence that this initiation process results from damage to DNA leading to gene mutations in single target cells in tissues. The critical damage is likely to be coincident damage to both DNA strands (DNA double strand breaks, Section 2.2.2). Although a proportion of such double strand damage will be repaired, completely error free repair of such damage, even at low doses, is not expected. Neoplastic promotion can be seen as a process whereby initiated cells receive an abnormal growth stimulus and begin to proliferate in a semi-independent manner. Conversion of these pre-neoplastic cells to a form in which they are committed to become fully malignant is a central feature of the process of neoplastic development. Such changes are now believed to be driven by further gene mutations accumulating within the expanding population of pre-neoplastic cells. Once the potential for full malignancy has been established, the subsequent progression of the disease may depend upon further cellular changes that allow invasion of adjacent normal tissues, the circulation of neoplastic cells in the blood and lymphatic systems and the establishment of metastases (secondary tumour growths) at other sites in the body. It is this invasive process that provides principally for the fatal effects of most common human tumours. On this basis, a single mutational event in a critical gene in a single target cell in vivo can create the potential for neoplastic development. Thus, a single radiation track traversing the nucleus of an appropriate target cell has a finite probability, albeit very low, of generating the specific damage to DNA that results in a tumour initiating mutation. These initiated cells can then develop by multistage processes into an overt malignancy. As a consequence, at the level of DNA damage, there is no basis for assuming that there is likely to be a dose threshold below which the risk of tumour induction would be zero. For radiation protection purposes, a progressive increase in risk with increasing dose, with no threshold, is therefore assumed [95C2]. Whilst such a multistage mechanism is considered to be the cause of many human tumours there are likely to be some tumours that may arise in tissues where there has been deterministic damage (fibrosis) for such tumour types a threshold dose may need to be exceeded before the tumour will occur. There are many examples of such tumour types in animals and the development of radiation-induced bone tumours in man may also require a threshold dose to be exceeded [00U8]. Radiation appears to be capable of causing tumours in nearly all tissues of the body, although the frequency of appearance following a unit dose may vary markedly from one tissue to another. Information on the dose related frequency of tumour induction by radiation is gained through follow-up of groups of persons exposed to radiation. The observed tumour frequency can then be compared with an age and sex matched control group, not exposed to radiation, to determine the increase in frequency due to the radiation exposure. Extensive follow-up studies have been carried out on groups of persons exposed to either external radiation or to internally incorporated radionuclides.
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Tumours induced by radiation are in general indistinguishable from those occurring spontaneously and since cancer is not uncommon (about one in five die as a result of it in Western Europe and North America), the problem of determining a relatively small excess due to radiation exposure is difficult. In general, large exposed populations are necessary to obtain statistically meaningful results. The chief sources of information on the risks of radiation-induced cancer are the A-bomb survivors exposed to whole-body irradiation in Hiroshima and Nagasaki, patients with ankylosing spondylitis and other patients who were exposed to partial-body irradiation therapeutically, either from external radiation or internally incorporated radionuclides, and various occupationally exposed populations, such as uranium miners and radium-dial painters. Some quantitative information on thyroid cancers may also be obtained following the Chernobyl accident. Increasingly information is becoming available from epidemiological studies on groups of persons occupationally exposed to radiation. In general, however, the radiation exposures in these populations is relatively low and there is limited power in the studies to obtain quantitative estimates of risks of radiation-induced cancer. Reports by the United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) provide a comprehensive review of the data available [94U7, 00U8]. In it’s 1990 recommendations, the ICRP re-assessed the epidemiological data and this resulted in an increase in estimates of the lifetime risk of radiation-induced cancer. Partly, this arose as a result of revised dosimetry for the A-bomb survivors and a longer follow-up of the population, but mainly it was attributed to a change in the model now used to project lifetime risks [91I2]. Similar calculated values of lifetime risk have been published by UNSCEAR [94U7, 00U8].
2.4.2 Dose-response relationships 2.4.2.1 Assessment of lifetime risk There is always a minimum period of time between irradiation and the appearance of a radiation-induced tumour. This period is termed the latent period and its length varies with age and from one tumour type to another. Some types of leukaemia and bone cancer have latent periods of only a few years but many solid tumours have latent periods of ten or more years. For leukaemia and bone cancer there is fairly good evidence that the risk is almost completely expressed within about twenty-five years following exposure. For solid tumours of longer latency, such as those of the GI tract , liver and lung it is not yet clear whether the incidence of these tumours passes through a maximum and declines with time following exposure, whether the risk levels out, or alternatively increases indefinitely during the remainder of life. To project the overall cancer risk for an exposed population, it is therefore necessary to use models that extrapolate over time data based on only a limited period of the lives of the individuals. Two such projection models have generally been used: (a) the additive (absolute) risk model which postulates that radiation will induce cancer independently of the spontaneous rate after a period of latency, variations in risk may occur due to sex and age at exposure as well as the tissue exposed. (b) the relative (multiplicative) risk model in which the excess (after latency) is given by a constant (or time-varying) factor applied to the age dependent incidence of natural cancers in the population. In most cases the spontaneous risk of cancer increases with age and therefore the relative risk model will predict an increasing incidence of radiation-induced cancer with age. This model also gives different risks of radiation-induced cancer in different populations, depending on the national cancer incidence. Data available from the A-bomb survivors in Japan and from studies on uranium miners suggest the relative risk projection model gives a better fit to the data, at least for some of the most common cancer types. Despite this there are indications from a number of exposed groups that the risk of cancer starts to decline many years after exposure. This has been well documented for leukaemia, but has also been Landolt-Börnstein New Series VIII/4
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observed in the case of bone cancers (patients in Germany given 224Ra), thyroid cancers (US follow-up study after thymus irradiation), solid cancers (patients treated for ankylosing spondylitics) and possibly lung cancers in the uranium miners exposed to radon and its decay products [88U5; 00U8]. These results suggest that for the Japanese population the excess risk may ultimately decrease with time and thus relative risk projection models applied over a lifetime could result in an overestimate of the cancer risk. 2.4.2.2 Effects of dose and dose rate The total radiation dose and the dose rate both influence cancer induction and are linked to the form of the dose-response relationship. For radiological protection purposes tumour induction is generally assumed to increase with increasing dose, with no threshold, as indicated above. However, studies using cells in culture reveal that for many endpoints, including mutation, the dose-response for exposure to low-LET radiation is not linear, but that the effectiveness of radiation, per unit dose, increases as the dose increases. At very low doses, where there is a low probability of more than one radiation event occurring in a cell nucleus it may be expected that the effect is linearly related to dose. At higher doses, where multiple ionising events within a single cell are commonplace, damage arising from interactions between two or more events becomes more probable. Ultimately, at high doses cell killing will progressively reduce the risk of tumour induction. For single (acute) radiation exposures cell killing starts to become significant at doses of a few gray. The generalised dose-response is given in Equation 3 (Section 2.2.3.2). The difficulty in assessing risks of cancer following exposures to low-LET radiation at low doses and dose rates is illustrated in Fig. 2.9. This gives, schematically, data points and possible dose-response curves for cancer induction. Frequently, as in this example, information is only available at relatively high doses. An approach commonly used in risk assessment is to fit a linear dose-response relationship to the data (curve B) a procedure usually considered to give an upper limit to the risk at low doses. This will be the case unless significant cell killing has occurred. If this linear relationship is due to single tracks acting independently then the effect per unit dose would be expected to be independent of dose magnitude and dose rate. In practice, however, this is not generally observed and the linear quadratic relationship (curve A) frequently gives a better fit to the data at low to intermediate doses implying that at higher doses damage is the result of both single and multiple tracks. At still higher doses cell killing becomes significant with a consequent reduction in tumour yield.
ì í Low LET î
A - High absorbed doses and high dose rates B - Linear, no threshold C - Low dose rate D - Limiting slope for low dose rate
Induced incidence of cancer
B
High LET
A
C D
Absorbed dose D [Gy]
Landolt-Börnstein New Series VIII/4
Fig. 2.9. Dose-response relationship for radiationinduced cancer: possible inferences are illustrated in extrapolating data available at high doses and high dose rates to response at low doses and dose rates for lowLET radiation; [based on 90N2].
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With a progressive lowering of the dose and the dose rate, allowing more opportunity for repair of damage, curve C might be obtained. A point may ultimately be reached at which multiple track events make a negligible contribution to tumour incidence and damage is produced only as a result of single tracks acting alone giving a linear response (curve D) with the effect directly proportional to dose (slope α1, the risk coefficient). A similar response would be obtained by lowering the dose rate alone as even with high total doses the rate of build up of lesions would be slower and the opportunity for multiple track events would decrease. Hence in the limit, curve D, could be achieved either by reducing the dose to very low values so that effects are independent of dose rate or by reducing the dose rate to very low values. The approach used for assessing risks at low doses and low dose rates of low-LET radiation is described in Sections 2.4.5 and 2.4.6. For high-LET radiation it is assumed that there is no dose rate effect and the response is proportional to dose for doses below those at which there is cell killing. The data on the A-bomb survivors provide information on risks of cancer in a range of tissues, although to date no quantitative information is available for radiation-induced cancers of the liver, cells on bone surfaces, thyroid and skin. Information on radiation-induced cancer in these tissues is, however, available from other epidemiological studies summarised in Table 2.6. The principal studies used to quantify the effects of both external radiation and internally incorporated radionuclides are summarised below. Table 2.6. Human populations available for risk estimation Atomic bombs Medical diagnosis Medical therapy
Occupational exposure Radiation accidents
Japanese survivorsa Marshall islandersa,b Multiple fluoroscopies (breast)a Prenatal irradiationa Thorotrast injectionsc Pelvic radiotherapy (cervix)a Spinal radiotherapy (ankylosing spondylitis)a Neck and chest radiotherapy (thyroid)a,b Scalp radiotherapya Radium treatmentc Uranium minersc Radium ingestion (dial painters)c Radiation workersa,b,c Chernobyla,b
a) Exposure to external radiation b) Internal exposure to β/γ internal emitters c) Internal exposure to α emitters
2.4.3 Exposures to external radiation 2.4.3.1 The A-bomb survivors in Japan The mortality experience of the Hiroshima and Nagasaki A-bomb survivors has been the single most important source of information on the risk of radiation-induced cancer. This population has been the subject of a comprehensive follow-up since 1950. Information is available on the exposure of individuals to whole body radiation at a range of ages. Data on mortality from radiation-induced cancer that became available in the 1980s on the population of more than 90,000 people in the Life Span Study (LSS) necessitated a revision of previous risk estimates [87P3, 90S3]. There were a number of components to this change. The first was a revision of the dosimetry (termed DS86) to allow, amongst other factors, for the high humidity in the air over the cities which has substantially reduced the neutron dose at Hiroshima from the earlier 1965 (T65) estimates which were based on measurements in the dry atmosphere of the Nevada desert. Improved estimates were also made of the yield of the Hiroshima bomb (increased from 12.5 to 15 ktonnes), the shielding provided by buildings and of tissue and organ doses. The second was Landolt-Börnstein New Series VIII/4
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that the number of excess fatal cancers in the population increased due to the longer period of follow-up (to 1985) and an estimate of the cancers occurring in the period 1945-1950 was made. The third, and most significant change, was that relative risk, rather than additive risk models appeared to provide a better basis for assessing lifetime risk of most solid cancers (Section 2.4.2.1). UNSCEAR [88U5] in a report to the General Assembly provided the first information on radiation-induced cancer risks for a number of tissues in the Japanese population based on relative risk projection models. The total cancer risk in the population, at high dose and high dose rate, was then estimated to be 7-11 × 10−2 Sv−1 using age-averaged and age-specific constant relative risk models. This compared with the Committee's 1977 assessment of 2.5 × 10−2 Sv−1 [77U2] at high dose rate using the additive model. Because children and young persons are more sensitive to radiation than adults. The application of age specific risk coefficients therefore increases the predicted numbers of radiation-induced cancers in the whole population compared with that for a working population (Section 2.4.6). These risk estimates for whole body radiation exposure were based on an extrapolation into the future which is somewhat uncertain for solid cancers because two-thirds of the Japanese survivors were still alive and two-thirds of the cancer risk had still to be expressed. Up to 1985 about 80 excess leukaemias and 260 excess solid cancers had occurred in the LSS population for whom DS86 doses were available out of a total of about 6000 cancer deaths [87P3]. The risk of radiation-induced leukaemia is more certain than that for solid cancers, however, as few more excess cases are now expected. There remain uncertainties in extrapolating the cancer risks based on the Japanese population exposed to radiation at high dose rates to the low doses and dose rates relevant for radiological protection purposes (see Section 2.4.5). In a more recent report on the LSS, Pierce et al [96P1] reported on five more years of follow-up (1986-1990). Their analysis included an additional 10,500 survivors (86,572 in total). During 1950-1990 there have been 7827 cancer deaths, of which it is estimated there are 87 excess leukaemias and 334 solid cancers. The mortality curve for all solid cancers combined shows essentially a linear dose-response in the range 0-3 Sv, whereas for leukaemia the trend in dose is non-linear with an upward curvature. The radiation-induced leukaemia risk seems to have been almost completely expressed during the follow-up period, and the lifetime excess absolute risk of leukaemia associated with an acute dose of 1 Sv has been estimated as being about 1 %. However, in contrast to leukaemia, nearly a quarter of the radiation-induced solid cancers are estimated to have arisen in the most recent five-year period of the mortality follow-up, i.e. 1986-90 [96P1]. Since most of the A-bomb survivors exposed at young ages are still alive, the future pattern of cancer risks in this group will be important in determining lifetime risks. A significant increase in the risk of solid cancers is now seen at doses down to about 50 mSv [00U8]. 2.4.3.2 Thyroid cancer A number of epidemiological studies provide information on cancer risks in individual tissues. Groups of children and young persons who received thyroid irradiation, and who can be used to derive risk coefficients for thyroid cancer, include children who received X-ray treatment for thymic enlargement, patients treated in US hospitals for thyrotoxicosis and other benign lesions of the neck and patients who received X-ray treatment for thyroid disease [85N1, 85S4, 00U8]. In the majority of cases, particularly in the young, thyroid cancer is not fatal. The mortality from radiation-induced thyroid cancer is expected to be about 10 % of the incidence. There is also evidence that the risk in adults is about half that in children and that the risk in females is about twice that in males. For a population uniformly exposed to external radiation the risk of fatal thyroid cancer is estimated to be 8.0 × 10−4 Sv−1 assuming a 5 year latent period [91I2]. In human populations given iodine-131 for non-therapeutic reasons, and who received doses well below 2 Gy, no significant excess of thyroid cancers has been observed. This suggests a risk coefficient 3 to 4 times less than that obtained following external radiation at high dose rates [85N1]. Data on thyroid cancer incidence in children in areas of the former Soviet Union that were contaminated with fall-out from Chernobyl indicate an increased risk of thyroid cancer in some areas. To date the data are insufficient to provide quantitative risk estimates [00U8]. Thyroid cancer risks from exposures to radioiodine are considered further in Section 2.4.4. Landolt-Börnstein New Series VIII/4
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2.4.3.3 Skin cancer An ICRP Task Group [91I3] reviewed data on the risks of skin cancer. Most of the data came from groups given partial body irradiation in the course of medical treatment, although some data were also available from occupationally exposed groups, in particular radiologists and radiation technicians and uranium mining populations. Little information is available from the A-bomb survivors. On the basis of a relative risk model, the Task Group calculated a risk of fatal skin cancer for exposure of a general population of 2 × 10−4 Sv−1 at low doses, on the assumption that 0.2 % of cases would be fatal. They stressed the uncertainty in assessing the temporal pattern of radiation-induced skin cancers. 2.4.3.4 Breast cancer Data are available on radiation-induced breast cancer from follow-up studies on the A-bomb survivors as well as from studies of patients in North America given fluoroscopy examinations for tuberculosis or treated for acute postpartum mastitis. Risks calculated from either population are little different, based on additive projection models. ICRP has based its risk estimate of 2 × 10−3 Sv−1, for a mixed population of men and women, on the data on the A-bomb survivors. The risk of breast cancer also varies considerably with age at exposure. Thus, for exposure in the first decade of life, the risk is about 4 times that at ages 40-50 years [93M4].
2.4.4. Exposure to internally incorporated radionuclides Human data on cancer induction from internally incorporated radionuclides are available for only a few radionuclides and have been reviewed by UNSCEAR [94U7, 00U8]. Quantitative data for risk estimation are available only for alpha particle emitting radionuclides. Limited data are available on humans exposed to β/γ emitters. A number of epidemiological studies have followed groups exposed to 131I. These studies cover a wide range of doses, varying from very high doses delivered in the treatment of hypothyroidism to the low doses received by patients exposed to diagnostic procedures or exposed to radiation from fallout in the Marshall Islands. The information available provides little evidence that exposure to 131I is associated with a risk of thyroid cancer, although in some cases the follow-up is relatively short. This lack of effect, compared with the effect of external radiation, may be due to an effect of dose rate or to differences in the distribution of dose within the gland. There may also be differences due to ages at exposure. As in the case of external radiation the groups were predominantly young persons. The extent to which exposures to 131I has contributed to the increased risk of thyroid cancer following the Chernobyl accident is still uncertain. Some very sparse data on tumour induction are available on a few individuals given 32P, 35S and 59Fe for medical reasons and there is some information on persons in the Southern Urals exposed to 90Sr who used water from the Techa River for drinking and irrigation [94K1]. A number of studies have also considered the effects of radionuclides in weapons fallout or in discharges to the environment from other nuclear facilities. These data do not at present provide a basis for assessing risks from intakes of β/γ emitting radionuclides. The available information on α-particle emitters covers groups exposed to radium isotopes (224Ra, 226 Ra, 228Ra) where bone tumours are the predominant late effect, and Thorotrast (colloidal 232ThO2) which principally results in irradiation of the liver, spleen and bone marrow, with tumours arising mainly in the liver and bone marrow (leukaemia). Information is also available in man on lung cancer following occupational exposure to radon and its decay products. A number of epidemiological studies of domestic exposure to radon have been published and others are presently under way, to date the data are generally consistent with risks obtained from worker studies although exposures are lower and have a reduced sensitivity for obtaining quantitative risk estimates. Twenty-six men who worked with plutonium in North America on the Manhattan project during the Second World War have also been studied (estimated body Landolt-Börnstein New Series VIII/4
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contents 52-3180 Bq). Seven individuals had died by 1991. The causes of death were lung cancer (2 cases), myocardial infarction, arteriosclerotic heart disease, accidental injury, respiratory failure due to pneumonia/congestive heart failure and osteosarcoma of the sacrum. Three men also had a history of skin cancer [91V3]. There is a high probability that the bone cancer was caused by exposure to plutonium as the spontaneous risk is about 1 in 2000. ICRP [91I2] has recommended the use of radiation weighting factors wR for calculating the equivalent dose to tissues and thus interpolating between the effects of high and low-LET radiation. The wR for αparticle irradiation is taken to be 20. 2.4.4.1 Radium-226/228 luminisers An increased incidence of bone cancer and of head sinus carcinoma has been observed in persons in the USA exposed to long-lived radium, particularly in painters of luminous dials, but also radium chemists or persons treated with radium salts for a possible therapeutic effect [86R2, 94R1]. These persons became internally contaminated with pure 226Ra (t1/2 = 1,600 years) in some cases, and in other cases with various mixtures of 226Ra and 228Ra (t1/2 = 5.8 years). Bone cancers and head sinus carcinomas have arisen in these populations. The majority of these cancers had appeared by 1969, although three bone tumours have appeared since then and more recently head cancers have appeared at a greater rate than bone cancers. The radium isotopes deposit principally in the skeleton and the bone sarcomas appear to have been induced by α particles from either the 226Ra or 228Ra decay series. The head sinus carcinomas are thought to be caused mainly by the accumulation of decay products of radon (222Rn) gas in the frontal sinuses and mastoid air cells. This radon is produced by the decay of 226Ra in the bone. Except for the bone sarcomas and head sinus carcinomas no definite excess in other types of malignancy, including leukaemia, is presently ascribed to the internal deposition of long-lived radium. The follow-up study on this population was essentially discontinued in the USA in the mid 1990s. 2.4.4.2 Radium-224 patients The effects of intakes of radium has also been studied in German patients injected with 224Ra shortly after World War II. The study group consists of a population of 682 adults and 218 juveniles (age at first injection varied between 1 and 20 years) who received weekly or twice weekly intravenous injections of 224 Ra, mainly for the treatment of bone tuberculosis or ankylosing spondylitis [86M1, 94S5]. The last bone tumour occurred in 1988, 41 years after the injection of 224Ra into a three-year-old boy and is the only bone sarcoma reported in this series since 1974. Very few new tumours are now expected and follow-up of the population is now limited. Based on the information on the incidence of bone cancers following intakes of 224Ra and average bone dose from its deposition in the skeleton, ICRP [91I2] has adopted a total risk estimate for fatal cancer of 5 × 10−4 Sv−1 (assuming a radiation weighting factor wR for α-particle irradiation of 20). 2.4.4.3 Miners exposed to radon An increased mortality from lung disease has been observed in under-ground miners working in Czechoslovakia, Canada, United States of America and Sweden exposed to radon (222Rn) and its decay products [88B3, 98B5]. The increase in mortality from lung cancer has been correlated with air concentrations of radon in different mines and the duration of exposure. Bronchial stem cells and secretory cells in the airways are considered to be the main target cell for the induction of lung cancer resulting from radon exposure. There are many difficulties in calculating the radiation dose to these cells as a result of exposure to radon
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decay products (expressed in working level months1). The radiation dose over the working life must be taken into account and the dust loading of the atmosphere known as it determines the extent of the uptake of radon decay products onto the respirable particles. In addition to any possible synergistic effects between smoking and radon exposure, the presence of dust, diesel fumes and other possible carcinogens in the mine atmosphere causes some uncertainty as to whether an excess of cancer can be attributed to radiation alone. The BEIR VI Committee [98B5] recommended two models for estimating radon risks based on its analysis of the data on radon-exposed miners, without expressing a preference for either. One of the BEIR VI models takes account of factors such as total exposure, age and average radon concentration. Risks predicted under the latter model are about 50 % greater than those based on the former model. The BEIR VI Committee also considered both multiplicative and submultiplicative versions of these models. The risk predicted for smokers under the submultiplicative form of each model is only slightly smaller than that based on the multiplicative version. In contrast, the risk for non-smokers under the submultiplicative assumption is about twice that under the multiplicative version of the corresponding model. Based on the various combinations of the BEIR VI models, the lifetime risk of lung cancer for smokers in the UK would lie in the range 10 %-15 %, while that for non-smokers would be in the range 1 %-3 %. For a general population of smokers and non-smokers, the range in lifetime risks would be about 3 %-5 %. The BEIR VI model can also be used to calculate total risks of lung cancer in a population in absolute terms. Thus in the UK population lung cancers attributable to the mean domestic radon concentration of 20 Bq m−3 would be in the range of 2000-3300 per year, based on the above models. Taking into account the proportion of non-smokers in the population, it can be estimated that about 500-1300 of radon-associated deaths would arise among non-smokers. For a working population, ICRP [91I2] have adopted a risk factor for lung cancer of 0.68 × 10−2 Sv−1 based on data from the the A-bomb survivors. 2.4.4.4 Thorotrast patients Thorotrast is colloidal thorium dioxide. In the late 1920s it began to be injected into the arteries of patients for use in diagnostic radiology as an X-ray contrast material. The average dose of about 25 ml of Thorotrast contained 5 g of thorium with an activity (from α-particles) of about 20 kBq 232Th with additional radioactivity from its decay products. The colloidal Thorotrast was cleared from the bloodstream by uptake into phagocytic cells depositing about 60 % in liver, 30 % in spleen and 10 % in red marrow. Extensive epidemiological studies in Portugal, Sweden, Denmark, the United States, the Federal Republic of Germany and Japan have shown that retention of thorium dioxide particles in the liver and in the bone marrow resulted in an increased risk of liver tumours and leukaemias as well as liver cirrhosis and other cardiovascular diseases [84V1, 94V2]. On the basis of an injected dose of 25 ml the dose to the liver is estimated to be 0.25 Gy y−1 (high-LET). Present estimates, based on a latent period of 20 years, suggest a lifetime risk of liver cancer following exposure to Thorotrast of about 0.15 × 10−2 Sv−1 (assuming a wR for α-particle irradiation of 20), about half this risk is expected to be expressed by 40 years after exposure [88B3, 91I2].
1
1 WL is any combination of the short-lived decay products of radon per litre of air which will result in the ultimate emission of 1.3 × 105 MeV of α particle energy. A WLM results from exposure to a concentration of decay products in air of 1 WL for an average working month of 170 hours.
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2.4.5 Dose and dose rate effectiveness factors (DDREFs) Risk coefficients for radiation-induced cancer are based mainly on population groups exposed at high doses and high dose rates as described above. Studies at the molecular, cellular, tissue and whole animal level have demonstrated that radiation damage increases with dose and that, at least for low-LET radiation, at high dose rates it is often greater per unit of exposure than at low dose rates. Thus, although the assumption normally made for radiation protection purposes is that the dose-response curve for cancer induction is linear, with the risk proportional to dose, in practice a dose and dose rate effectiveness factor (DDREF) has commonly been used to allow for a reduced effectiveness of radiation in inducing cancer in man at low doses and low dose rates. The choice of a suitable DDREF has caused considerable debate with relevant data being available from cellular and animal studies, as well as human epidemiology. ICRP in its 1990 recommendations based estimates of DDREF principally on an analysis by Pierce and Vaeth [89P2] of the data from the Japanese survivors. This analysis shows that the data do not allow for a reduction factor of much more than about 2. Other epidemiological data showed little evidence of dose rate effects although studies on thyroid cancer incidence [85S4] and breast cancer mortality [89M3] indicate possible reduction factors of up to 3 or 4. As a consequence ICRP adopted a DDREF of 2, recognising that ‘the choice is somewhat arbitrary and may be conservative’. In practice, the DDREF would be expected to vary with tissue and with exposure conditions although a single value had to be assigned for protection purposes. A better understanding of the mechanisms involved will be essential for improving understanding of the effects of both dose and dose rates on radiation-induced tumour induction in man. A summary of values of DDREF recommended by national and international bodies is given in Table 2.7. No DDREF is recommended for high-LET radiation (i.e. DDREF = 1). Table 2.7. Summary of dose and dose rate effectiveness factors for radiation-induced cancer Reference DDREF Source ICRP 1977 77I1 2 NCRP 1980 90N2 2-10 UNSCEAR 1986 86U4 up to 5 UNSCEAR 1988 88U5 2-10 BEIR 1990 90B4 2 ICRP 1991 91I2 2 UNSCEAR 1993 93U6 <3 UNSCEAR 2000a 00U8 <3 a)
3 for hereditary disease
2.4.6 Risk coefficients for protection In the last few years a number of reports have been published which have calculated risks of radiationinduced cancer for different populations. They have been based predominantly on information derived from the A-bomb survivors but supplemented by data from other epidemiological studies as summarised above. Most risks have been calculated for the general population, although a number of reports have also given risks for workers. These tend to be lower (by about 20-40 %) because of the greater risk to children and young persons calculated using the relative risk projection model for most solid cancers. The assumption made for protection purposes is that the incidence of radiation-induced cancer increases with the dose, with no threshold. Thus they are stochastic in nature. Tables 2.8 and 2.9 summarise the information on risks of radiation-induced cancer at high doses and high dose rates published in recent years by UNSCEAR [88U5, 00U8], BEIR [90B4], NRPB [93M4] and ICRP [91I3], using mainly relative risk projection models for most solid cancers. In the majority of studies lifetime risks of cancer have been calculated, although NRPB also gave risks to 40 years after exposure (the then period of follow-up of the A-bomb survivors). UNSCEAR [88U5] calculated risks based on both an ageLandolt-Börnstein New Series VIII/4
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averaged and an age-specific constant relative risk models. BEIR V [90B4] calculated risks to a US population and gave values for a number of tissues using time-varying relative risk models for some cancers (leukaemia, respiratory tract, breast cancer in females). It is noteworthy that BEIR V, unlike UNSCEAR, calculated excess cancer deaths, not early deaths. The former risk is about 20-25 % less than the latter reflecting the baseline cancer rate in the population. ICRP [91I2] calculated risks for a 'world' population based on an average value for five populations (Japan, UK, USA, Puerto Rico, China) and on transferring both absolute and relative risks across populations. Table 2.8. Estimated lifetime fatal cancer risks in populations (all ages, both sexes) associated with exposure to low-LET radiation at high doses and high dose rates, based on a multiplicative projection model Source UNSCEAR 1977 UNSCEAR 1988 BEIR V 1990 ICRP 1991 Muirhead 1993 a) b) c) d) e)
Reference
Population
77U2 88U5 90B4 91I2 93M4
Japan USA Five nations UK
Fatal cancer risk [10−2 Sv−1]
2.5a 7-11b 7.9c 10.0d 4.9 - 11.80
additive model range based on age-averaged and age-specific constant relative risks see text (Section 2.4.6) average value based on US, UK, Japan, Puerto Rico and Chinese populations. Risk for workers 8.0 × 10−2 Sv−1 risk calculated to 40 years after exposure and lifetime assuming age-specific relative risks. Risk for workers 5.9-10.1 × 10−2 Sv−1.
Note: These values are for acute doses only and do not include an adjustment for dose rate.
Table 2.9. Lifetime fatal cancer risks given by UNSCEAR 2000 [00U8]. Fatal cancer risk
Leukaemia Solid Cancers Males Females
1 Sv
0.1 Sv
1%
0.05 %
9% 13 %
0.9 % 1.3 %
NOTE: These values are for acute doses only and do not include an adjustment for dose rate
Overall the lifetime risks calculated in recent years are not too different for the various studies, the lowest value being for UNSCEAR [88U5] using age-averaged risk coefficients. ICRP [91I3] adopted a rounded value of 10 × 10−2 Sv−1 for the risk coefficient for fatal cancer at high doses and high dose rate following exposure of a mixed population of all ages. Applying a DDREF of 2 gives a risk of 5 × 10−2 Sv−1 for radiation protection purposes. Risk coefficients for individual tissues are given in Table 2.10. For workers the risk coefficient adopted for radiation protection purposes is 4 × 10−2 Sv−1. These risk coefficients have been used by ICRP in developing the dose limits given in the 1990 recommendations [91I3] and provide the basis for the International Basic Safety Standards [96I5] and the European Basic Safety Standards [96E2].
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Table 2.10. Risk coefficients for fatal cancer adopted by ICRP Organ or tissue Bladder Red bone marrow Bone surface Breast Colon Liver Lung Oesophagus Ovary Skin Stomach Thyroid Remainder Gonads (hereditary disease) Total
Fatal cancer risk coefficient [10-2 Sv-1] ICRP 1991b ICRP 1977a Population Workers
0.05 0.50 -
0.30 0.50 0.05 0.20 0.85 0.15 0.85 0.30 0.10 0.02 1.10 0.08 0.50 -
0.24 0.40 0.04 0.16 0.68 0.12 0.68 0.24 0.08 0.02 0.88 0.06 0.40 -
1.25
5.0
4.0
0.20 0.05 0.25 0.20
a) 77I1 b) 91I2
2.4.7 Low dose studies The majority of studies on which risk estimates for radiation-induced cancer are based are for populations exposed at high doses and high dose rates. Studies of low dose rate exposure generally involve low doses and because of the likely low excess risks are hampered by lack of statistical power and possibly also by confounding factors. However low dose rate studies can provide a check on the risks derived by extrapolation from high dose rate studies. The main studies of interest are on workers who are occupationally exposed although some data are also available on risks in children following exposures in utero and on persons from areas of high natural background. 2.4.7.1 Occupational exposures Several studies have been conducted of nuclear industry workers. In the USA, Gilbert et al, [89G1] performed a joint analysis of data for about 36,000 workers at the Hanford site, Oak Ridge National Laboratory and Rocky Flats weapons plant. Neither for the grouping of all cancers nor for leukaemia was there an indication of an increasing trend in risk with dose. In 1976 NRPB set up the National Registry for Radiation Workers (NRRW). The NRRW was designed to investigate the effects of occupational exposures to ionising radiation by direct epidemiological observations. The first analysis of the NRRW was published in l992 [92K2]. The main findings were: • • •
a strong “healthy worker effect”, a statistically significant trend in leukaemia risk with dose, and weaker evidence of a trend with dose for solid tumours.
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A “healthy worker effect” means that death rates in the group of workers studied are lower than in a group of the general population of the same age and sex. This is a common finding in epidemiological studies of working populations. The observation of a healthy worker effect in the NRRW cohort was reassuring, but not unexpected. The more important findings were a trend with dose for leukaemia and a (non-significant) trend for solid tumours. The risk factors in the first NRRW analysis were compatible with those recommended by the ICRP but the confidence intervals were wide. To obtain more precise information on the risks of radiation work NRPB carried out a second analysis of the NRRW with a larger cohort and a longer period of follow-up [99M6]. A comparison of the main features of the two studies is shown in the box. Comparison of NRRW cohorts for the first and second analyses First
Second
No. of Workers
95,217
124,743
Collective dose [man Sv]
3,198
3,810
Mean dose [mSv]
33.6
30.5
Person-years
1.2 million
2 million
Total deaths
6,660
12,972
The second NRRW analysis provided further information on mortality among UK radiation workers. As in the first analysis, there was a strong healthy worker effect, with mortality from all causes and all malignancies less than that expected from national rates. The 90 % confidence intervals for the trend in cancer risk with external dose are tighter than before, and they now exclude values more than four times those seen among the Japanese A-bomb survivors, although they are also generally consistent with no raised risk. For leukaemia excluding chronic lymphatic leukaemia (CLL) there is evidence, of borderline statistical significance, of an increasing risk with dose and, as with solid cancers, the data are consistent with the A-bomb findings. Further analyses should provide more information on risks in relation to occupational radiation exposure. A combined analysis of mortality among 95,673 workers (85.4 % men) in the US, the UK and Canada has been published [95C1]. The combination of the data from the various studies increases the overall power to study associations between radiation and specific cancers. The combined analysis covered a total of 2,124,526 person-years (PY) at risk and 15,825 deaths, 3,976 of which were due to cancer. As with the NRRW, mortality from leukaemias, excluding CLL was significantly associated with cumulative external radiation exposure. There was no evidence of an association between radiation dose and mortality from all cancers. It was concluded that the results of the study did not suggest that current radiation risk estimates for cancer at low levels of exposure are appreciably in error. 2.4.7.2 Background radiation Studies of exposure to natural radiation (other than radon) have generally involved looking for any geographical correlation with cancer rates. Such studies are difficult to interpret, however, owing to the effect of confounding factors such as socio-demographic variables and other factors that vary geographically.
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2.5 Hereditary disease Radiation damage to the male and female germ cells, resulting in an increase in hereditary disease, is the other main late effect resulting from exposure to ionising radiation. Only very limited data on genetic damage are available from human populations and estimates of risk have to be derived mainly from animal studies.
2.5.1 Categories of genetic damage Inheritance is the process by which the genetic information carried by the DNA in the cell nucleus is passed from one generation to the next. This is essentially an orderly process but mutations do arise spontaneously giving a positive background level of genetic damage. Hereditary effects may occur by changes arising in the base sequence in the DNA of a single gene leading to gene mutation, or by rearrangement of collections of genes within and between chromosomes causing chromosomal aberrations. Both may be produced by radiation. There are three main types of gene mutation namely dominant, recessive and X-linked. Every individual receives a set of genes from each parent. If dominant, a gene mutation in one set of genes but not in the other can express itself in spite of its counterpart from the other parent being normal. A recessive gene mutation cannot be expressed unless the genetic material from both parents carries the same mutation. Females have two X chromosomes and males one X and one Y, the Y chromosome being virtually inert apart from factors for maleness. An X-linked gene mutation can readily express itself in the male whereas in the female X-linked mutations will not express themselves unless both X chromosomes carry the same mutation. The normal chromosome number in man is 46. Chromosomal mutations are due to alterations in chromosome numbers or structure. If the number of chromosomes is increased or decreased in the fertilised egg this produces such profound effects that, except in a few instances, death is likely to occur soon after conception; if a child survives it is likely to have severe physical and/or mental defects. There is at present no good evidence for radiation-induction of diseases of chromosomal origin [82U3, 88U5]. The genetics of some inherited diseases are more complicated because some relatively common chronic diseases have a genetic element but additional factors such as environment play a part in their expression. Changes in the mutation rate will also alter the incidence of these 'multifactorial' diseases. Examples of various categories of hereditary diseases have been reviewed by UNSCEAR [77U2]; examples are given in Table 2.11. Most live-born children with inherited chromosomal mutations exhibit mental and/or physical abnormalities. There is little or no chance of sufferers who reach adulthood reproducing and so passing these defects on to their children. These conditions are therefore maintained in the population by new mutations arising either spontaneously or induced by an environmental insult such as radiation. Dominant mutations show up in the first generation after exposure as do X-linked mutations in males and may occur in subsequent generations if they do not prevent childbearing. Recessive mutations, however, tend to occur in later generations. When assessing the risks of radiation it is therefore necessary to allow for hereditary effects which may not appear for several generations.
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Table 2.11. Examples of hereditary diseases [77U2]. Dominant disorders Congenital cataract Cystic kidney disease Huntington’s chorea (progressive mental retardation) X-linked diseases Haemophilia Albinism Colour blindness Heart valve defects Autosomal recessive diseases Cretinism Disorders of amino acid metabolism Aplastic anaemia Muscular dystrophy Multifactorial diseases Ankylosing spondylitis Varicose veins Cleft palate Diabetes mellitus Schizophrenia Asthma Chromosome anomalies Down’s syndrome
2.5.2 Risk coefficients for hereditary disease So far no hereditary effects at levels that are statistically significant have been observed in human populations exposed to radiation [88U5]. Neel et al [89N4] have reviewed all the genetic studies in Hiroshima and Nagasaki on the children born to irradiated survivors. The 'end points' considered were congenital defect, survival of liveborn infants, sex-chromosome aneuploidy and balanced chromosomal exchanges, cancer with onset below age 20, mutations altering protein charge or activity, sex ratio, and physical growth and development. The average conjoint parental gonad exposures for the parents was about 0.5 Gy (based on DS86 dosimetry), the exact figure depending on the radiation histories of the parents whose children formed the basis for a specific end point. No statistically significant effects were observed. Taken together, the data suggest a lower limit for the doubling dose for genetic damage following acute irradiation of approximately 1.4-1.8 Sv. This compares with a value of 0.3 Sv in the mouse for acute exposure and 1 Sv for chronic exposure. The assumption made in calculating risks for radiation protection purposes is that the incidence of radiation-induced hereditary disease increases with the dose, with no threshold. Thus they are stochastic in nature. In the absence of direct quantitative human data, animal studies have been used by UNSCEAR [88U5] to assess the risk of radiation-induced hereditary disease in human populations. Tables 2.12 and 2.13 gives some estimates of the incidence of hereditary diseases in a population recommended by UNSCEAR and ICRP in recent years. They are based on a doubling dose for hereditary disease of 1 Gy derived from animal studies. In the most recent report by UNSCEAR [01U9] the risk of radiation-induced hereditary disease has been appreciably reduced compared with its previous advice.
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Table 2.12. Incidence of genetic disease at equilibrium from parental exposure. Incidence (low-LET)a [10−2 Gy−1] Disease classification Chromosomal anomalies Dominant and X-linked Recessive Multifactorial Total
UNSCEAR, 1977 [77U2]
0.4 1.0 0.45 1.85
ICRP, 1977 [77I1]
UNSCEAR, 1982 [82U3]
UNSCEAR, 1988 [88U5]
ICRP, 1991 [91I2]
2.0
0.04 1.0 0.45 1.49
0.04 1.0 0.15 1.19
0.04 1.0 0.15 3.5b 2.4
a) Assuming doubling dose of 1 Gy. b) Severity less than other genetic diseases, weighted by factor of 1/3.
Table 2.13. Incidence of genetic disease from one-generation exposure to low-LET, low dose rate or chronic radiation [01U9]. Disease class Chromosomal anomalies Dominant and X-linked Recessive Multifactional Chronic multifactional Congenital anomalies Total a)
Baseline frequency per 106 live births 4,000 16,500 7,500
Incidence (low-LET)a [10−2 Gy−1] 1st generation 2nd generation 0.075-0.15 0.05-0.10 -
650,000 50,000 738,000
~0.025-0.12 ~0.2 ~0.3-0.47
~0.025-0.12 ~0.04-0.10 0.11-0.32
Assuming doubling dose of 1 Gy
In its most recent recommendations ICRP [91I2] gives a risk of hereditary disorders of 2.4 × 10−2 Sv−1 expressed over all generations following exposure of either parent (Table 2.12). This risk factor includes a risk factor for multifactorial diseases. At present the information on such diseases is very limited and only a tentative value is available. ICRP [91I2] have assessed the risk as 3.5 × 10−2 Sv−1 following exposure of either parent. The severity of multifactorial diseases are considered to be not as great as other hereditary diseases so they are weighted by a factor of three, giving a risk factor of 1.2 × 10−2 Sv−1. The genetically significant exposure in a population will be less than this because a proportion of the population are older than child bearing age. If the mean age of child bearing is 30 years and average life expectancy is 75 years then the probability of genetic harm resulting from exposure of the entire population is 30/75 (= 0.4) × 2.4 × 10−2 Sv−1 ≈ 10−2 Sv−1 [91I2] (Table 2.14). For a working population the reproductive fraction is less than the entire population. For a working population, the reproductive fraction is (30-18)/(65-18) = 0.25. The risk factor for workers is thus about 0.6 of that for an entire population (0.25/0.4) giving a risk factor for workers of 0.6 × 10−2 Sv−1 (60 % of 1 × 10−2 Sv−1) (Table 2.14).
2.6 Irradiation in utero For the developing embryo and fetus there is evidence that deterministic effects, severe mental retardation and cancer induction may occur following irradiation in utero. The risk of hereditary disease may be taken to be the same as after birth.
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2.6.1 Deterministic effects Evidence of the deterministic effects of radiation on the embryo and fetus is derived almost entirely from animal experiments. It is necessary to extrapolate the results of these studies to predict the consequences of radiation exposure in man. The effects of radiation on the embryo depend on the time of exposure relative to its development. When the number of cells in the embryo is small (i.e., in the first few days after fertilisation) and they are not yet specialised, damage is frequently seen in animals as failure of the conceptus to implant or loss of embryos, which would be seen in humans as miscarriage. However, recent evidence from in vitro human embryo research has shown that the survival of even one cell in the early embryo before implantation can allow normal development of cells to occur, since all the necessary genetic components are present in each cell of the embryo at this early stage of development. The consequences of any of these cells carrying a point mutation are unquantified, but the possibility of stochastic effects occurring cannot be dismissed. Malformations have been observed in rodent embryos at a stage when organs such as the brain, skeleton, eyes and heart are developing. Congenital abnormalities are commonly found in the offspring of rodents but any attempts to project the results to predict effects in man are fraught with difficulties. With this cautionary note and bearing in mind, proposed human threshold doses for radiological protection purposes for low-LET radiation: 0.05 Gy for reabsorption of pre-implantation embryos; 0.05 Gy for minor skeletal abnormalities; 0.2 Gy for functional disorders of the central nervous system; and between 0.2 and 0.5 Gy for serious skeletal abnormalities and growth retardation, such information provides a basis for guidelines to ensure that pregnant women are adequately protected [03I1]. Protraction of the dose will reduce any effect.
2.6.2 Brain function The human brain is probably the most complex organ in the body and its proper development and function depend upon an elaborate sequence of events which must be coordinated temporally and spatially. Any disturbance of this sequence could lead to abnormality since the normal function of the nervous system depends upon the proper location of the neuronal cells. A study of about 1600 children exposed in utero at Hiroshima and Nagasaki to various radiation doses and at various developmental stages has shown about 30 cases of clinically severe mental retardation with a greater incidence than expected in the higher dose groups. Excess mental retardation was not observed following exposure up to 8 weeks from conception, was at a maximum between 8 and 15 weeks and then was somewhat lower between 16 and 25 weeks. No effect was observed following exposures later than 25 weeks [84O2; 88O3]. The period of maximum sensitivity (8-15 weeks) corresponds with the timing of both of the major waves of neuronal proliferation and migration within the cerebral cortex. Although the number of cases is small, the data indicate an excess probability of 40% at 1 Sv received during the 8-15 weeks after conception. The current results of IQ tests amongst those children exposed in utero indicates a general downward shift in the distribution of IQ with increasing dose. A coefficient of about 30 IQ points Sv-1 relates to in utero exposure between 8-15 weeks after conception. A smaller shift is identified in the 16-25 week period [88S1]. This downward shift in IQ of 30 points Sv−1 shown schematically in Figure 2.10 is consistent with the observation of an incidence of 0.4 for a dose of 1 Sv. At doses of the order of 0.05 Sv, no effect would be detectable in the general distribution of IQ, but at somewhat larger doses the effect might be sufficient to show an increase in the numbers classified as seriously mentally retarded. The net result is that the end point of serious mental retardation would appear to demonstrate a threshold, which is reasonably consistent. The ICRP now believes that the phenomenon is deterministic with a threshold related to the minimum shift in IQ that can be measured. It is not therefore included in the definition of radiation detriment used for protection purposes.
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∆x
∆f −x
m
Retarded fraction f
Ψ( X )
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Fig. 2.10. The shift to the left from ψ(x) by ∆x (30 IQ points) increases the background retarded fraction f by ∆f. –xm denotes the number of standard deviations below IQ 100 to classify an individual as mentally retarded [91I2].
x=0 (IQ 100)
2.6.3 Risk coefficients for cancer Information on the risk of cancer following irradiation in utero has been reviewed by UNSCEAR [72U1, 77U2] and by the BEIR-III Committee [80B2]. Current risk estimates for radiation-induced childhood cancer are based mainly on data collected in the Oxford Survey of Childhood Cancers (OSCC) concerning obstetric radiography [75B6]. This study contains information on over 150,000 childhood cancer deaths in Great Britain during 1953-81 and the same number of matched control children [87K3]. The OSCC, in common with other, smaller, case-control studies, indicates a relative risk of about 1.4 (40 % increase in risk) for childhood cancer associated with prenatal irradiation [89B7]. Concerns about possible bias and confounding in these case-control studies have been raised – for example, by Boice and Miller [99B8] – in view of issues such as the lack of evidence for a raised risk from cohort studies, and the similarity of the relative risks for leukaemia and other cancers in the OSCC. In their review of these issues, Doll and Wakeford [97D2] concluded that there is strong evidence against bias and confounding as alternative explanations for the raised risks seen in the OSCC and other case-control studies. The doses received by the fetus are uncertain; based on estimated average doses of about 10-20 mGy. Based on data from the OSCC and information from UNSCEAR [72U1] on doses received in utero from obstetric radiography the number of excess cancer cases (to 15 years of age) following irradiation in utero is calculated to be about 6 × 10−2 Gy-1 [93M4]. Since slightly less than 50 % of childhood cancers consist of leukaemia and other lymphatic/haematopoietic cancers [81O1] and the relative risks are similar for these and other cancers, a risk of 2.5 × 10−2 Gy−1 is calculated for leukaemia and 3.5 × 10−2 Gy−1 for solid cancers. As approximately half of all childhood cancers are fatal [81O1], the number of excess cancer deaths is calculated to be 3 × 10−2 Gy−1 (low-LET), comprising 1.25 × 10−2 Gy−1 for leukaemias and 1.75 × 10−2 Gy−1 for solid cancers. These risks are derived principally from follow-up studies on children irradiated in utero with radiation doses up to a maximum of 10-20 mGy (low-LET). They are therefore applicable for estimating risks at low doses and dose rates. There is also likely to be an additional risk of cancer that will appear late in life but the information is very limited. In addition, follow-up of persons exposed to A-bomb radiation in utero in Hiroshima and Nagasaki indicates that the raised cancer risk continues into adulthood, although quantification of this risk is difficult [88Y1, 97D1].
2.6.4 Hereditary disease Hereditary disease is considered in Section 2.5. Genetic studies in the offspring of atomic bomb survivors have not shown any significant radiation-related increases in any measure of genetic damage employed. In experiments in mice the sensitivity of fetal gonads was comparable to that of adult gonads or a little lower [74S2]. It is therefore assumed that the risks of hereditary disease from in utero irradiation are the same as after birth (2.4 × 10−2 Sv−1) following exposure of either male or female germ cells. It may be the risk will be lower in early embryogenesis and fetogenesis prior to the establishment of germinal tissues.
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2.7 Summary of risk factors for cancer and hereditary disease The ICRP, in its most recent 1990 recommendations [91I2], considers four components of the detriment (health effects) due to irradiation of the tissues and organs of the body at low doses when assessing the overall effects of radiation. These include the probability of fatal cancer; the probability of non-fatal cancer and the probability of severe hereditary disease, both weighted for severity relative to fatal cancer; and the time scale of appearance of these detrimental effects. The risk factors developed by ICRP for protection purposes are summarised in Table 2.14. The overall weighted severity values assigned to the non-fatal cancers and severe hereditary diseases (including multifactorial diseases) each amount to about one-fifth of the detriment associated with fatal cancer. In summary the aggregated detriment amounts to 7.3 × 10−2 Sv−1 for a nominal population. It is somewhat less (5.6 × 10−2 Sv−1) for a population aged 18-64 years who are occupationally exposed, when account is taken of the omission of younger persons who are more radio-sensitive and the shorter mean potential period of reproduction. The temporal pattern of fatal cancer risk is such that the period of maximum risk occurs in the seventh and eighth decades of life if the multiplicative projection model is used to calculate the lifetime expression of the cancers in persons exposed continuously to small annual doses at or below the dose limits. Table 2.14. Risk factors for protection [10−2 Sv−1] [77I1, 91I2] ICRP 1991 ICRP 1977 Public Workers Fatal cancer 1.25 5.0 4.0 Hereditary defects 0.4a 1.0b 0.6b Total 1.65 6.0 4.6 Total (weighted)c 7.3 5.6 a) Two generations b) All generations c) To allow for non-fatal cancers and years of life lost for cancers and hereditary disease.
2.8 Conclusions Deterministic effects in tissues and organs are the result of the loss of substantial numbers of stem cells, thereby cutting off the supply of functional cells. The consequence can be a temporary or permanent loss of tissue function which may be life threatening. A characteristic of the dose-response relationship for deterministic effects is that they are avoidable below a dose threshold. This is a reflection of sufficient numbers of stem cells maintaining functional cell populations. Knowledge of dose thresholds has been derived from the tolerance doses observed in radiotherapy. The tolerance doses vary with the tissue - the gonads, the bone marrow, the gastrointestinal and the lens of the eye being the most sensitive. It is the opinion of the ICRP that deterministic effects can be avoided if the presently recommended effective dose limits (based upon limiting stochastic effects) and the annual equivalent doses for the lens of the eye and the skin are not exceeded. There are a number of important questions that remain to be answered in the assessment of the risks of radiation-induced cancer in human populations. Very limited information is available at the low doses and low dose rates that are important for radiation protection purposes and the risks have to be assessed from populations exposed at high doses and dose rates by applying an appropriate dose and dose rate effectiveness factor. Increasingly, however, epidemiological studies on groups of workers in the nuclear industry are providing information on exposures at low doses and dose rates although at present any estimates of risk have large uncertainties associated with them. With the development of these national studies and by pooling them internationally these uncertainties should be progressively reduced. Landolt-Börnstein New Series VIII/4
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Continued follow-up of exposed populations, in particular the A-bomb survivors in Japan is needed for validating current lifetime projection models. It seems unlikely that epidemiological studies will be able to answer all the questions concerned with the effects of dose, dose rate, radiation quality and individual sensitivity to cancer induction. Ultimately this must depend on a much better understanding of the response of tissues to radiation. This will come increasingly from cellular and molecular studies designed to understand the fundamental mechanisms involved in cancer induction. The assumption made for protection purposes is that the incidence of radiation-induced cancer and hereditary disease increases with the dose, with no threshold. Thus it is not possible to completely prevent any risks. Protection standards must, therefore, be set to keep any risk to an acceptable level. Table 2.15 gives a chronology of the first century of radiation protection. Table 2.15. Chronology of the first century of radiation protection Year Event 1895 Discovery of X-rays (November 8) 1896 X-ray report made public (January 3) Discovery of radioactivity (February) First reports of possible X-ray injury; Damage to eyes (March 3) Skin effects first noted (April 18) 1901 X-ray lethality to mammals demonstrated experimentally 1904 First death in X-ray pioneer attributed to cumulative overexposure (October) 1906 Law of radiosensitivity of tissues put forth 1911 1915 1920 1921 1925 1927 1928 1929 1931 1932 1934 1936 1941 1944 1950 1991
International radium standard and Curie unit British Roentgen Society adopts radiation protection recommendations First Standing X-ray Protection Committee British X-ray and Radium Protection Committee issues first memorandum First “tolerance dose” proposed Genetic effects of X-rays shown (drosophila) Roentgen unit formally adopted, International X-ray and Radium Protection Committee formed (forerunner of ICRP) US Advisory Committee on X-ray and Radium Protection formed (forerunner of NCRP) USACXRP publishes first recommendations - 0.2 R/day Concept of greater permissible dose for partial body irradiation (hands) introduced Discovery of neutrons ICXRP recommends permissible dose of 0.2 R/day USACXRP recommends reduction in permissible dose to 0.1 R/day USACXRP recommends adoption of maximum body burden of 0.1 µCi for radium Suggested maximum permissible dose of 0.02 R/day Maximum permissible concentration for inhaled radioactivity introduced Rem and Rep introduced ICRP set up Publication of 1990 Recommendations of ICRP
Investigator Roentgen Becquerel Edison Morton Stevens Rollins Dally Bergonie Tribondeau Curie ARRS Mutscheller Müller
Failla Chadwick
Taylor Parker Parker
Please Note: Work on preparation of this Chapter was completed in March 2002. Readers should note that ICRP is planning to issue new recommendations on radiological protection in 2005. This will take into account more recent information on health effects published since its 1990 recommendations [91I2].
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2.9 References 72U1 74S2 75B6 77I1 77U2 80B2 81O1 82U3 84O2 84V1 85N1 85S4 86B1 86M1 86R2
86U4 87K3 87P3 88B3 88O3 88S1 88U5 88W1 88Y1
UNSCEAR: Ionizing Radiation : Levels and effects. 1972 Report to the general assembly, with annexes. Vol. II: Effects. New York: United Nations, 1972. Searle, A.G.: Adv. Radiat. Biol. 4 (1974) 121. Bithell, J.F., Stewart, A.M.: Br. J. Cancer 35 (1975) 271. ICRP.: Recommendations of the international commission on radiological protection. ICRP Publication 26. Oxford: Pergamon Press, 1977. UNSCEAR: Sources and effects of ionising radiation. 1977 report to the general assembly, with annexes, 1977. BEIR III.: Committee on the biological effects of ionising radiations. The effects on populations of exposure to low levels of ionizing radiation. National Academy of Sciences. Washington DC: National Academy Press, 1980. OPCS. Cancer statistics: Incidence, survival and mortality in England and Wales. Studies on medical and population subjects, No. 43, London: HMSO, 1981. UNSCEAR: Ionizing radiation sources and biological effects. Report to the general assembly, with annexes, 1982. Otake, M., Schull, W.J.: Br. J. Radiol. 57 (1984) 409. van Kaick, G., Muth, H., Kaul, A.: The German thorotrast study. Report No. EUR 9504 EN. Luxembourg CEC (quoted in BEIR IV), 1984. National council on radiation protection and measurements, NCRP, Report No. 80. Induction of thyroid cancer by ionising radiation, 1985. Shore, R.E.: J. Natl. Cancer Inst. 74 (1985) 1177. Barendsen, G.W.: Responses of cultured cells, tumours and normal tissues to radiations of different linear energy transfer. IN: Current Topics in Radiation Research, Vol. 4, pp 293-356 (Ebert and Howard, eds). North-Holland Publishing Company, Amsterdam. Mays, C.W., Spiess, H., Chemelevsky, D., Kellerer, A.; in: Proc. Symp. "The Radiobiology of Radium and Thorotrast", Gössner, W., Gerber, G.B., Hagen, J., Luz, A. (eds.), Neuherberg, 2931 Oct, 1984. Strahlentherapie 80, Suppl. 14-21 (1986) 27. Rundo, J., Keane, A.T., Lucas, H.F.: Current (1984) status of the study of 226Ra and 228Ra in humans at the Center for Human Biology; in: Proc. Symp. "The Radiobiology of Radium and Thorotrast", Gössner, W., Gerber, G.B., Hagen, U., Luz, A. (eds.), Neuherberg 29-31 Oct, 1984; Strahlentherapie 80 Suppl. 14-21 (1986). UNSCEAR. Genetic and Somatic Effects of Ionizing Radiation. 1986 Report to the General Assembly, with annexes (1986). Knox, E.G., Stewart, A.M., Kneale, G.W., Gilman, E.A.: J. Soc. Radiol. Prot. 7 (1987) 177. Preston, D.L., Pierce, D.A.: The effects of changes in dosimetry on cancer mortality risk estimates in the atomic bomb survivors. Hiroshima, Radiation Effects Research Foundation, RERF TR 0-87, 1987. BEIR IV.: Health risks of radon and other internally deposited alpha-emitters. Washington DC: National Academy Press, 1988. Otake, M., Yoshimaru, H., Schull, W.J.: Severe mental retardation among prenatally exposed survivors of the atomic bombing of Hiroshima and Nagasaki: A comparison of the T65DR and DS86 dosimetry systems. Hiroshima, Radiation Effects Research Foundation, RERF TR16-87, 1988. Schull, W.J., Otake, M., Yoshimaru, H.: Effect on intelligence test score of prenatal exposure to ionising radiation in Hiroshima and Nagasaki: A comparison of the T65DR and DS86 dosimetry systems. Hiroshima, Radiation Effects Research Foundation, RERF TR3-88, 1988. UNSCEAR: Sources, effects and risks of ionizing radiation. 1988 Report to the general assembly, with annexes, 1988. Ward, F.: Prog. Nucleic Acid Res. Mol. Biol. 35 (1988) 95. Yoshimoto, Y., Kato, H., Schull, W.J.: Lancet 2 (1988) 665.
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89P2 90B4 90N2 90N3 90S3 91I2 91I3 91V3 92K2 93I4 93M4 93U6 94G2 94K1 94R1 94S5 94U7
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Bithell, J.F., Epidemiological studies of children irradiation in utero. J.F. Bithell (1989), in: Low dose radiation: biological bases of risk assessment, Baverstock, K.F., Stather, J.W. (eds), London: Taylor and Francis, 1989. Edwards, A.A., Lloyd, D.C., Prosser, J.S.: Chromosome aberrations in human lymphocytes - a radiobiological review; in: Low dose radiation: biological bases of risk assessment, Baverstock, K.F., Stather, J.W. (eds),. London: Taylor and Francis, 1989, p. 423. Gilbert, E.S., Fry, S.A., Wiggs, L.D., Voelz, G.L., Cragle, D.L., Petersen, G.R.: Radiat. Res. 120 (1989) 19. Kronenberg, A., Little, J.B.: Mutagenic properties of low doses of X-rays, fast neutrons and selected heavy ions in human cells; in: Low dose radiation: Biological bases of risk assessment, Baverstock, K.F., Stather, J.W. (eds), London: Taylor and Francis, 1989, p. 423. Miller, A.B., Howe, G.R., Sherman, G.J., Lindsay, J.P., Yaffe, M.J., Dinner, P.J., Risch, H.A., Preston, D.L.: N. Engl. J. Med. 321 (1989) 1285. Muirhead, C.R., Kneale, G.W.: J. Radiol. Prot. 9 (1989) 209. Neel, J.V., Schull, W.J., Awa, A.A., Satoh, C., Otake, M., Kato, H., Yoshimoto, Y.: The genetic effects of the atomic bombs: Problems in extrapolating from somatic cell findings to risk for children. Low dose radiation: Biological bases of risk assessment, Bayerstock, K.F., Stather, J.W. (eds.), London: Taylor and Francis, 1989, p. 42. Pierce, D.A., Vaeth, M.: Cancer risk estimation from the A-bomb survivors: Extrapolation to low doses, use of relative risk models and other uncertainties; in: Low dose radiation biological bases of risk assessment, Baverstock, K.F., Stather, J.W. (eds.), London: Taylor and Francis, 1989, p 54. BEIR V.: Health effects of exposure to low levels of ionizing radiation. Washington DC: National Academy Press, 1990. NCRP.: Influence of dose and its distribution in time on dose-response relationships for lowLET radiation. Washington DC: NCRP Report No. 64, 1980. NCRP.: The relative biological effectiveness of radiations of different quality. NCRP Report No. 104, 1990. Shimizu, Y., Kato, H., Schull, W.J.: Radiat. Res. 121 (1990) 120. ICRP.: 1990 Recommendations of the international commission on radiological protection. ICRP Publication 60. Annals of the ICRP 21, No.1-3. Oxford: Pergamon Press, 1991. ICRP.: The biological basis for dose limitation in the skin. ICRP Publication 59. Annals of the ICRP 22, No. 2. Oxford: Pergamon Press, 1991. Voelz, G.L., Lawrence, J.N.P.: Health Phys. 61 (1991) 181. Kendall, G.M., Muirhead, C.R., MacGibbon, B.H., O’Hagan, J.A., Conquest, A.J., Goodill, A.A., Butland, B.K., Fell, T.P., Jackson, D.A., Webb, M.A., Haylock, R.G.E., Thomas, J.M., Silk, T.J.: Br. Med. J. 304 (1992) 220. ICRU.: Quantities and units in radiation protection dosimetry. Publication 51, 1993. Muirhead, C.R., Cox, R., Stather, J.W., MacGibbon, B.H., Edwards, A.A., Haylock, R.G.E.: Docs. NRPB 4(4) (1993) 15-157. UNSCEAR: Sources, effects and risks of ionizing radiation. 1993 Report to the general assembly, with annexes, 1993. Goodhead, D.T.: Int. J. Radiat. Biol. 65 (1) (1994) 7-17. Kellerer, A.M., Burkhart, W. (eds.): Radiation exposure in the Southern Urals. The Science of the Total Environ. Special Issue, 1994. Rowland, R.E.: Radium in Humans. A Review of US Studies. ANL/ER-3 UC-408 Argonne National Laboratory, 1994. Spiess, H.: The Ra-224 Study: Past, Present and Future; in: Proc. Int. Seminar on Health Effects of Internally Deposited Radionuclides: Emphasis on Radium and Thorium, Heidelberg, Germany, 1994. UNSCEAR: Sources, effects and risks of ionizing radiation. 1994 Report to the general assembly, with annexes, 1994.
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95C1 95C2 96E2 96I5 96P1 97D1 97D2 98B5 98M2 99B8 99M6 00U8 01U9 03I1
2 Biological effects of ionising radiation van Kaick, G., Wesch, H., Luehrs, H., Liebermann, D., Kaul, A.: Epidemiological results and dosimetric calculations - an update of the German thorotrast study; in: Proc. Int. Seminar on Health Effects of Internally Deposited Radionuclides: Emphasis on Radium and Thorium, Heidelberg, Germany, 1994. Cardis, E., Gilbert, E.S., Carpenter, L.: Radiat. Res. 142 (1995) 117. Cox, R., Muirhead, C.R., Stather, J.W., Edwards, A.A., Little, M.P.: Docs NRPB 6 (1) (1995). EURATOM.: Basic safety standards for the protection of the health of workers and the general public against the dangers arising from ionizing radiation. Council Directive 96/29/Euratom at 13 May 1996. Off. J. Eur. Commun. L159 39. IAEA.: International basic safety standards for protection against ionising radiation and for the safety of radioactive sources. Jointly sponsored by FAO, IAEA, ILO, NEA/OECD, PAHO and WHO. Vienna, IAEA, Safety Series No. 115, 1996. Pierce, D.A., Shimizu, Y., Preston, D.L., Vaeth, M., Mabuchi, K.: Radiat. Res. 146 (1996) 1. Delongchamp, R.R., Mabuchi, K., Yoshimoto, Y.: Radiat. Res. 147 (1997) 385. Doll, R., Wakeford, R.: Br. J. Radiol. 70 (1997) 130. BEIR VI.: Effects of exposure to Radon. Washington DC: National Academy Press, 1988. Mill, A.J., Frankenberg, D., Bettega, D., Hiever, L., Saran, A., Allen, L.A.: J. Radiol. Prot. 18 (2) (1998) 79. Boice, J.D., Miller, R.W.: Teratology 59 (1999) 227. Muirhead, C.R., Goodhill, A.A., Haylock, R.G.E.: J. Radiol. Prot. 19 (1999) 3. UNSCEAR: Sources and effects of ionizing radiation. 2000 Report to the general assembly, with scientific annexes. Vol II: Effects, 2000. UNSCEAR: Hereditary effects of radiation. 2001 Report to the general assembly, with scientific annex, 2001. ICRP. Biological Effects after Prenatal Irradiation (Embryo and fetus). ICRP Publication 90. Annals of the ICRP No. 1-2 Oxford, Pergamon Press (2003).
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This chapter introduces the basic properties of atomic nuclei, their structure characteristics and transformations by radioactive decay and nuclear reactions. The particle and electromagnetic radiation associated with nuclear transformation is classified and their interaction with matter briefly described. Special emphasis is put on fission products leading to the radioactive inventory of nuclear reactors.
3.1 Natural radioactivity Radioactivity was observed for the first time in 1896, when Henri Becquerel [96Bec] in Paris found that photographic plates were blackened even in the absence of light, if they were in contact with uranium containing minerals. In 1898 the same observation was made for thorium by Marie Curie [98Cur] in France and by G. C. Schmidt [98Sch] in Germany. Marie Curie found differences in the radioactivity of uranium and thorium and concluded that these elements must contain unknown radioactive elements. Together with her husband, Pierre Curie, she discovered polonium in 1898, and radium in the same year. More historical details can be found e.g. in [95AMC]. Since radioactive radiation cannot be detected by human senses, it is necessary to use for its detection and the identification of radioactive substances solid, liquid, or gaseous detectors, which indicate the radiation via its interaction with the respective materials. Radioactive elements are widely distributed in the earth's crust in more or less small concentrations. They originate on one hand from extremely long-lived primordial radionuclides formed together with the other stable elements and from their longer or shorter lived decay products like the important minerals of uranium and thorium and their radioactive decay chains. Uranium and thorium are common elements in nature. Their concentrations in the crust of the earth are about 4 and 13 mg/kg, respectively, and the concentration of uranium in seawater is about 3 µg/l. The most important uranium mineral is pitchblende (U308). The most important thorium mineral is monazite, which contains between about 0.1 and 15 % Th. Natural radioactivity is mainly observed with heavier primordial elements and seldom with light ones (e.g. 40K and 87Rb). On the other hand, radioactive elements can be formed - especially in case of light shorter lived elements - by interactions of cosmic radiation with the atmosphere. This radiation entering the atmosphere from outside originates from the sun as well as from material in the deep interstellar space. It produces a variety of elementary particles (protons, neutrons, electrons, positrons, mesons, photons) and of radioactive atoms via cascades of interactions with the gas molecules in the atmosphere. Among others, 14C, 10Be , 7Be, and 3H (tritium) are produced in the atmosphere by cosmic radiation. Examples for the production rate are: about 2.2 × 104 atoms of 14C per second per m2 of the earth's surface and about 2.5 × 103 atoms of tritium per second and m2. Taking into account the radioactive decay and the dwell times in the atmosphere, these data result in a global equilibrium inventory of about 6.3 × 104 kg of 14 C and of about 3.5 kg of 3H. The measurement of local concentrations of natural radionuclides like 14C, 40K, 87Rb is very useful for the determination of the age of the respective material. However, other sources have to be taken into account like the production of these nuclides in nuclear power plants and by explosions of nuclear bombs. Landolt-Börnstein New Series VIII/4
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3.2 Elements, isotopes and radionuclides 3.2.1 Atoms, electrons and the Periodic Table of Elements 3.2.1.1 The atom The atom is the smallest particle of an element, which can no more be subdivided by chemical methods. It consists of a (heavy) nucleus surrounded by a number of (light) electrons. The elements differ from each other by the structure of their atoms i.e. the composition of the nucleus and the number and distribution of the electrons. The diameter of an atom is approximately 10−8 cm. The nucleus contains positively charged protons and electrically neutral neutrons and has, therefore, a positive charge. Its diameter is in the order of 10−13 cm, which is five orders of magnitude smaller than the atomic diameter. Since atoms in their normal state are electrically neutral, the number of (negative) electrons in the shell must balance the number of (positive) protons in the atomic nucleus. 3.2.1.2 The electron shells The electron distribution around the atomic nucleus is described and denoted according to Bohr's shell model. Electrons move in discrete shells with energy states (binding energies) decreasing with increasing distance from the nucleus. They are numerated beginning with n = 1 and denominated by capital letters K, L, M, .... The maximum number of electrons in one shell is 2n. By energy transfer electrons can be excited into higher lying weaker bound states. The electrons are released into an unbound state by an energy transfer of at least the binding energy of the respective shell leaving the residual atom in an ionised state. The remaining hole in the electron shell is filled by one of the outer electrons releasing the energy difference in form of characteristic photon radiation (fluorescence) or by radiationless energy transfer to another bound electron, which in turn is emitted (Auger electron). 3.2.1.3 The Periodic Table of Elements When ordering all elements according to their atomic weight, one finds a periodical repetition of special properties leading to several groups of elements with a remarkable similarity of their chemical and physical properties. Within these groups very similar differences of the atomic weights of neighbouring group members are found. The most outstanding element group is formed by the noble gases, similar observations are made for the halogens fluorine, chlorine, bromine, and iodine, as well as for the chalcogens sulphur, selenium, and tellurium. First indications of this periodical behaviour were already described in the early 19th century. In 1869 L. Meyer and D. Mendeleev succeeded in placing all known elements satisfactory in the periodical system. In this periodical system all elements, except of the rare earth elements (lanthanides) and the transuranium elements (actinides) are sorted into 6 periods. This ordering shows two short periods with 8 elements each and three long periods comprising 18 elements, taking into account some not naturally occurring elements, like technetium. The three long periods are subdivided into two parts. Between these one finds three triples of very similar elements: Fe-Co-Ni, Ru-Rh-Pd, Os-Ir-Pt, respectively. In this arrangement chemically similar elements form columns combined in 8 groups, subdivided into main groups and auxiliary groups of transition elements. The 8th main group containing the noble gases by convention is called group 0. This ordering was explained later on by Bohr's atomic model. The periods correspond to the filling of the different electronic shells, which can take up 2n electrons each. The noble gases in group 0 correspond to a completely filled shell in each period. Landolt-Börnstein New Series VIII/4
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3.2.2 Atomic nuclei, nuclides and the Chart of Nuclides 3.2.2.1 Nuclei and nuclides Atomic nuclei are composed of protons and neutrons having the common denotation nucleons. The corresponding number of protons Z is called the atomic number. It characterizes the element and its chemical behaviour, while the physical properties are defined by Z as well as by the neutron number N. Atoms with a defined nucleus and the corresponding number of electrons are called nuclides. Each nuclide is identified by its mass number A = Z + N and is marked by the following denotation for the element symbol El: A A 4 235 He = He - 4 = 2 He42 , 235U = U - 235 = 92 U143 Z El N or in shorter form El (or El-A), e.g.: since Z defines uniquely the element and N is given as well by N = A − Z. The following denotation is used for neighbouring nuclides (cf. Fig. 3.1): - nuclides with equal
atomic number Z are called isotopes mass number A are called isobars neutron number N are called isotones - nuclides in long-lived excited states are called isomers. Different isotopes of an element exhibit almost the same chemical behaviour (very small differences are observable in special experiments only: “isotopic effect”). In general, the elements represent a mixture of all of their stable isotopes.
Atomic number Z
Isobars
Isotones
49 48 47
74 Isotopes
45 44
72
43 60
62 64 66 68
70
Fig. 3.1. Representation of nuclides in a chart of nuclides as proposed by E. Segrè and definition of isotopes, isotones, and isobars.
Neutron number N
Today nearly 3000 nuclides are known most of which are extremely short-lived and can be produced in complicated experiments only. More than 300 nuclides are found in nature and can be assigned to one of four categories: 1. 258 stable nuclides 2. 25 nuclides with Z <80 having extremely long half-lives (radioactive decay not proved uniquely in all cases) 3. about 15 quasistable nuclides with half-lives >105 years (including 238U, 235U, 232Th and 244Pu) existing since the genesis of the elements and called primordial radionuclides. Among these 238U, 235 U and 232Th and their radioactive decay products form the main sources of natural radioactivity 4. radionuclides continuously produced by the impact of cosmic radiation: e.g. 14C, 10Be, 7Be, 3H. 3.2.2.2 The Chart of Nuclides In order to get a synopsis of this vast amount of nuclides several types of schematic representations have been introduced in the past. The most widespread scheme has been proposed by E. Segrè and has been realized among others in the Karlsruhe Chart of Nuclides [98PKS]. In this scheme the nuclides are arranged in such a way (cf. Fig. 3.1) that the proton number Z is given on the ordinate and the neutron number N = A − Z on the abscissa, respectively. Each experimentally observed nuclide is represented by a square containing the symbol of the element and the number of nucleons (mass number) and in addition Landolt-Börnstein New Series VIII/4
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various decay or structure data. For example, in the Karlsruhe Chart of Nuclides [98PKS] besides the half-lives, the decay modes, the types of emitted radiation and the energies of the most abundant radiation types are given as observable in respective experiments. Other charts, nowadays often distributed in electronic form (e.g. [99Mag]), are based on (evaluated) nuclear data files enabling access to all known decay data of all known radionuclides. In these charts the stable nuclides lie around a central branch with approximately N = Z for the lighter nuclides. With increasing mass this branch flattens due to the increasing neutron excess necessary for the stability of heavier nuclei. This branch (valley of beta-stability) separates the neutron rich nuclides undergoing β−-decay from the proton rich nuclides undergoing β+-decay. Furthermore, with increasing distance from the valley of stability, i.e. with increasing neutron- or proton excess, the stability and consequently the half-lives of the nuclides decrease drastically. Thus, radionuclides with half-lives in the range of hours or longer are only found in narrow bands along the valley of stability. The so called strong interaction, by which the nucleons are bound in the atomic nucleus, is a very short ranged pairing force binding the nucleons very close together. This leads to a very high density of the nuclear matter. With increasing mass number stable nuclei exhibit an increasing neutron excess (up to 50 % for the heaviest nuclei) necessary to balance the increasing repulsive force of the positively charged protons. Nuclei − like the electron shells − can be excited by energy transfer into higher energetic states. The de-excitation of such states occurs under emission of characteristic nuclear radiation, which is in general γ-radiation. At sufficiently high excitation energies (above the separation energy of nucleons) the emission of nucleons or nucleon clusters becomes possible.
3.3 The structure of the atomic nucleus 3.3.1 Elementary particles Though the modern high energy and particle physics deals with elementary particles far below the level of those forming the atomic nuclei, only the latter shall be discussed here and are − according to classical nuclear physics – denominated as elementary particles of the nucleus, the nucleons. 3.3.1.1 Charge, mass and stability of the nucleons The proton is stable and has a positive electric charge equal and opposite to the elementary charge. The neutron is electrically neutral and its mass is slightly larger than that of the proton. The free unbound neutron is unstable and decays with a half-life of 10.25 min into a proton and an electron and an antineutrino. The following units are used for atomic and nuclear quantities (see also Table 3.1), making use of the energy-mass equivalence in Einstein's equation E = mc2, which leads to the correspondence 1 kg = 5.6095892 × 1029 MeV: elementary charge: e = 1.602176462×10−19 C mass: atomic mass unit: u = 1.66053873×10−27 kg = 9.31494013×102 MeV/c2, where the basis of this unit is 1/12 of the mass of the neutral 12C atom (including the masses of its 6 electrons).
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Table 3.1. Frequently used general constants (main values taken from [99MoT]) and conversion factors. The standard deviations are given in parentheses. Symbol c h ħ = h/2π u mn
value
quantity −1
299792458 m s 6.62606876 (52)×10−34 J s 1.054571596 (82)×10−34 J s 1.66053873 (13)×10−27 kg 9.31494013 (37)×102 MeV/c2 1.00866491578 (55) u
speed of light Planck constant atomic mass unit neutron rest mass
−27
mp
1.67492716 (13)×10 kg 9.39565330 (38)×102 MeV/c2 1.00727646688 (13) u
proton rest mass
−27
me e e/me re = e2/mec2 a0
α R∞ NA Vm R k = R/NA F = NA·e
1.67262158 (13)×10 kg 9.38271998 (38)×102 MeV/c2 9.10938188 (72)×10−31 kg 0.510998902 (21) MeV/c2 1.602176462 (63)×10−19 C 4.803204197 (19)×10−10 esu 1.758820174 (71)×1011 C kg−1 5.27281023 (21)×1017 esu g−1 2.817940285 (31)×10−15 m 5.291772083 (19)×10−11 m 7.297352533 (27)×10−3 1.0973731568549 (83)×107 m−1 6.02214199 (47)×1023 mol−1 2.2413996 (39)×10−2 m3 mol−1 8.314472 (15) J mol−1 K−1 8.314472 (15)×107 erg mol−1 K−1 1.3806503 (24)×10−23 J K−1 8.617342 (15)×10−5 eV K−1 9.64853415 (39)×104 C mol−1 2.892557769 (12)×1014 esu mol−1
electron rest mass elementary charge specific electron charge electron radius Bohr radius fine structure constant Rydberg constant Avogadro constant molar volume of an ideal gas at s. t. p. molar gas constant Boltzmann constant Faraday constant
Some useful conversions: 1 Ci = 3.7×1010 Bq (= disintegrations s−1) 1 W ^ 3.1×1010 fissions s−1 1 MWd ^ 2.7×1021 fissions ^ 1g fissionable material
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1 µA = 6.241509745×1012 e s−1 1 eV/Atom ^ 23 kcal mol−1 1 MeV Ci ^ 5.93×10−3 W
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Average binding energy per nucleon B / A [MeV ]
9 8 7 6 5 4 3 2 0
Fig. 3.2. Average binding energy per nucleon in dependence on the mass number. 8 16 24 30
60 90 120 150 180 210 240 Mass number A
3.3.1.2 Binding energy The experimentally measured atomic weights M of stable pure isotopes are not integer multiples A of the atomic mass unit u. This deviation is called the mass defect ∆M = M – A·u. In addition, also the mass M(Z, A) of a nucleus differs from the mass sum of its nucleons and is always smaller: M(Z, A) < Z·Mp + N·Mn. According to Einstein's equation, the mass difference ∆M = Z·Mp + N·Mn − M(Z, A) corresponds to an energy B = ∆M·c2 = [Z·Mp + N·Mn − M(Z, A)]·c2, which is necessary to separate all nucleons of the nucleus. B is called the total binding energy of the nucleus and is always positive for stable nuclides. B/A is the average binding energy per nucleon, which has a mean value of about 8 MeV/nucleon, except for the very light and heavier elements (cf. Fig. 3.2). This behaviour has the most important consequence that energy can be gained by fission of heavy as well as by fusion of light nuclei, both leading to a higher energy for the product(s).
3.3.2 Nuclear transformations 3.3.2.1 Nuclear reactions The vast amount of radioactive nuclides as shown in charts of nuclides do not occur in nature but can only be produced artificially by means of nuclear reactions. In a nuclear reaction a normally stable nuclide is bombarded with a beam of one of many types of radiation (charged particle or neutron radiation, electromagnetic radiation, electrons). The simplest form of a nuclear transformation is the radioactive decay caused by internal excess energy: A → B + b + ∆E. This is a mononuclear reaction. Binuclear reactions, denoted generally as “nuclear reactions”, are induced by bombarding target nuclei with a beam of specific projectiles. They are described by: A + a → B + b + ∆E, where A is the target nuclide, a the projectile, B the product nuclide, and b the “ejectile” (particle or photon emitted). The energy ∆E is also called the Q value of the reaction. The first nuclear reaction was observed in a cloud chamber in 1919 by Rutherford: 14
N + 4He → 17O + 1H, in shorter form N(α,p)17O or generally A(a,b)B. 14
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When comparing chemical and nuclear reactions one finds the following main differences: • In chemical reactions macroscopic amounts of matter as a whole are altered, in nuclear reactions individual atoms. • In chemical reactions the total involved mass is conserved, in nuclear reactions the sum Σ(E + mc2) of the energies and the mass-energy equivalents. • The energies of chemical reactions are relatively small and comparable with the energies of chemical bonds (of the order of eV), whereas the energies involved in nuclear reactions are about 6 orders of magnitude higher (of the order of MeV). Many nuclear reactions pass over a transition state, similarly to chemical reactions: A + a → (C)* → B + b + ∆E The transition state (C)*, which is excited to an elevated energy, is also called a compound nucleus; its lifetime is very short (<10−13 s). While the probability for a mononuclear reaction (i.e. for radioactive transmutation) is given by the decay constant λ, in the compound nucleus model the reaction probability is determined by both the probability that the projectile a will react with the nuclide A and by the probability that the nuclide B is produced. The time duration of a nuclear reaction is in between about 10−23 and 10−13 s. The lower limit is given by the crossing time of a particle passing the nucleus with the velocity of light, and the upper limit holds for slow reactions, e.g. with thermal neutrons. 3.3.2.2 Projectiles for nuclear reactions Charged particles Positively charged particles inducing nuclear reactions, such as protons, deuterons or ions with higher atomic numbers Z, need a minimum energy to surmount the repulsive Coulomb barrier formed by the protons of the nuclei. The Coulomb barrier U can be calculated approximately from the equation U≈
Z A Za e2 Z Z ≈ 1/ 3 A a 1/ 3 4πε 0 r AA + Aa
(3.3.1)
where ZA and Za are the charge numbers of the nuclide A and the projectile a, respectively, e is the unit charge, ε0 the electric field constant, and r the distance, which in this approximation is set to be r ≈ r0(AA1/3 + Aa1/3)
(3.3.2)
AA and Aa are the mass numbers of the nuclide A and the projectile a, respectively. For the Coulomb barrier U the following approximate values can be calculated: • • • • •
U ≈ 1.8 MeV for the reaction of a proton with 12C, U ≈ 13 MeV for the reaction of a proton with 238U, U ≈ 24 MeV for the reaction of an α-particle with 238U, U ≈ 130 MeV for the reaction of 12C with 238U and U ≈ 700 MeV for the reaction of 238U with 238U.
With increasing atomic numbers the approximate formula (3.3.1) becomes incorrect. For the reaction of 238U with 238U, e.g., a more exact value for the Coulomb barrier is U ≈ 1500 MeV.
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Charged particles for nuclear reactions originate from particle accelerators. They are produced in ion sources by bombarding a gas with energetic electrons. The positive ions are extracted by means of an electrode. For the further acceleration various set-ups are used. The two main groups are linear accelerators (linacs) with single (e.g. Cockroft-Walton type) and multiple (e.g. tandem Van de Graaff accelerators) stages and circular accelerators. With single-stage linear accelerators proton or deuteron beams of up to about 10 mA with energies of up to about 4 MeV are obtained. These accelerators are often applied as injectors in larger machines for the production of high-energy particles. In two-stage tandem Van de Graaff accelerators proton beams of about 20 MeV and α-particles of about 30 MeV are obtained. In addition, lighter heavy ions can be efficiently accelerated, too. The intensity of the beam current varies between about 10 and 100 µA. In cyclotrons the ions move on spiral paths with increasing radius guided by a suitable magnetic field, while in synchrotrons the circular orbit of the particles remains constant and the guiding magnetic field increases with the momentum of the ions. In modern machines, protons, deuterons and α-particles can be accelerated to energies of several 100 MeV up to about 1 GeV. For the production of radionuclides, relatively small accelerators are used producing particle energies of 10 to 30 MeV and ion currents of the order of 100 µA. With light charged particles mainly nuclides on the left hand side of the valley of β-stability are produced (β+- and electron capture activities). Increasing projectile energies lead then to an increasing neutron deficit of the produced radionuclides. Heavier ions are produced in special types of linear or circular accelerators. The term “heavy ions” is used in this context for all ions heavier than α-particles and includes “light ions” e.g. of lithium, carbon or oxygen as well as “heavy ions” of elements up to uranium. These projectiles are widely used in basic as well as in applied research and techniques. Due to the great number of nucleons, which they can transfer to the target nuclide, the main application fields of heavy ions are: synthesis of new (including the socalled super-heavy) elements, production of nuclides far off the line of β-stability, investigation of nuclear matter at high densities, production of small holes with defined diameters in thin foils and irradiation of tumors in medicine. Neutrons Neutrons are frequently used projectiles for nuclear reactions. Since they do not carry a positive charge, they do not experience Coulomb repulsion, and even low-energy (thermal and slow) neutrons can easily enter the target nuclei. Neutrons with energies of the order of 1 to 10 eV (resonance neutrons) exhibit relatively high absorption maxima. Furthermore, neutrons are available in large quantities in nuclear reactors with fluxes of the order of about 1010 to 1016 cm−2 s−1. Neutrons may also be produced by nuclear reactions, such as 9Be(α,n)12C, 9Be(d,n)10B, 9Be(γ,n)2α, d(γ,n)p, d(d,n)3He or t(d,n)α. Alpha particles are available from radionuclides like 226Ra or 210Po, and RaBe neutron sources have formerly often been applied in experimental neutron physics. Gamma-rays of sufficiently high energy for (γ,n) reactions may be supplied by radioactive nuclides, such as 124Sb. Neutrons are also available from spontaneously fissioning nuclides like 252Cf (1 µg of 252Cf emits 2.3 × 106 neutrons per second). In neutron generators deuterons with energies between 0.1 and 10 MeV are produced in small accelerators and impinge on a suitable target, e.g. a tritium target. The neutron fluxes available from neutron sources and neutron generators vary between about 105-108 and 108-1011 cm−2s−1, respectively. Neutrons produced by the reaction t(d,n)α have energies of about 14 MeV. Reactions with neutrons are used to produce nuclides on the right-hand side of the line of β-stability, i.e. nuclides undergoing β−-decay.
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Electrons and γ-rays For the acceleration of electrons to high velocities approaching the velocity of light, much less energy is needed than for the acceleration of ions. They are accelerated in linear or circular (betatrons, synchrotrons) accelerators. Depending on the size of the installations, energies of the order of 10 to 300 MeV are obtained in linear accelerators and betatrons, whereas in electron synchrotrons electrons with energies up to the order of 10 GeV are available. Electrons are not directly used for the production of radionuclides. However, the high-energetic bremsstrahlung, which is emitted, when the electrons are slowed down in a target of high atomic number, is applied for the induction of nuclear reactions (photo-nuclear reactions). The maximum energy of this bremsstrahlung corresponds to the energy of the incident electrons and is in the range of high-energy γ-rays. Gamma rays are also available from γ-emitters, such as 60Co or 124Sb, but with comparably lower energies and intensities. The monoenergetic synchrotron radiation emitted by electrons moving in the magnetic field of a synchrotron has found many applications because of the high intensities, broad energy ranges, and small beam diameters available. The energy of the synchrotron radiation is in the range of the energy of X-rays. 3.3.2.3 Artificial radionuclides for medical and technical applications Numerous radionuclides are especially produced at accelerators or nuclear reactors for dedicated applications in medicine and technology. In particular, in nuclear medicine new diagnostic methods have been established and are applied with increasing frequency such as single photon emission computed tomography (SPECT) and positron emission tomography (PET). The main advantage of these methods as compared to other imaging processes is their capability not only to provide an image of inner organs but also to give information about their function (metabolism). These methods use short living accelerator isotopes like 11C, 13N, 18F, 123I or 153Gd. The diagnostical goals are detection of malfunction or cancer stroke of thyroid gland, kidneys, liver, heart and brain. For so-called targeted radiotherapy of cancer various β-emitters e.g. 90Y, 125I and 107Pd are being used. Quite recently, α-emitters like 211At, and 213Bi/225Ac are under development for such purposes which promise to have thousands fold better biological effectiveness. Technical application of primordial (241Am) and artificial radionuclides (e.g. 55Fe, 57, 60Co, 137Cs) range from smoke detection, measurement of moisture content in materials via thickness measurements for various materials to wear diagnostics of machine parts (57Co). The latter method is industrially applied by many car factories. 3.3.2.4 Excited states, level- and decay schemes Like the electrons, which can occupy only discrete states in the shell with defined energies, also the nucleons in the atomic nucleus are in discrete states with well-defined energies. The state with the lowest energy is called the groundstate, its energy is per definition set to zero and all higher lying states (excited levels) are referred to this groundstate. In many cases of radioactive decay and even more often after nuclear reactions the product of the transformation is not in the groundstate but rather in an excited state. These and other excited states correspond to a series of defined energy levels, which are specific for each nuclide. They release their excitation energy by emission of one or several γ-rays, mostly within about 10−13 s after their formation by a nuclear reaction or by a preceding α- or β-decay. In some special cases, e.g. for very low transition energies, γ-transitions can be “hindered” (forbidden transitions) resulting in longer half-lives of these metastable “isomeric states”. In order to visualize the level structure of a nuclide, the levels are ordered in a scheme according to their energy above the ground state and are denoted − besides the energy − by the specific quantum numbers nuclear spin and parity, resulting from the nucleons contributing to the excitation. In addition, Landolt-Börnstein New Series VIII/4
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the assigned γ-transitions and for isomeric states the half-lives are indicated. In the decay schemes the nuclides with higher atomic number are ordered to the right hand side and the decay types and transition energies to the levels in the daughter nuclide are specified. The construction of level schemes is performed either directly by sophisticated nuclear reaction experiments and/or by a comprehensive investigation of the radioactive decay of the mother nuclide feeding several states with subsequent γ-decay. While for light radionuclides the decay schemes are mostly relatively simple due to the fact that only few levels can be fed by the decay of the mother nuclide, heavier nuclides exhibit mostly very complicated schemes with many levels and γ-transitions.
3.4 Radioactive decay 3.4.1 Basic properties 3.4.1.1 Decay law for a single radionuclide From the observation that the number of nuclei decaying per unit time is proportional to the actual number N of radioactive nuclei the following time law results: dN/dt = −λN
(3.4.1)
The constant of proportionality λ is characteristic for each nuclide and is called “decay constant”. It corresponds to the fraction of the actually present nuclei, which decays per unit time. The integration of equation (3.4.1) results in the general decay law: N(t) = N(0)·e−λt
(3.4.2)
The number of atoms of a nuclide decaying per unit time is called its decay rate (λN). It is measured in Bequerel (1Bq = 1/sec). The formerly used unit of radioactivity was the Curie (1 Ci corresponding to 3.7 × 1010 Bq). A specified activity refers to the respective nuclide only, independent of possible subsequent radioactive decays of daughter nuclides. The decay rate λN is usually set equal to the activity A. In general, however, an experimentally measured activity A is lower but always proportional to the decay rate of a sample A = FλN. This is due to the fact that the measured rate depends on the experimental setup, the sensitivity of the detector used and the thickness, backing and area of the sample. In the following, this factor F will be set to 1. The general decay equation becomes therefore:
λN(t) = λN(0)·e−λt
(3.4.3)
A(t) = A(0)·e−λt
(3.4.4)
or and ln A(t) = ln A(0) − λt
(3.4.4a)
where A(t) means the activity at time t and A(0) the activity at time t = 0. If the original number of atoms N(0) has decreased to one half, N(t) = N(0)/2 and t = T (= T⅛: halflife), equ. (3.4.2) gives λ = ln 2/T ≈ 0.693/T. With this expression for λ it follows from equ. (3.4.4a): ln A(t) = ln A(0) − ln(2)·t/T
(3.4.4b) Landolt-Börnstein New Series VIII/4
Ref. p. 3-39]
3 Physical fundamentals
3-11
The use of the half-life is often more convenient than that of the decay constant λ. Hence, in final expressions λ is replaced by ln 2/T. Half-lives are usually given in units of seconds (s), minutes (m), hours (h), days (d) and years (y). Examples of radioactive decays into a stable final state are: 60Co, 128I, ... 3.4.1.2 Branching radioactive decay Several nuclides can decay via more than one decay branch, e.g. by electron capture as well as by β− or β+-decay. The decay constant λ being characteristic for the nuclide's decay is in this case composed of the partial decay constants of the individual decay branches:
λ = λ1 + λ 2 + ... + λ n
(3.4.5)
Since the activity for each decay branch is Ai = λiN, the total activity of the nuclide is given by: or
A = A1+ A2 + ... + An = λ1N + λ 2N + ... + λ nN
(3.4.6)
A(t) = λ1N(0) e−λt + λ 2N(0) e−λt + ...+ λ nN(0) e−λt
(3.4.7)
The partial activities are proportional to the total number N of the respective nuclide decreasing with its corresponding half-life. Example: 64Cu → 64Ni + 64Zn (T=12.7 h; β+=19 %, β−=39 %, EC = 42 %) 3.4.1.3 Mixture of several nuclides without genetic relations The total activity of a mixture of radionuclides at any time is proportional to the sum of the individual activities at the respective time: A = A1 + A2 + ...+ An = A1(0)·e−λ1t + A2(0)·e−λ 2t + ...+ An(0)·e−λ nt
(3.4.8)
3.4.1.4 Activity of nuclides with genetic relations For activity chains with generic mother-daughter relations the activity of the mother nuclide (index 1) is given by equ. (3.4.4). For the determination of the daughter activity (index 2) one has to consider both the formation via decay of the mother nuclide with the decay constant λ1 and the decay of the daughter with its own decay constant λ 2: dN2/dt = λ1N1 − λ 2N2
(3.4.9)
The number of daughter nuclei at time t is: N 2 (t ) = N1 (0)
(e
λ1
−λ 1t
−e
−λ 2t
) + N ( 0) e
−λ 2t
2 λ 2 − λ1 λ2 −λ t −λ t −λ t A2 (t ) = A1 (0) e 1 − e 2 + A2 (0)e 2 or λ 2 − λ1
(
A2 (t ) = A1 (0)
T1 T1 − T2
)
t ln 2 t ln 2 ⎛ − t ln 2 ⎞ − − ⎜ e T1 − e T2 ⎟ + A (0) e T2 2 ⎜ ⎟ ⎝ ⎠
If at t = 0 the daughter activity is zero, equ. (3.4.10b) becomes Landolt-Börnstein New Series VIII/4
(3.4.10) (3.4.10a)
(3.4.10b)
3-12
3 Physical fundamentals
T1 ⎛⎜ − e A2 (t ) = A1 (0) T1 − T2 ⎜ ⎝
t ln 2 T1
−e
−
t ln 2 T2
⎞ ⎟ ⎟ ⎠
[Ref. p. 3-39
(3.4.11)
3.4.1.5 Secular radioactive equilibrium (T1 » T2) If the half-life T1 of the mother nuclide is much larger than that of the daughter nuclide T2 the latter can be discarded in the quotient of equ. (3.4.11): t ln 2 ⎛ − t ln 2 ⎞ − T1 ⎜ − e T2 ⎟ A2 (t ) = A1 (0) e ⎜ ⎟ ⎝ ⎠
(3.4.12)
If in addition t « T1, equ. (3.4.12) becomes: t ln 2 ⎛ ⎞ − ⎜ A2 (t ) = A1 (0) 1 − e T2 ⎟ ⎜ ⎟ ⎝ ⎠
(3.4.13)
what means that the increase of the daughter activity depends only on T2. After a period of about 6 - 7 T2 the daughter activity equals approximately that of the mother nuclide. The total activity for periods small when compared to T1 is constant and equals 2A1. Example: 90Sr → 90Y → 90Zr (T1=28.64 y, T2=64.1 h) 3.4.1.6 Transient equilibrium (T1 > T2) If T1 > T2 but both are of the same order of magnitude equ. (3.4.11) is valid. It can be simplified for periods t larger than about 6-7 T2, since the expression
e
−
t ln 2 T2
decreases much faster with time than
e
−
t ln 2 T1
− T A2 (t ) = A1 (0) 1 e T1 − T2
t ln 2 T1
(3.4.14)
The system reaches an equilibrium state, where the daughter activity is A2 =
T1 A1 T1 − T2
and decreases with the half-life of the mother T1. The daughter activity A2 for t = 0 equals zero and approaches zero again for very large times t. The time tm, where A2 has a maximum, is found from the first derivative of equ. (3.4.11): Landolt-Börnstein New Series VIII/4
Ref. p. 3-39]
tm =
3 Physical fundamentals
T1T2 T ln 2 (T2 − T1 )ln 2 T1
3-13 (3.4.15)
For the total mother- and daughter activity
Atotal = A1 + A2 = A1 (0)e
−
t ln 2 T1
+ A1 (0)
T1 T1 − T2
⎛ − t ln 2 − ⎜ e T1 − e ⎜ ⎝
t ln 2 T2
⎞ ⎟ ⎟ ⎠
(3.4.16)
the time for the maximum is found correspondingly: tm =
2
T1T2 T2 ln (T2 − T1 ) ln 2 2T1T2 − T2 2
(3.4.17)
Example: 132Te → 132I → 132Xe (T1=76.3 h, T2=2.3 h) 3.4.1.7 Similar half-lives (T1 ≈ T2) If the half-lives of mother and daughter nuclide are very similar, equ. (3.4.11) can no more be used. By means of series development one obtains: A2 (t ) = A1 (0)
t ln 2 − e T2
t ln 2 T2 ⎛
⎜⎜1 + ⎝
t (T1 − T2 )ln 2 ⎞ ⎟⎟ 2T1T2 ⎠
(3.4.18)
The term t (T1 − T2 )ln 2 2T1T2
can be neglected if it adopts values «1. Equation (3.4.18) has then the following form: A2 (t ) = A1 (0)
tln 2 − e T2
t ln 2 T2
(3.4.19)
For t (T1 − T2 )ln 2 ≅1 2T1T2
the bracket in equ. (3.4.18) cannot be neglected. For t (T1 − T2 )ln 2 >1 2T1T2
instead of equ. (3.4.18) equ. (3.4.11) has to be used again. For the special case T1 = T2 one obtains formally again equation (3.4.19).
Landolt-Börnstein New Series VIII/4
3-14
3 Physical fundamentals
[Ref. p. 3-39
From the first derivative of equ. (3.4.19) one gets the time tm of the maximum activity: tm =
T 1 = ln 2 λ
(3.4.20)
If the mother nuclide at time t = 0 did not contain any daughter activity, tm corresponds to the time where mother and daughter activities are equal. Example: 101Mo → 101Tc → 101Ru (T1=14.6 m, T2=14.2 m) 3.4.1.8 Half-life of mother nuclide shorter than half-life of daughter (T1 < T2) In case of T1 < T2 equ. (3.4.11) can be rewritten as:
T1 ⎛⎜ − A2 (t ) = A1 (0) e T2 − T1 ⎜ ⎝
t ln 2 T2
−e
−
t ln 2 T1
⎞ ⎟ ⎟ ⎠
(3.4.21)
In this equation T1/(T2−T1) <1, if 2T1 < T2. For times t « T2 the increase of the daughter activity A2 depends only on the half-life T1 of the mother nuclide in contrast to the cases of secular and transient equilibrium:
A2 (t ) = A1 (0)
− T1 ⎛⎜ 1− e T2 − T1 ⎜ ⎝
t ln 2 T1
⎞ ⎟ ⎟ ⎠
(3.4.22)
For times t > T1 the daughter activity decreases with its individual half-life: − T1 A2 (t ) = A1 (0) e T2 − T1
t ln 2 T2
(3.4.23)
In this case no equilibrium is reached, since the ratio of daughter and mother activities increases permanently as function of time: A2 (t ) T1 ⎛⎜ T1T2 e = A2 (t ) T2 − T1 ⎜ ⎝
T2 −T1
t ln 2
⎞ − 1⎟ ⎟ ⎠
(3.4.24)
The daughter activity passes a maximum, like in the cases 3.4.1.6 and 3.4.1.7. The corresponding time tm can be calculated from equ. (3.4.15), too. Example: 135I → 135Xe → 135Cs (T1=6.61 h, T2=9.1 h) 3.4.1.8 Successive activities in longer decay chains (radioactive families)
In case of further decay of a daughter nuclide (decay chains, radioactive families) the activity A3 at time t of the next member in the chain is given by equ. (3.4.25) provided that at time t = 0 only the mother activity existed:
Landolt-Börnstein New Series VIII/4
Ref. p. 3-39]
3 Physical fundamentals
t ln 2 t ln 2 ⎛ − − − T3 T2 T1 e e T2 + e T1 + A3 (t ) = A1 (0)T1 ⎜ ⎜ (T1 − T2 )(T1 − T3 ) (T2 − T1 )(T2 − T3 ) (T3 − T1 )(T3 − T2 ) ⎝
3-15 t ln 2 T3
⎞ ⎟ ⎟ ⎠
(3.4.25)
Example: 211Pb → 211Bi → 207Tl → 207Pb (T1=36.1 m, T2=2.17 m; T3=4.77 m) For T1 » T2 and T1 » T3 and for times t « T1 equ. (3.4.25) is simplified to: ⎛ − T2 e A3 (t ) = A1 (0)⎜1 − ⎜ T2 − T3 ⎝
t ln 2 T2
+
− T3 e T2 − T3
t ln 2 T3
⎞ ⎟ ⎟ ⎠
(3.4.26)
For T2 > T1 and T2 > T3 and for times t » T1 and t » T3 equ. (3.4.25) is further simplified to: − T1T2 e A3 (t ) = A1 (0) (T2 − T1 )(T2 − T3 )
tln 2 T2
(3.4.27)
and for T1 < T2 and T3 « T2 to: A3 (t ) = A1 (0)
− T1 e T2 − T1
t ln 2 T2
(3.4.28)
= A2 (t )
The general equation for the activity of the nth member of a radioactive decay chain was given by Bateman [10Bat] for the case that at t = 0 there are no decay products of the mother nuclide: ⎛ − A n (t ) = A1 (0)T1 ⎜ C1e ⎜ ⎝
t ln 2 T1
+ C2e
−
t ln 2 T2
+ ...... + C n e
−
t ln 2 Tn
⎞ ⎟ ⎟ ⎠
(3.4.29)
with Cn =
(n −2 )
(n − 2 )
(n − 2 ) Tn T1 T2 , C2 = , C1 = (Tn − T1 )(Tn − T2 )...(Tn − Tn −1 ) (T2 − T1 )(T2 − T3 )...(T2 − Tn ) (T1 − T2 )(T1 − T3 )...(T1 − Tn )
3.4.2 Decay modes In order to minimise their total energy, many nuclides undergo radioactive decay. Depending on the lowest reachable energy state various decay types occur either caused by the strong interacting nuclear force (non-isobaric decay with change of the mass number) or by the weak interaction (isobaric decay with constant mass number). The most common radioactive decay modes observed for naturally occurring nuclides are α-decay (non-isobaric), β-decay (isobaric) in the 3 types β−-, β+-decay, and electron capture, and γ-decay (electromagnetic transitions between nuclear states). Much less abundant decay modes are the spontaneous fission and the emission of nucleons (protons, neutrons) and nucleon clusters. The following scheme (Fig. 3.3) shows the nuclear transmutations (changes of atomic and mass number) originating from the main decay modes. A more complete compilation is given in Table 3.3.
Landolt-Börnstein New Series VIII/4
3-16
3 Physical fundamentals
Z, A α
β+
Z−2 , A − 4
Z−1 , A
Z, A
Z, A
Z, A
EC
β−
γ
Z+1 , A
Z, A
[Ref. p. 3-39
Fig. 3.3. Transmutations of atomic and mass numbers during the main nuclear decay modes.
During the α-decay 4He2+ ions are emitted by the unstable radionuclide, whose mass and atomic numbers decrease consequently by 4 and 2 units, respectively. The transitions take place between well defined states in the mother and daughter nuclides leading to one or several monoenergetic α-lines for each α-decaying nuclide. The energy available for the α-decay can be obtained from the Einstein relation and the average binding energy per nucleon (Fig. 3.2) resulting in the fact that all nuclides with A > 140 should be unstable against α-decay. However, further effects have to be considered. In order to leave the nucleus, the α-particle has to surmount a high potential barrier, which can be passed at high energies only. Nevertheless, α-particles are observed with energies well below this barrier. This is a consequence of the quantum mechanical tunnel effect according to which a certain probability exists to pass this barrier at lower energies. Since the width of this barrier decreases with increasing potential, one could expect that α-particles with higher energy tunnelling at a narrower barrier width would pass the barrier with higher probability. This was observed phenomenologically already in 1911 by Geiger and Nutall who found a clear correlation between half-life and range of α-particles, the latter being a direct measure for the αenergy: log λ = a·log Eα + b (Geiger-Nutall rule). Table 3.2. Compilation of nuclear decay modes Decay mode Symbol Radiation emitted helium nuclei He2+ α-decay α
Decay process (Z, A)→(Z−2, A−4) + 4He2+
β−
electrons e−, antineutrinos
(Z, A)→(Z+1, A) + e− + ν ¯e
β+
positrons e+, neutrinos
(Z, A)→(Z−1, A) + e+ + νe
electron capture (EC)
ε
γ-transition
γ
isomeric transition (IT)
Iγ
internal conversion (IC)
e−
proton decay spontaneous fission
p sf
characteristic X-rays/ Auger (Z, A)→(Z−1, A) + νe electrons of the daughter nuclide, neutrinos release of nuclear excitation photons (hν) energy after particle emission delayed release of nuclear photons (hν) excitation energy conversion electrons and transfer of nuclear excitation energy characteristic X-rays to an inner shell (K-,L-,...) electron protons (Z, A) → (Z−1, A−1) + p fission products, neutrons (Z, A) → (Z', A') + (Z−Z', A−A'−x) +xn
β-decay
The denotation β-decay combines all nuclear decay modes, by which the atomic number Z is changed by one unit, while the mass number A remains unchanged (cf. Fig. 3.3). During β−-decay a neutron in the nucleus is transformed into a proton, a negative electron together with an (anti-)neutrino is emitted and the atomic number changes from Z to Z+1. In contrast, during β+-decay a proton in the nucleus is transformed into a neutron, a positive positron together with a neutrino is emitted and the atomic number changes from Z to Z−1. From the mass balance of this process it follows that it can take place only, if the atomic masses of mother and daughter nuclide differ by at least two electron rest mass units. The neutrinos emitted together with the β-particles are uncharged nearly mass-less particles which give rise to the continuous form of the β-spectra, since the available transition energy is shared between the β±-particle and the usually not observed neutrino. The third possibility competing to the β+-decay, the electron capture, results from the fact that the electron orbits of the inner shells - especially the K-shell Landolt-Börnstein New Series VIII/4
Ref. p. 3-39]
3 Physical fundamentals
3-17
can pass the nucleus and the electron can be captured by an excess proton forming a neutron. The hole in the electron shell is filled by an electron from outer shells giving rise to the emission of characteristic X-rays of the daughter nuclide, which are the only external radiation of this process. 5+ 60
5.272y
0
Co Eβ = 315 keV 4+
2505 E γ = 1173.24 keV
2+
1332
E γ = 1332.50 keV 0+
0 60
Fig. 3.4. Main part of the decay scheme of 60Co decaying by β−-decay to 60Ni. On the right hand side of the states their energy referred to the ground state is given (in keV), on the left hand side the nuclear spin and parity characterising the respective level.
Ni
Quite often the radioactive decay modes as discussed above do not populate the ground states but rather excited states of the daughter nuclides (cf. Fig. 3.4). These are de-excited mostly immediately by one or more γ-transitions, until the ground state is reached. The energy of the γ-rays corresponds to the energy difference of the nuclear levels involved in the respective transition. Besides of γ-ray emission an excited nucleus can interact directly with a bound shell electron by transfer of its energy leading to emission of the electron. This process - called internal conversion - is a one-step process without the production of an intermediate γ-ray. It competes with the normal γ-emission, preferentially for heavy nuclei, and obeys the same selection rules like γ-decay. The emission of an electron of the K-shell is called K-conversion and analogous for the other shells L-, M-, N-conversion. Spontaneous fission is a decay mode by which heavy nuclides with high neutron excess separate under energy release into two lighter fragments and several neutrons. Among natural radioactive elements it occurs only in very few cases for primordial or very long-lived nuclides heavier than uranium. Its probability compared to the competing α- or β-decay is very small for uranium (238U: ~1:106), but increases markedly with increasing atomic number Z and neutron excess. In addition, several short-lived isomers decaying exclusively by spontaneous fission (fission isomers) have been observed for the transuranium elements Np to Bk. Spontaneous fission is described by the transformation: (Z, A)!(Z', A') + (Z−Z', A−A'−x) + xn + ∆E, where (Z', A') and (Z−Z', A−A') denote the two fission fragments, x is the number of emitted neutrons and ∆E the energy released mainly in form of kinetic energy of the excited fragments and γ-radiation from their de-excitation. More details on fission products and further physical background of fission will be given in Chapter 3.6. A further decay mode occurring, however, very rarely and mostly for nuclides far off the valley of β-stability is the emission of protons, neutrons or heavier nucleon clusters (5-7Li, 7-9Be, 11-14C, 14-16N, 19-22F and 20-25Ne). With increasing proton excess on the left-hand side of the valley of β-stability, the binding energy of the last proton decreases markedly, and proton emission from the ground state (p-decay) becomes energetically possible by tunnelling through the energy barrier of the nuclear potential as in the case of α-decay. The emission of monoenergetic protons of 1.06 and 1.23 MeV, respectively, by transmutation of the ground state of the mother nuclide into the ground state of the daughter nuclide was first observed for 147 Tm→146Er + p (Ep = 1.051 MeV, t1/2 = 0.56 s) and 151Lu→150Yb + p (Ep = 1.233 MeV, t1/2 = 85 ms). Due to the competition with β+-decay, which is favoured in most cases, p-decay from the ground state Landolt-Börnstein New Series VIII/4
3-18
3 Physical fundamentals
[Ref. p. 3-39
occurs very rarely. Much more frequently, p-emission occurs after β+-decay feeding an excited state in the daughter nuclide, from which the proton can easier surmount the energy barrier (β+-delayed proton emission). β+-delayed emission of protons, in some cases even of two protons or an α-particle, was found for the lightest known isotopes of most of the elements between B and Zr. The binding energy of additional neutrons is higher than that of additional protons and approaches zero only at large distances from the line of β-stability. Therefore, on the right-hand (neutron-rich) side of the line of β-stability all known nuclides are energetically stable to neutron emission from the ground state, which has not been observed up to now. In contrast, neutron emission immediately following β−-decay (β−-delayed neutron emission) is observed for many neutron-rich nuclides and many fission products. In very recent years spontaneous emission of particles heavier than α-particles (cluster-emission) has been observed in several cases. Spontaneous fragmentation of nuclei with atomic numbers Z >40 by emission of cluster nuclei is energetically possible with extremely large partial half-lives. Consequently, cluster radioactivity is a very rare event when compared to other decay modes.
3.4.3 The natural radioactive decay families After the first detection of radioactivity in 1896 many other radioactive substances were identified in the investigated uranium and thorium minerals. Among these, there are three primordial isotopes: 232Th, 238U and 235U, which initiate three naturally occurring decay series. They were called the thorium, uraniumradium and actinium families, the latter two according to their most important members 226Ra and 227Ac, respectively. In these three decay series, only α- and β−-decay occurs. By the emission of an α-particle the mass number decreases by 4 and the atomic number by 2 units, whereas by the emission of a β−-particle the mass number remains unchanged and the atomic number increases by 1 unit. From this fact it follows that all members of such a decay series have mass numbers differing from each other only by multiples of 4 units. This means that all mass numbers occurring in a series have mass numbers A = 4n + b, where n varies within the series and b depends only on the mass number of the starting primordial isotope. Consequently, the thorium family originating from 232Th (A = 232 = 4n with n = 58) is characterised by the common label A = 4n; by variation of n, all possible mass numbers of the members of the decay series are obtained. For the uranium-radium family starting with 238U, the respective label is A = 4n + 2, and for the actinium family starting with 235U, A = 4n + 3. One radioactive decay series with A = 4n + 1 is obviously missing in nature. Members of this family can be produced, however, artificially via nuclear reactions. The longest half-life in this decay series belongs to 237Np; therefore it is called the neptunium family. It was probably present in natural matter after the genesis of the elements before about 5 × 109 y but decayed due to the relatively short half-life of 237Np. The final members of all these decay series are of course stable nuclides: 208Pb at the end of the thorium family, 206Pb for the uranium-radium family, 207Pb for the actinium family, and 209Bi for the neptunium family. Furthermore, in all four decay series one or more branchings occur due to the capability of several nuclides to decay by α-decay as well as by β−-decay. In nearly all cases, however, one branch preponderates strongly and the weak branch ends up at the same common stable end product of the family. In the early years of radioactivity only few radioactive elements and no concept of isotopy had been established. Consequently, the newly detected activities, mainly distinguished from their differing halflives, were given historical names, like mesothorium, actinouranium, thoron, ..., according to the main family members chemically identified at that time. These denominations are frequently found in older literature. In Table 3.4 the abbreviations of these historical names are assigned to the respective isotopes.
Landolt-Börnstein New Series VIII/4
Ref. p. 3-39]
3 Physical fundamentals
3-19
In the four diagrams in Fig. 3.5 all members of the 4 families are given together with their half-lives and decay modes (arranged in a compressed form of the Segrè type of charts of nuclides). For branching decays the main decay branch is given first and indicated by a thick arrow pointing to the daughter nuclide. A thin arrow marks a second weak branch (mostly < 1%) and is also used for subsequent decays in these branches. Table 3.3. Historical names (abbreviations) of the members of the three natural radioactive families Thorium family A = 4n 232
Th Th
228
228 Ra Ac 228Th MsTh1 MsTh2 RdTh
224
Ra ThX
220
Rn Tn
216
Po ThA
212
Pb ThB
212
Bi ThC
212
Po ThC'
208
Tl ThC''
208
Pb ThD
227
Th RdAc 207 Tl AcC''
223
Fr AcK 207 Pb AcD
223
Ra AcX
219
219
Rn An
215
215
Po AcA
211
Pb AcB
234
230
226
222
218
214
218
218
Actinium family A = 4n + 3 235
U AcU 215 At —
231
Th UY 211 Bi AcC
231
Pa — 211 Po AcC'
227
Ac — 211m Po —
At —
Bi —
Uranium-radium family A = 4n + 2 238
U UI 214 Bi RaC
234
Th UX1 214 Po RaC'
Landolt-Börnstein New Series VIII/4
234m
Pa
UX2 Tl RaC''
210
234
Pa UZ 210 Pb RaD
U UII 206 Hg —
Th
Io 210
Bi RaE
Ra — 206 Tl RaE''
Rn — 210 Po RaF
Po RaA 206 Pb RaG
Pb RaB
At —
Rn —
3-20
3 Physical fundamentals
[Ref. p. 3-39
Fig. 3.5a-d (following). Decay chains of the 4 natural radioactive families in a compressed Segrè type representation (N−Z over Z). In this representation α-decay leads vertical to the second field below the mother nuclide, while β− decay leads diagonal to the left field above. All chain members are given together with their half-lives and decay modes. For branching decays the main decay branch is given first and indicated by a thick arrow pointing to the daughter nuclide. A thin arrow marks a second mostly very weak branch and is also used for subsequent decays.
A = 4n
N-Z
44
46
48
50
52
Z 232
Th; α 1.405.1010 y
228
Th; α 1.913 y
90
228
Ac; β6.13 h
89 224
Ra; α 3.66 d
88
228
Ra; β5.75 y
87 220
Rn; α 55.6 s
86 85 84
212
Po; α 0.15 s
212
Bi; α, β60.60 m
83 82
216
Po; α 300 ns
208
212
Pb; β10.64 h
Pb stable
81
208
Tl; β3.053 m
(a) Thorium family
Landolt-Börnstein New Series VIII/4
Ref. p. 3-39]
3 Physical fundamentals
3-21
A = 4n + 1
N-Z
43
45
47
49
51
53
Z 241
Am; α 432.2 y
95
[238U(α,n) ⇒] 94
241
Pu; β-, α 14.35 y
237
Np; α 2.144.106 y
93 233
U; α 1.592.105 y
92
Pa; β27.0 d
229
Th; α 7880 y
90
233
Th; β22.3 m
Ac; α 10.0 d 225
Ra; β14.8 d
88 221
Fr; α 4.9 m
87 217
Rn; α 0.54 ms
86
217
At; α, β32.3 ms
85 213
Po; α 4.2 µs
84
81
[(232Th+n) ⇒]
225
89
82
U; β6.75 d
233
91
83
237
209
213
Bi; β-, α 45.59 m
Bi stable 209
Pb; β3.253 h 209
Tl; β2.16 m
(b) Neptunium family This decay chain is not directly initiated by a primordial nuclide but rather by nuclear reactions of the naturally occurring nuclides 238U and 232Th, respectively, with α-particles and fission neutrons from their decay as indicated in the square brackets. Landolt-Börnstein New Series VIII/4
3-22
3 Physical fundamentals
[Ref. p. 3-39
A = 4n + 2
N-Z
42
44
46
48
50
52
54
Z 234
238
U; α 2.455.105 y
92
U; α 4.468.109 y 234g/m
Pa; β- 6.7 h/1.17 m
91 230
Th; α 7.54.104 y
90
234
Th; β24.10 d
89 226
Ra; α 1600 y
88 87
222
Rn; α 3.825 d
86 218
At; α ≈2 s
85 84
210
81 80
218
Po; α, β3.05 m
Po; α 164 µs
210
Bi; β-,α 5.013 d
83 82
214
Po; α 138.38 d
206
214
Bi; β-,α 19.9 m
210
Pb; β-,α 22.3 y
Pb stable 206
Tl; β4.2 m
214
Pb; β26.8 m
210
Tl; β1.3 m
206
Hg; β8.15 m
(c) Uranium-Radium family In the Uranium-Radium family the decay of 234Th results in the population of the isomeric state (m) as well as the groundstate (g) of 234Pa which in turn decay to 234U.
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A = 4n + 3
N-Z
43
45
47
49
51
Z 235
U; α 7.038.108 y
92 231
Pa; α 3.276.104 y
91 227
Ac; β-, α 21.773 y
223
Ra; α 11.43 d
88
223
Fr; β-, α 21.8 m
87 219
Rn; α 3.96 s
86 215
211
Po; α, β1.78 ms
211
Bi; α, β2.17 m
207
(d) Actinium family
215
Bi; β7.6 m
211
Pb; β36.1 m
Pb stable
81
At; α, β0.9 m
215
Po; α 0.516 s
83
Landolt-Börnstein New Series VIII/4
219
At; α 0.1 ms
85
82
Th; β25.5 h
227
89
84
231
Th; α 18.72 d
90
207
Tl; β4.77 m
3-24
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3.5 Radioactive radiation 3.5.1 Types of radiation For the identification and investigation of radionuclides as well as for radiation protection the properties and interactions of nuclear radiation have to be known. The most important aspect is the interaction of radiation with matter, which determines the detection methods and the actions on inorganic and organic matter. Charged particles or photons, such as α-particles, protons, electrons, positrons, γ- or X-rays, induce ionisation processes in gases, liquids or solids and are called, therefore, ionising radiation. In addition to the production of excited atoms or molecules also chemical reactions may be induced. The minimum energy needed for ionisation or excitation of atoms or molecules is of the order of several eV and depends on the type of the atoms or molecules involved. The photons of visible light and neighbouring wavelenghts have energies between about 1 eV (wavelength λ = 1240 nm) and 10 eV (λ = 124 nm). If their energy exceeds the ionisation energy of the absorbing matter, they lose it in a single ionisation process. On the other hand, charged particles have energies in the range of 0.1 to 10 MeV and produce a large number of ions, electrons and excited atoms or molecules in many interaction steps. 3.5.1.1 Particle radiation
This type of radiation comprises the frequently and also naturally occurring α- and β-radiation, the neutron radiation accompanying mainly nuclear fission, and the quite rare types of nucleon- and clusteremission. The main properties of these particles influencing their physical behaviour like interaction with and absorption in matter are their mass, charge and energy. Ions and excited atoms or molecules produced in the primary interactions of charged particles with matter give rise to further (secondary) physical and chemical reactions. Many of these secondary reactions are very fast and happen relatively frequently. The concentration of such reaction products along the path of the primary particle is proportional to its energy loss per unit path length. For example, α-particles of 1 MeV loose 190 eV/nm in water, whereas 1 MeV electrons loose only 0.2 eV/nm, resulting in shorter ranges for the α-particles but higher concentrations of reaction products, respectively. Very high energetic particles may also induce nuclear reactions. Electrons loose in the force field of nuclei part of their energy by generating bremsstrahlung. For electron energies of the order of 1 MeV (e.g. β-radiation), these photons have energies in the X-ray range and for energies >10 MeV the bremsstrahlung photons have the energies of γ-rays. Since neutrons carry no charge, they can interact with matter via the very short-range nuclear force, and they loose their energy stepwise mainly by collisions with atomic nuclei or they may induce nuclear reactions. This requires a large number of scattering processes and especially light materials to slow down and absorb neutrons. From range calculations it follows that α-radiation is easily absorbed quantitatively (e.g. by a sheet of paper), for absorption of β-radiation several millimetres or centimetres of material are necessary, and for absorption of γ-radiation thick layers of either lead or concrete are needed taking into account that the absorption of γ-rays follows an exponential law. The ratio of the absorption coefficients for α-, β- and γradiation of equal energy is about 104:102:1.
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3.5.1.2 Electromagnetic radiation
The types of electromagnetic radiation considered here are mostly produced during de-excitation processes in the electron shells or the nuclei of atoms. Due to historical reasons, this electromagnetic radiation − despite of similar fundamental properties − is differently denominated according to different origins: • • • •
emission from de-exciting outer electron shells: emission from de-exciting inner electron shells: emission from de-exciting nuclei: emission during deceleration of charged particles:
light X-rays γ-radiation bremsstrahlung.
The energy range of X-rays lies within about 100 eV to 100 keV (= wavelengths of about 10 nm to 10 pm), and that of γ-rays within about 10 keV to 104 MeV (= wavelengths of about 0.1 nm to 10−7 nm). That means there is an overlap in the energy ranges of X-rays and γ-rays. Electrons with energies >10 MeV decelerated in a substance of high atomic number induce the emission of high-energetic (“hard”) bremsstrahlung. In contrast to the γ-rays emitted from nuclei, this bremsstrahlung shows a continuous energy distribution. X-rays are emitted from the electron shell of the nuclides after formation of a hole in one of the inner shells either by an external process (bombardment with photons, electrons or heavy charged particles) or by an internal decay process like electron capture or internal conversion. The X-ray spectrum of an individual nuclide shows a distinct line structure corresponding to the different transition possibilities in the electronic shell and reflecting its structure. The γ-rays emitted by an excited nucleus have well-defined energies, which correspond practically to the differences in the excitation energies of the nuclei (the recoil energies transferred to the emitting nuclei are very small). Gamma spectroscopy is, therefore, the usual method to investigate the level structure and decay schemes of atomic nuclei. Generally, the γ-radiation is emitted immediately after a preceding α- or β-decay, since the lifetimes of excited states are of the order of 10−13 s. However, if immediate γ-transitions are “forbidden” because of high differences of the nuclear spins of the involved states and the ground state in combination with the conservation laws of nuclear momentum and parity (selection rules), a metastable or isomeric state results which decays with its own half-life different from that of the mother nuclide. The transition from the isomeric to the ground state is called isomeric transition (IT). Some long-lived nuclear isomers in isotopes with a stable ground state are of practical importance as pure γ-emitters. 3.5.1.3 Conversion electrons
If a γ-transition in an excited nucleus is hindered by the selection rules, a certain probability exists increasing with increasing “hindrance” to transfer the excitation energy directly to a shell electron, which is emitted monoenergetically. The energy Ee of these conversion electrons is given by Ee = Eγ − EB where Eγ is the respective γ-energy, and EB the binding energy of the electron in the respective shell. Since the higher shells L, M, ... have sub-shells with slightly different binding energies, the conversion electron spectra contain 3 lines for L-conversion and 5 lines for M-conversion. The internal conversion leaves a hole in the electron shell, which by recombination causes the emission of characteristic X-rays or Auger electrons. The latter − in contrast to conversion electrons − gain their energy from transitions between electron shells, which is in general much lower than the nuclear transition energies of conversion electrons. The relative abundance of internal conversion is given by the total conversion coefficient α = Ne/Nγ, which is composed from the partial conversion coefficients for the individual electron shells and subshells: α = αΚ + αLΙ + αLII + αLIII + αM..., each defined as the ratio of the number of the respective conversion electrons to the number of γ-rays emitted.
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3.5.2 Physical properties of radiation In order to investigate radionuclides and to develop protection methods against radioactive radiation, one has to deal with their properties and the interaction processes between the different radiation types and matter. In principle, the absorption of electromagnetic radiation like γ- and X-rays is different from that of particles. While the latter lose their energy by successive collisions, photons give off their energy mostly in one process. Because they are chargeless, their interaction with matter is small. The absorption of γrays follows an exponential law: I = I0 e−µd
(3.5.1)
where µ is the absorption coefficient and d the absorber thickness. The exact validity of this exponential law is, however, restricted to monoenergetic γ-radiation, a narrow beam of γ-rays and a thin absorber. 3.5.2.1 Interaction of charged particles with matter
The primary interactions of fast charged particles with matter can be classified as follows: a) b) c) d) e)
elastic collisions with atomic electrons inelastic collisions with atomic electrons elastic collisions with nuclei inelastic collisions with nuclei nuclear reactions and interaction with nuclear forces
All these interactions can contribute in principle to deflection and deceleration of incoming particles. Since more than 104 collisions are necessary to stop particles with energies in the MeV range, this is a multi-scattering process composed statistically in different ways of the individual processes. Heavy charged particles (p, d, α, ions) loose most of their kinetic energy by process (b). The electrons of the scattering atom are excited by the energy transfer, most of them to energies high enough to ionise the atoms. Elastic collisions of heavy particles (protons, deuterons, α-particles, ...) with nuclei (c) are rare in comparison to (b), for light particles (electrons, positrons), however, much more frequent. Inelastic scattering from nuclei (d) also occurs rarely and is negligible for deceleration of heavy particles, for light particles, however, it is important at higher energies giving rise to the emission of bremsstrahlung. Consequently, there are characteristic differences for the deceleration of heavy and light particles. On the other hand, the detection of all types of charged particles, as well as of electromagnetic radiation, is based in general on their ionisation effects. This holds also for the damaging effects in biological material. Deceleration of heavy particles
By the inelastic scattering of fast particles from shell electrons the atoms are excited or preferentially ionised. For the resulting loss of kinetic energy per unit path one obtains: −
dE 4π z 2 e 4 = N0B ds me v 2
(3.5.2)
In this expression − also called stopping power − z means the charge of the fast particle, v its velocity and N0 the number of nuclei per cm3 in the absorber material. Landolt-Börnstein New Series VIII/4
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The atomic stopping number B is a function, depending on the energy of the incoming particle. From a quantum mechanical calculation H. Bethe obtained the following expression for the atomic stopping power B under consideration of relativistic effects at high particle energies: ⎡ 2m v 2 C ⎤ B = Z ⎢ln e − ln(1 − β 2 ) − β 2 − K ⎥ I Z ⎦ ⎣
(3.5.3)
where Z means the atomic number of the absorber, β = v/c and I = 11.5·Z the average ionisation potential; CK is a correction factor depending on E and Z, which is of influence only at very low projectile energies and takes values in between 0 and ~1. For the stopping power it follows: −
⎡ 2m v 2 C ⎤ dE 4π z 2 e 4 = N 0 Z ⎢ln e − ln (1 − β 2 ) − β 2 − K ⎥ 2 I Z ⎦ ds me v ⎣
(3.5.4)
The schematic course of equ. (3.5.4) shown in Fig. 3.6 demonstrates qualitatively that in the intermediate region for energies much larger than the ionisation potential and much less than the particle rest mass (I « E « Mc2) the energy loss decreases approximately as 1/E, since the logarithmic term varies slowly and the relativistic terms are negligible. At high velocities, however, these terms give rise to a slight increase causing a flat minimum at E ≈ 3Mc2. At low energies the logarithmic term prevails and the curve decreases steeply below E ≈ 500·I. From this behaviour it follows that the ionisation density produced by the decelerated particle in the absorber increases markedly before the end of the path s and decreases steeply behind this point. Furthermore, the 1/v2 dependence of the first term in equ. (3.5.4) causes a stretching of the energy scale for different masses of the incoming projectiles (cf. Fig. 3.6).
Loss of kinetic energy per unit path - d E ds
Loss of kinetic energy per unit path - d E [ MeV/ cm ] dx
0.05
~ = 500 l
3 Mc 2 Particle energy E
0.04
µ
d
α
0.03 e
_
p
0.02
0.01
0 10 -2
10 -1
1 10 2 10 Particle energy E [MeV]
10 3
10 4
Fig. 3.6. Schematic course of the stopping power with energy (left) and dependence on the projectile type for electrons, muons, protons, deuterons and α-particles (right).
3.5.2.2 Interaction of neutrons with matter
Neutrons are emitted by radioactive decay during spontaneous fission and in rare cases after β-decay of a very neutron-rich nuclide to an excited level of the daughter nuclide. In nuclear reactions and in particular in nuclear fission neutrons are of high importance, especially for the production of radionuclides. Since from radioactive sources mostly only small fluxes can be obtained, neutron sources based on nuclear reactions, like 9Be(α,n)12C, 9Be(d,n)10B or 9Be(γ,n)2α, are used. The highest fluxes of neutrons are available in nuclear reactors. Landolt-Börnstein New Series VIII/4
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The interaction of neutrons with matter happens mainly with nucleons and nuclei via elastic and inelastic scattering and nuclear reactions, because they are electrically neutral and interact hardly with electrons. Thus, ionisation by neutrons is negligible. In elastic collisions the total kinetic energy remains constant and only deflection occurs, while in inelastic collisions part of the kinetic energy is used for excitation of the collision partner. Depending on the neutron energy different types of interactions occur and, correspondingly, several energy ranges are distinguished: • • • • •
0-0.1 eV: 0.1-100 eV: 1-10 keV 0.1-100 keV: 0.1-10 MeV:
thermal neutrons (energies similar to those of gas molecules at room temperature) slow neutrons epithermal neutrons neutrons of intermediate energies fast neutrons
In contrast to charged particles, neutrons do not undergo Coulomb interaction with nuclei. However, low-energy (thermal and slow) neutrons are very effectively absorbed by a great number of nuclei, giving rise to nuclear reactions. Elements such as B, Cd, Sm, Eu, Gd and Dy are used as good neutron absorbers. Epithermal neutrons are also called resonance neutrons, because there exist absorption maxima or resonances at defined energies for distinct absorber nuclei. Neutrons with energies corresponding exactly to the excitation energies of excited levels of the nucleus are absorbed with very high probability. Fast neutrons are decelerated mainly by elastic and inelastic collisions. The energy released in one elastic collision depends on the collision angle and the mass number of the target nucleus. The lighter the nucleus, the higher is the energy loss of the neutron. Consequently, hydrogen or hydrogen-containing substances like water or paraffin are very effective to reduce (“moderate”) the energy of neutrons. Graphite can also be used as moderating material, but needs larger material thickness due to its higher mass number. After deceleration the “slow neutrons” are captured by nuclei, giving rise to nuclear reactions. High-energy (fast) neutrons may also induce nuclear reactions, but the probability and consequently the contribution of this interaction type is relatively small. 3.5.2.3 Interaction of electromagnetic radiation with matter
Electromagnetic radiation passing a material experiences an intensity attenuation according to equ. (3.5.1) (see above) via interactions with the various components of matter. For instance, γ-rays interact with: • • • •
atomic electrons, nuclei, electrical fields of the electrons and nuclei, meson fields of the nuclei.
These interactions can result in energy losses, alterations of the propagation direction and polarisation. The effects on γ-rays are: • total absorption, • inelastic scattering (incoherent), • elastic scattering (coherent, Thomson scattering). Though all of the possible combinations of interactions and processes can occur, most of them result in very weak effects, which can normally be neglected for attenuation considerations. Important are: • the photoeffect, • the Compton effect, • the pair formation. In the photoeffect the incoming photon is totally absorbed by an atomic electron of the inner shells (preferably of the K- or L-shells). A free electron cannot take up the total energy of the photon because of the momentum conservation. In contrast, this is possible for the bound atomic electrons, because the atom Landolt-Börnstein New Series VIII/4
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3 Physical fundamentals
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takes up the residual momentum as recoil. Consequently, the stronger bound inner electrons exhibit the strongest absorption. The photoelectrons in turn are emitted with a kinetic energy Ekin corresponding to the total photon energy h·ν reduced by the electron's binding energy Eb, which depends on the atomic number of the respective atom: Ekin = h·ν − Eb
(3.5.5)
Consequently, for high photon energies and light absorbers the photoelectrons carry nearly the total photon energy. In addition, they exhibit a pronounced angular distribution favouring the forward direction with increasing photon energy. Additional (isotropic) radiation is emitted originating from the recombination process of the electron hole left by the emitted photoelectron. During this recombination characteristic X-rays and Auger electrons are released by the energy set free by an outer electron filling up the hole in the inner shell. Furthermore, bremsstrahlung is produced by the photoelectrons especially in heavy materials. All these radiation types connected with the absorption of photons together with the ionised residual absorber atoms can result in a high local energy density. The Compton effect describes the inelastic scattering of photons from outer shell electrons, where the photon looses only part of its energy, which is transferred to the electron. Both, the recoiling electron and the photon are scattered with respect to the original photon direction, and the scattering into forward directions is again favoured with increasing photon energy. The relative energy transfer to the electron increases with increasing photon energy and also with increasing photon scattering angle. On the other hand, the highest backscattering contributions occur at low photon energies. The pair formation occurs in the electric field of an atomic nucleus, where the γ-ray can be spontaneously transformed into an electron-positron pair, if its energy is larger than the sum of the rest masses of the pair: Eγ > 2mec2 = 1,022 MeV. The nucleus remains unchanged but is necessary for the conservation of momentum. The positron annihilates by recombination with an electron of the absorber atoms, whereby two annihilation γ-quanta are produced with an energy of 511 keV each.
Compton effect
Total
Pair formation
Absorption coefficient m [cm -1]
Absorption coefficient m
Photoeffect 1.5
1.0 Pb
0.5 Cu Fe
10 -1
1 10 γ - energy E γ [MeV ]
10 2
Fig. 3.7. Schematic representation of the partial and total absorption coefficients of a heavy absorber in dependence on the γ-ray energy.
0 0.1
Al 1 γ - energy E γ [MeV ]
10
Fig. 3.8. Schematic representation of the total absorption coefficients of different absorbers in dependence on the γ-ray energy.
For the attenuation of photons in matter, mainly the three processes photoeffect, Compton effect and pair formation have to be taken into account. Fig. 3.7 shows schematically their contributions to the total absorption coefficient µ in dependence on the photon energy. A primary photon beam is attenuated by scattering as well as by absorption. The main parameters for the attenuation are the photon energy and the Landolt-Börnstein New Series VIII/4
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atomic number and density of the absorber material (cf. Fig. 3.8). At low photon energies, classical coherent (Thomson) scattering predominates, at higher energies incoherent scattering via the Compton effect. Absorption of photon energy occurs mainly via the photoeffect and pair formation and partially via the Compton effect. The lower energy threshold for the photoeffect is given by the binding energy of the inner shell electrons, for the pair formation by the rest mass of the electron-positron pair (1022 keV). At high atomic numbers Z and low photon energies the photoeffect dominates, at high energies the pair formation. For energies between about 1 to 5 MeV the Compton effect dominates at all atomic numbers Z. The resulting Compton electrons give, therefore, the most important contribution to the energy dose in human tissue and to the biological effect of radiation.
3.6 Nuclear fission and fission products 3.6.1 Particle induced nuclear fission Besides the few heavy nuclides decaying by spontaneous fission many other heavy isotopes undergo fission after bombardment with particles, especially neutrons. Fission of uranium was first observed but misinterpreted in 1934 by Fermi [34Fer] in an attempt to produce transuranium elements by irradiation of uranium with slow neutrons. Similar experiments were performed by several groups, but only in 1937 Hahn, Meitner and Strassmann [37HMS] identified the observed radioactive products of the fission of uranium to have appreciably lower atomic mass, such as 140Ba. Fission of heavy nuclei always leads to products with a high neutron excess due to the much larger neutron-to-proton ratio of heavy nuclides. The primary fission products are formed in about 10−11 s by fission and emission of prompt and β−-delayed neutrons and γ-rays from the highly excited fragments. They always lie on the right hand side of the valley of β-stability and decay by several successive β−-decays following isobaric chains into nuclides of increasing atomic number Z ending up with the first stable isobar in the chain. The fission process exhibits different features depending mainly on the energy of the inducing particles and on the atomic number Z of the fissioning nuclide. For fission induced by low-energy neutrons with energies up to about 10 MeV two fission products with mass numbers in the range between about 70 and 160 and ≈ 2-3 neutrons are emitted. The energy ∆E released by fission is relatively high (∆E ≈ 200 MeV), since for the light fission products the binding energy per nucleon is higher than for the heavy fissioning nuclei. When comparing the naturally occurring uranium isotopes 234U (abundance: 0.0055 %), 235U (0.720 %), and 238U (99.2745 %), it is found that 235U undergoes fission by slow (i.e. thermal) neutron capture with a remarkably large cross section of σn,f = 586 × 10−24 cm2. This holds also for other odd-mass (even Z, odd N) nuclei like 233U, 235U, 239Pu and 241Pu. In all these cases the binding energy of an additional neutron is very high resulting in high cross sections σn,f for fission by slow (thermal) neutrons. To induce fission in 238U, the neutron must have an energy of 1 MeV. The natural isotope 232Th undergoes fission by 1.1 MeV neutrons. Using higher energetic particles, either neutrons, deuterons, or α-particles, it is possible to induce fission in any element with atomic number larger than 73.
3.6.2 Fission products One of the most characteristic features of the fission process − very important also for the activities that have to be regarded in nuclear technologies − is the resulting mass distribution of the various fission products. As an example, Fig. 3.9 displays the yield of fission products found after thermal fission of 235U Landolt-Börnstein New Series VIII/4
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3 Physical fundamentals
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as a function of the mass number A. The maximum yields are around A = 90 to 100 and A = 133 to 143, respectively. For these mass ranges the fission yields are about 6 %, whereas symmetrical fission around A = 117 occurs with much smaller probability ( ≈0.01 %). It should be reminded that the sum of the fission yields is 200 %, because each fission produces two fission products. When investigating the mass distributions from fission with thermal neutrons one observes that those of 233U and 239Pu are similar to that of 235U. In the case of 239Pu, however, the low mass maximum is slightly shifted to higher masses, while the maximum for heavy fission products remains nearly unchanged. This tendency continues with increasing mass of the fissioning nuclei, and in the case of 258 Fm the two maxima are superimposed. An increase of the energy of the neutrons leads to a strong increase of the probability for symmetric fission and gives rise to a flattening of the valley of the mass distribution by up to two orders of magnitude. Increase of symmetric fission is also observed for nuclides with lower atomic numbers Z. For 227 Ac (Z = 89) symmetric and asymmetric fission have nearly the same probability, resulting in three maxima in the mass distribution. 10
Fission yield [%]
1
10 -1
10 -2
10 -3
10 - 4 80
100
120 Mass number A
140
160
Fig. 3.9. Yields of fission products for the fission of 235 U by thermal neutrons.
It should be mentioned that the mass distribution curves as discussed before give the total yields of the decay chains of mass numbers A. The independent yields for individual members of the decay chains, i.e. the yields for direct formation in the fission process, are often difficult to determine, especially if in the precursor chain short half-lives occur. The total energy ∆E released during fission consists of kinetic energy and excitation energy of the primary fission fragments, where the kinetic energy resulting mainly from the Coulomb repulsion of the two fission fragments gives the main contribution. In the case of low-energy fission the kinetic energy Ekin is given by the empirical relation Ekin ~ Z2 / A1/3 where Z and A are the atomic number and the mass number of the fissioning nucleus, respectively. The primary fission fragments release their excitation energy by emission of prompt neutrons with energies between 0 and about 10 MeV (mean value ≈2 MeV) and of prompt γ-rays. The number of prompt neutrons emitted increases with the mass number of the fissioning nuclei and depends mainly on the excitation energy of the primary fission fragments. Furthermore, an average number of 7.5 γ-rays with a mean energy of about 1 MeV are emitted per fission, as well as several low-energy transitions in form of conversion electrons and X-rays. Landolt-Börnstein New Series VIII/4
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In very rare cases high-energy charged particles such as p, d, t, α-particles, 3He, 7Li, 8Li, 9Li, 9Be, 10Be and isotopes of B, C, N and O are also emitted at an early stage of the low-energy fission process, when the fission fragments are still very close to each other. The probability of this so-called ternary fission, i.e. formation of three fragments increases strongly with the excitation energy of the fissioning nuclei. For example, high-energy fission of 232Th with 400 MeV argon ions leads to a ratio of ternary to binary fission of about 1:30. In contrast to low-energy fission, high-energy fission induced by neutrons or other high-energy particles leads to marked changes in the mass distribution of fission products. Among others, the probability of symmetrical fission increases considerably with increasing excitation energy of the target nuclei, resulting in a single flat maximum of the fission yield curve slightly below half the target mass number.
3.6.3 Nuclear reactors When comparing the energies available from combustion of carbon or carbon compounds (or in general: from a chemical reaction like oxidation) with those from nuclear fission, one has to compare the orders of magnitude of the binding energies in the electron shells and in the atomic nuclei, which are in the order of eV and MeV, respectively. This means a difference of six orders of magnitude: 1 kg of carbon produces an energy of 9.4 kWh, 1 kg of uranium can produce a maximum of 1.85 × 107 kWh. This high energy release as already observed with the first fission experiments caused immediate considerations to make use of this enormous energy potential. In addition, the neutrons released in each fission process could initiate further fissions leading in principle to a continuous generation of energy. However, it soon turned out that a safe control of a continuously proceeding fission process required the solution of various technological questions and problems. Consequently, the first application of nuclear fission was the construction of a nuclear explosive (“atomic bomb”) on the basis of uncontrolled self-amplifying fission in the form of chain reactions. As mentioned above, the fission process in the first step results in two fragments with high neutron excess and in some prompt high energy neutrons. Furthermore, from the decay chain of the highly excited primary fragments besides β−-particles and γ-rays about l % of β-delayed neutrons are emitted which are delayed by at least 0.01 second and about 0.07 % which are delayed by as much as 1 minute. This would be a satisfactory condition for a chain reaction provided the neutrons originating from the fission process have a possibility to react with other atoms of 235U. The neutrons generated during fission are, however, no thermal neutrons but have energies up to the order of 1 MeV. If these neutrons are slowed down to thermal velocities they can excite other 235U atoms to fission, so that the reaction would proceed with increasing amplification as a chain reaction if only 235U would be present. In natural uranium this process is not possible because of the high concentration of 238U atoms which can capture higher energy neutrons without undergoing fission. In addition, there exist well-defined energies, where a resonance absorption (in a very sharp energy window) of neutrons in 238U occurs with very high cross sections (up to 1200 × 10−24 cm2). Consequently, the fast neutrons have to be effectively slowed down, whereby a sufficient number must escape the resonance absorption in 238U in order to reach thermal velocities. This is achieved by use of a moderator, an element of small atomic weight like deuterium, helium, beryllium, and carbon, whose atoms will not capture the neutrons but rather scatter them elastically.
3.6.4 Nuclear explosives The high amount of energy released by nuclear fission led very early to the production of nuclear explosives. Since 235U, 233U and 239Pu have sufficiently high fission cross sections for fast neutrons, they can be used as nuclear explosives if the respective critical masses are brought together. Without a neutron Landolt-Börnstein New Series VIII/4
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reflector increasing the number of neutrons within the explosive material, a sphere of about 50 kg uranium metal containing 94 % 235U or a sphere of about 16 kg plutonium metal (239Pu) reaches criticality. If a neutron reflector is used, the critical masses are about 20 kg for 235U and about 6 kg for 239 Pu. The critical masses for 233U are similar to those for 239Pu. The use of plutonium in nuclear weapons requires a low concentration of 240Pu in the plutonium, because its presence leads to the production of high numbers of neutrons by spontaneous fission. Consequently; a too high concentration of 240Pu would initiate the neutron multiplication too early with a relatively small multiplication factor and a relatively low energy release. Higher concentrations of 241Pu also interfere, because of its decay into 24lAm with a half-life of only 14.4 y. To minimise the formation of 240Pu and 241Pu, Pu for use in weapons is, in general, produced in special reactors with low burn-up (« 20000 MWthd per ton). Criticality can be reached by shooting two under-critical hemispheres onto each other by means of normal explosives (gun-type) or by compressing an under-critical spherical shell into a supercritical sphere (implosion-type). The bomb ignited over Hiroshima (energy release corresponding to ≈15 kilotons of TNT) was of the gun type using 235U, whereas that ignited over Nagasaki was of the implosion type using 239Pu (energy release corresponding to ≈22 kilotons of TNT). Generally, the fissile core is surrounded by a heavy material, in order to reflect the neutrons and to increase the inert mass and consequently the time in which the super-critical configuration is held together. The explosion of fissile material leads to temperatures of about 108 K which are sufficient to initiate fusion between deuterium and tritium. This is the basis of the development of hydrogen bombs, in which the energy of fission is used for ignition of fusion. LiD serves as a source of D and T, the latter being produced by thermal (6Li(n,αn)d) and by fast neutrons (7Li(n,αn)t). If the temperature is high enough, the D-D reaction can contribute to the energy production. The fast neutrons released by the fusion reactions react very effectively with natural or depleted U initiating fission of 238U. By these kinds of weapons large amounts of fission products are formed (“dirty weapons”). If a surrounding of non-fissile heavy material is used, fission products are released only by the ignition process (“clean weapons”).
3.6.5 Radioactive inventory and nuclear waste The radioactive inventory in a nuclear power reactor originates from: • fission products • uranium and transuranium elements formed by direct neutron induced reactions and their decay chains • isotopes produced by nuclear reactions in the cladding material of fuel rods, the reactor vessel components and the coolant. The dominating longer-lived fission products occurring in spent fuel elements are 85Kr, 131I, 133I, Xe, 135Xe, 134Cs and 137Cs. Radionuclides produced by nuclear reactions in the coolant are 3H, 14C, 13N, 16 N, 19O, 18F and 41Ar. Furthermore, fission products or actinides may leak into the cooling system from faulty fuel elements. Other radionuclides are produced by reactions with metals and their corrosion products in various reactor vessel components like 51Cr, 54Mn, 59Fe, 58Co, 60Co, 65Zn, 124Sb. 133
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10 3
Rel.activity
a 10 2 b
Fig. 3.10. Radioactive decay of the sum of all fission products (a), 106Ru (≈0.5 %, T = 373.6 d) (b), and 137Cs (≈ 6 %, T = 30.17 y) (c).
10 c 5 0
100 200 300 400 500 600 700 800 900 Decay time [d ]
Shortly after shut-down of a nuclear reactor the activity of the fuel is ≈1.7 × 1017 Bq per MW of thermal energy produced. 237U (from the reactions 235U(n, γ)236U(n, γ)237U and 238U(n, 2n)237U) causes a relatively high initial uranium activity. Since it decays with a half-life of 6.75 d, it vanishes rapidly after the necessary storage of the discharged fuel elements. A global composition of spent nuclear fuel from light water reactors after storage of 1 year is given in Table 3.5. The radioactive waste produced during the operation of nuclear reactors is usually classified according to the state of matter (gaseous, liquid or solid) and according to the activity level as low-active waste (LAW), medium-active waste (MAW), and high-active waste (HAW). The largest amount of radioactivity is concentrated in the spent fuel elements representing highly-active waste (HAW). Table 3.4. Main components of spent nuclear fuel from a light-water reactor with an initial enrichment of 3.3 % 235U, a burn-up of 34000 MWd per ton and a storage time of 1 year (from [97Lie]).
Nuclide
Weight percent
uranium and transuranium elements 235 U 0.756 236 U 0.458 237 U 3×10−9 238 U 94.2 237 Np 0.05 238 Pu 0.018 239 Pu 0.527 240 Pu 0.220 241 Pu 0.105 242 Pu 0.038 americium isotopes 0.015 curium isotopes 0.007
Nuclide
Weight percent
fission products 85 K 90 Sr 129 I 134 Cs + 137Cs others Stable fission products
0.038 0.028 0.09 0.275 0.19 3.0
Reprocessing of nuclear fuel transforms all waste types into liquid solutions and results in the following amounts per ton of U: ≈1 m3 HAW (fission products and actinides in HNO3 solution), ≈3 m3 MAW as organic solution, ≈17 m3 MAW as aqueous solution, ≈90 m3 LAW (aqueous solution). By further processing a volume reduction is achieved: ; ≈0.1 m3 HAW, ≈0.2 m3 MAW (organic), ≈8 m3 MAW (aqueous), ≈3m3 LAW (aqueous). After respective storage times the HAW solutions are transformed by calcination or vitrification into stable forms like ceramics or glasses suitable for long-term disposal and also in order to reduce the volume. Landolt-Börnstein New Series VIII/4
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After one year of intermediate storage and reprocessing, the initial activity of the HAW solutions is of the order of 1014 Bq/l from which the activity due to 90Sr and 137Cs is about 1013 Bq/l; after 10 y the activity of the HAW solution decreases approximately with the half-life of these nuclides (28.64 y and 30.17 y, respectively, cf. Fig. 3.10). After 1000 y the residual activity (of the order of 104 Bq/l) is determined by long-lived fission products like 99Tc, 129I and actinides. Solid MAW and LAW originate from structure material of the fuel elements, undissolved, dispersed and filtered particles of metals or metal oxides (e.g. Ru, Rh, Mo, Tc), and gaseous components like tritium (as T2 or HTO), 14C (as CO2), 85Kr, 129I or 106Ru (as RuO4) after adsorption in special adsorbents. Computer codes have been developed to calculate the inventory of all radionuclides in spent fuel rods after discharge and after several periods of storage. As an example in the Tables 3.6 and 3.7, taken from [95SSK], the activities of actinides, fission products and light elements from surrounding material are listed, which were obtained from a calculation with the code KORIGEN for a given reactor type, initial 235 U enrichment, and fuel burn-up at charge time, discharge time and several storage times. Further information on these calculations and their results as well as on general reprocessing, recycling and disposal concepts can be found e.g. in [83FiW], [93Wie], [95SSK], [80Clo] and [97Lie]. Table 3.5. Inventory of selected relevant radionuclides in Bq for a pressurized water reactor with a thermal power of 3.733 MW after an operation time of 1 day [95SSK].
Nuclide 85
Time after discharge 1h 6h
0h
24 h
120 h
Kr Kr 87 Kr 88 Kr 133 Xe 135 Xe Σ Kr-Xe
13
4.08×10 1.43×1018 2.86×1018 4.03×1018 2.89×1017 1.74×1018 1.03×1019
13
4.29×10 1.24×1018 1.68×1018 3.16×1018 3.10×1017 2.09×1018 8.47×1018
13
4.96×10 5.72×1017 1.10×1017 9.32×1017 4.04×1017 2.91×1018 4.93×1018
13
5.50×10 3.53×1016 6.02×1012 1.15×1016 6.15×1017 1.72×1018 2.38×1018
5.53×1013 1.25×1010 0.0 7.60×1005 5.82×1017 2.04×1015 5.84×1017
131
I I 133 I 134 I 135 I Σ Iodine
2.36×1017 9.66×1017 4.16×1018 8.76×1018 6.76×1018 2.09×1019
2.43×1017 9.61×1017 4.15×1018 6.53×1018 6.09×1018 1.80×1019
2.46×1017 9.26×1017 3.57×1018 2.63×1017 3.60×1018 8.61×1018
2.38×1017 7.92×1017 1.96×1018 2.23×1011 5.46×1017 3.54×1018
1.80×1017 3.38×1017 8.01×1016 0.0 2.32×1013 5.98×1017
89
7.29×1016 4.30×1014 5.51×1018 1.08×1014 4.11×1016 7.87×1016 4.28×1018 7.68×1014 1.58×1018 9.38×1017 6.80×1016 1.40×1018 1.22×1015 4.10×1017 2.27×1016 7.54×1017
7.42×1016 4.32×1014 5.14×1018 1.12×1014 4.37×1016 7.95×1016 4.11×1018 8.32×1014 l.56×1018 9.86×1017 6.81×1016 1.24×1018 1.22×1015 4.28×1017 2.32×1016 6.53×1017
7.41×1016 4.31×1014 3.57×1018 1.29×1014 5.47×1016 7.93×1016 3.35×1018 1.15×1015 1.48×1018 1.14×1018 6.78×1016 5.67×1017 1.22×1015 4.68×1017 2.36×1016 2.93×1017
7.33×1016 4.31×1014 9.59×1017 1.83×1014 7.28×1016 7.87×1016 1.60×1018 2.29×1015 1.23×1018 1.15×1018 6.69×1016 3.41×1016 1.22×1015 3.81×1017 2.09×1016 1.63×1016
6.94×1016 4.31×1014 8.71×1014 3.44×1014 7.60×1016 7.53×1016 3.12×1016 7.93×1015 4.48×1017 4.32×1017 6.24×1016 1.05×1010 1.21×1015 5.88×1016 1.02×1016 3.32×1009
85rn
132
Sr Sr 91 Sr 90 Y 91 Y 95 Zr 97 Zr 95 Nb 99 Mo 99m Tc 103 Ru 105 Ru 106 Ru 105 Rh 127 Sb 129 Sb 90
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Nuclide 127 Te 127rn Te 129 Te 129m Te 131m Te 132 Te 134 Cs 136 Cs 137 Cs 140 Ba 140 La 141 Ce 143 Ce 144 Ce 143 Pr 239 Np 238 Pu 239 Pu 240 Pu 241 Pu 241 Am 242 Cm 244 Cm Σ Aerosols
0h 1.07×1016 1.07×1013 6.70×1017 1.79×1015 1.72×1017 9.48×1017 3.67×1010 2.99×1014 4.56×1014 3.83×1017 7.20×1016 1.07×1017 2.66×1018 1.52×1016 7.23×1016 1.50×1019 8.38×1006 6.18×1011 5.25×1009 3.60×1009 3.87×1003 2.15×1002 9.62×10−04 3.52×1019
Time after discharge 1h 6h 1.13×1016 1.41×1016 1.16×1013 1.59×1013 17 6.43×10 3.42×1017 15 1.87×10 2.11×1015 17 1.70×10 1.52×1017 17 9.42×10 9.01×1017 10 3.77×10 4.02×1010 14 2.98×10 2.95×1014 14 4.58×10 4.58×1014 3.82×1017 3.78×1017 16 7.73×10 1.02×1017 17 1.13×10 1.30×1017 18 2.65×10 2.39×1018 16 1.52×10 1.52×1016 16 7.78×10 1.04×1017 19 1.51×10 1.43×1019 06 9.42×10 1.44×1007 11 6.67×10 9.09×1011 09 5.51×10 5.79×1009 09 3.60×10 3.60×1009 03 4.53×10 7.82×1003 02 2.67×10 4.95×1002 −03 1.26×10 2.46×10−03 19 3.46×10 3.00×1019
24 h 1.77×1016 3.06×1013 2.08×1016 2.27×1015 1.00×1017 7.68×1017 4.13×1010 2.83×1014 4.58×1014 3.63×1017 1.74×1017 1.40×1017 1.63×1018 1.52×1016 1.74×1017 1.15×1019 2.99×1007 1.67×1012 5.81×1009 3.60×1009 1.97×1004 1.01×1003 4.55×10−03 2.05×1019
120 h 9.85×1015 8.18×1013 1.37×1015 2.10×1015 1.09×1016 3.28×1017 4.12×1010 2.29×1014 4.58×1014 2.92×1017 2.90×1017 1.29×1017 2.18×1017 1.50×1016 2.68×1017 3.53×1018 7.04×1007 3.79×1012 5.81×1009 3.60×1009 8.28×1004 1.41×1003 5.40×10−03 6.37×1018
Total
6.27×1020
1.46×1020
3.29×1019
7.96×1018
7.09×1019
Table 3.6. Inventory of selected relevant radionuclides in Bq for a pressurized water reactor with a thermal power of 3.733 MW after an operation time of 333 days [95SSK]
Nuclide 85 Kr 85m Kr 87 Kr 88 Kr 133 Xe 135 Xe Σ Kr-Xe
0h 1.61×1016 1.19×1018 2.27×1018 3.22×1018 7.62×1018 1.83×1018 1.61×1019
Time after discharge 1h 6h 1.61×1016 1.61×1016 1.04×1018 4.78×1017 18 1.33×10 8.72×1016 18 2.53×10 7.45×1017 18 7.62×10 7.60×1018 2.20×1018 3.07×1018 19 1.47×10 1.20×1019
24 h 1.61×1016 2.95×1016 4.78×1012 9.19×1015 7.35×1018 1.82×1018 9.22×1018
120 h 1.61×1016 1.05×1010 0.0 6.08×1005 4.75×1018 2.16×1015 4.77×1018
131
3.50×1018 5.18×1018 7.63×1018 8.38×1018 7.14×1018 3.18×1019
3.50×1018 5.16×1018 7.50×1018 6.00×1018 6.43×1018 2.86×1019
3.25×1018 4.27×1018 3.52×1018 1.96×1011 5.77×1017 1.16×1019
2.33×1018 1.82×1018 1.43×1017 0.0 2.45×1013 4.30×1018
I I 133 I 134 I 135 I Σ Iodine 132
3.45×1018 4.99×1018 6.40×1018 2.33×1017 3.81×1018 1.89×1019
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Nuclide 89 Sr 90 Sr 91 Sr 90 Y 91 Y 95 Zr 97 Zr 95 Nb 99 Mo 99m Tc 103 Ru 105 Ru 106 Ru 105 Rh 127 Sb 129 Sb 127 Te 127m Te 129 Te 129m Te 131m Te 132 Te 134 Cs 136 Cs 137 Cs 140 Ba 140 La 141 Ce 143 Ce 144 Ce 143 Pr 239 Np 238 Pu 239 Pu 240 Pu 241 Pu 241 Am 242 Cm 244 Cm Σ Aerosols
0h 4.49×1018 1.26×1017 5.42×1018 1.30×1017 5.58×1018 6.70×1018 6.52×1018 6.55×1018 6.91×1018 6.05×1018 4.88×1018 2.80×1018 6.57×1017 2.66×1018 2.58×1017 1.01×1018 2.50×1017 2.80×1016 9.89×1017 1.47×1017 4.86×1017 5.12×1018 8.26×1016 6.90×1016 1.51×1017 6.87×1018 6.95×1018 6.45×1018 6.18×1018 3.25×1018 6.10×1018 6.25×1019 5.76×1014 8.83×1014 5.91×1014 1.10×1017 4.16×1013 4.73×1015 2.92×1013 1.67×1020
Time after discharge 1h 6h 4.48×1018 4.47×1018 1.26×1017 1.26×1017 18 5.05×10 3.50×1018 17 1.30×10 1.30×1017 18 5.57×10 5.57×1018 18 6.70×10 6.68×1018 18 6.26×10 5.10×1018 18 6.55×10 6.55×1018 18 6.84×10 6.49×1018 6.05×1018 5.95×1018 18 4.88×10 4.86×1018 18 2.47×10 1.13×1018 17 6.57×10 6.57×1017 18 2.66×10 2.57×1018 17 2.57×10 2.50×1017 17 8.69×10 3.89×1017 17 2.50×10 2.49×1017 16 2.80×10 2.80×1016 17 9.52×10 5.49×1017 17 1.47×10 1.46×1017 17 4.77×10 4.25×1017 18 5.08×10 4.86×1018 16 8.26×10 8.26×1016 16 6.89×10 6.81×1016 1.51×1017 1.51×1017 18 6.86×10 6.78×1018 18 6.94×10 6.93×1018 18 6.45×10 6.44×1018 18 6.09×10 5.49×1018 18 3.25×10 3.25×1018 18 6.10×10 6.10×1018 19 6.21×10 5.85×1019 14 5.76×10 5.77×1014 14 8.84×10 8.85×1014 14 5.91×10 5.91×1014 17 1.10×10 1.10×1017 13 4.16×10 4.17×1013 15 4.74×10 4.75×1015 13 2.92×10 2.93×1013 20 1.65×10 1.55×1020
24 h 4.43×1018 1.26×1017 9.42×1017 1.29×1017 5.54×1018 6.63×1018 2.44×1018 6.55×1018 5.37×1018 5.14×1018 4.80×1018 6.82×1016 6.56×1017 1.91×1018 2.19×1017 2.17×1016 2.32×1017 2.80×1016 1.20×1017 1.44×1017 2.81×1017 4.14×1018 8.25×1016 6.55×1016 1.51×1017 6.51×1018 6.85×1018 6.35×1018 3.76×1018 3.25×1018 6.04×1018 4.69×1019 5.80×1014 8.88×1014 5.91×1014 1.10×1017 4.21×1013 4.76×1015 2.94×1013 1.30×1020
120 h 4.19×1018 1.26×1017 8.55×1014 1.27×1017 5.29×1018 6.35×1018 4.75×1016 6.55×1018 1.96×1018 1.89×1018 4.47×1018 2.10×1010 6.51×1017 2.93×1017 1.07×1017 4.42×1009 1.30×1017 2.79×1016 8.65×1016 1.33×1017 3.05×1016 1.77×1018 8.22×1016 5.30×1016 1.51×1017 5.24×1018 5.91×1018 5.83×1018 5.01×1017 3.21×1018 5.21×1018 1.45×1019 5.88×1014 8.96×1014 5.91×1014 1.10×1017 4.40×1013 4.71×1015 2.95×1013 7.50×1019
Total
7.75×1020
3.12×1020
1.75×1020
9.69×1019
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3.6.6 Release of radionuclides from the radioactive inventory of a nuclear reactor In case of an accident of a less dangerous category mainly gaseous or volatile fission or decay products are released. Heavier elements or compounds are more or less retained, since they are hardly transported over longer distances. Consequently, the respective risk assessment studies show the highest risk potentials for the gaseous fission or decay products like: noble gases, iodine, radon, volatile elements and compounds (cf. e.g. [95SSK]). Furthermore, it has to be taken into account that the chemical and physical behaviour of many fission products may change considerably after nuclear transformations within their radioactive decay chains (e.g.: Ra → Rn). This often leads to consequences for the retention behaviour, release of and filter effectiveness for various nuclides in the radioactive inventory.
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3.7 References 10Bat 34Fer 37HMS 79Ewb 80Clo 83FiW 93Wie 95AMC 95SSK
96Bec 97Lie 98Cur
98PKS 98Sch 99Mag 99MoT
Bateman, H.: Proc. Cambridge. Philos. Soc. 15 (1910) 423. Fermi, E.: Nature (London) 133 (1934) 898. Hahn, O., Meitner, L., Strassmann, F.: Z. Phys. 106 (1937) 249. Ewbank, W.B., Ellis, Y.A., Scmorak, M.R.: Nucl. Data Sheets 26 (1979) 1. Closs, K.D. (ed.): Report KfK 3000, 1980. Fischer, U., Wiese, H.W.: Report KfK 3014, 1983. Wiese, H.W.: Nucl. Technol. 102 (1993) 68. Adloff, J.P., MacCordick, H.J.: Radiochim. Acta 70/71 (1995) 13. Leitfaden für den Fachberater Strahlenschutz der Katastrophenschutzleitung bei kerntechnischen Notfällen, Veröffentlichungen der Strahlenschutzkommission Band 13; Herausgegeben vom Bundesministerium für Umwelt, Naturschutz und Reaktorsicherheit; 2. überarbeitete Auflage, Stuttgart, Jena, New York: Gustav Fischer Verlag, 1995. Becquerel, H.: C. R. Seances Acad. Sci. (Paris) 122 (1896) 501. Lieser, K.H.: Nuclear and radiochemistry: Fundamentals and applications, D-69451 Weinheim (Federal Republic of Germany): VCH Verlagsgesellschaft mbH, 1997. Sklodowska-Curie, M.: C. R. Seances Acad. Sci. (Paris) 126 (1898) 1101. Curie, P, Sklodowska-Curie, M.: C. R. Seances Acad. Sci. (Paris) 127 (1898) 175. Curie, P., Curie, M., Bémont, G.: C. R. Seances Acad. Sci. (Paris) 127 (1898) 1215. Sklodowska-Curie, M.: Rev. Gén. Sci. Pures Appl. Bull. Assoc. Fr. Av. Sci. 10 (1899) 41. Curie, P., Curie, M., Bémont, G.: Sci. Am. 80 (1899) 60. Curie, P., Curie, S.: C. R. Seances Acad. Sci. (Paris) 134 (1902) 85. Pfennig, G., Klewe-Nebenius, H., Seelmann-Eggebert, W.: Chart of the nuclides (Karlsruher Nuklidkarte), 6th ed., reprint 1998, Forschungszentrum Karlsruhe: Technik und Umwelt, 1998. Schmidt, G.C.: Verh. Dtsch. Phys. Ges. 17 (1898) 14; C. R. Seances Acad. Sci. Paris 126 (1898) 1264. Magill, J.: Nuclides 2000: An electronic chart of the nuclides; EUR 18737 EN, 1999. Mohr, P.J., Taylor, B.N.: J. Phys. Chem. Ref. Data 28 (1999) 1713.
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4 Radiological quantities and units
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4 Radiological quantities and units
In this Chapter the fundamental quantities and units for ionising radiation and in addition specific quantities used in radiological protection are described.
4.1 Introduction While radiation field quantities, quantities describing radioactivity and absorbed dose quantities are based on physical phenomena only, specific dose quantities in radiation protection as e.g. effective dose, include factors which are based on judgements about the biological response of tissues, e.g. due to cancer induction. These factors have been changed in the past in view of new research results and ideas. The definitions given are mainly based on ICRU Report 51 [93I1], ICRU Report 60 [98I1], ICRP Publication 60 [91I1] and the ISO Standards Handbook, Quantities and Units [93I2].
Stochastic and non-stochastic quantities Physical quantities are used to describe physical phenomena or objects. Many physical processes, e. g. the decay of radionuclides, the number of interactions in a small volume irradiated or the energy transferred, are subject to inherent fluctuations. This situation is described by stochastic quantities the values of which follow a probability distribution. Some times this may be a Poisson distribution which is uniquely determined by its mean value. In many other cases a quantity is defined by averaging in time or over a volume which results in a single value with no inherent fluctuation. Those quantities, e. g. fluence or absorbed dose, are called non-stochastic quantities.
Units A unit is a reference sample of a quantity with which other quantities of the same kind are compared. Every quantity is expressed as a product of a numerical value and a unit. Generally the use of the International System of Units (SI) as given by the BIPM [91BI] is recommended which is based on the 7 base units meter, kilogram, second, ampere, kelvin, mole and candela. Derived SI-units are often given special names like joule, becquerel or gray. Some other units are, however, generally used which are outside of the international system, e. g. the electron volt (eV) and the atomic mass unit (u) - and the time units minute, hour, day and year are also generally permitted. Nevertheless, other units even if not recommended are still in use in radiation measurements and radiation protection. Table 1 presents some numerical relationships between those units and the SI-units recommended. Landolt-Börnstein New Series VIII/4
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Table 4.1. Former units and its relations to SI-units Quantity
Symbol
SI-unit Name
Former units rad
1 rad = 0.01 Gy
roentgen
1 R = 2.58 10−4 J kg−1
rem
1 rem = 0.01 Sv
absorbed dose
D
J kg−1
exposure
X
C kg−1
dose equivalent
H
J kg−1
sievert (Sv)
activity
A
s−1
becquerel (Bq) curie
potential alpha energy
cp
J m−3
gray (Gy)
concentration potential alpha energy exposure
1 Ci = 3.7 1010 Bq
Working level 1 WL = 2.08 10−5 J m−3 = 1.30 108 MeV m−3
Ep
J h m−3
Working level month (T = 170 h) 1 WLM = 3.54 10−3 J h m−3 = 2.21 1010 MeV h m−3
4.2 Radiation field quantities Radiation field quantities are non-stochastic quantities defined at any point of a radiation field. Radiation fields may consist of various types of particles and the field quantities are always related to a specific particle type. This is usually expressed by adding the particle name to the quantity, e.g. photon fluence or neutron flux. There are two classes of radiation field quantities referring either to the number of particles or to the energy transported by them. A radiation field of a specific particle type can be fully described by the number N of particles, their distribution in energy as well as their spatial, directional and temporal distribution. This needs the definition of scalar and vectorial quantities. While in radiation dosimetry mostly scalar field quantities are used, vectorial quantities are often needed and applied in radiation transport theory and calculations. The radiation field quantities are defined in specifying the field in increasing detail.
4.2.1 Scalar radiation field quantities Particle number, radiant energy The particle number, N, is the number of particles that are emitted, transferred, or received. Unit: 1
The radiant energy, R, is the energy (excluding rest energy) of the particles that are emitted, transferred, or received. Unit: joule, J
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For particles of energy E (excluding rest energy), the radiant energy, R, is equal to N·E. The distributions, NE and RE, of the particle number and the radiant energy with respect to energy are given by NE = dN/dE
and RE = dR/dE
(4.2.1a+b)
where dN is the number of particles with energy between E and E + dE and dR is their radiant energy. Flux, energy flux The flux, N& , is the quotient of dN by dt, where dN is the increment of the particle number in the time interval dt. N& = dN/dt Unit: s−1
The energy flux, R& , is the quotient of dR by dt, where dR is the increment of the radiant energy in the time interval dt. R& = dR/dt Unit: W
The term flux has often been employed for the quantity fluence rate (see below). This usage should be avoided. Fluence, energy fluence The quantity fluence is based on the idea of counting the number of particles incident or passing a small sphere. It is defined by: The fluence, Φ, is the quotient of dN by da, where dN is the number of particles incident on a sphere of cross-sectional area da. Φ = dN /da Unit: m−2
The energy fluence, Ψ, is the quotient of dR by da, where dR is the radiant energy incident on a sphere of cross-sectional area da. Ψ = dR /da Unit: J m−2
The fluence is independent of the directional distribution of the particles passing the sphere. In calculations, fluence is often expressed in terms of the length of trajectories of particles passing a volume dV. The fluence, Φ, is given by
Φ = dl /dV
(4.2.2)
where dl is the sum of the lengths of trajectories through this volume. The distributions, ΦE and ΨE, of the fluence and energy fluence with respect to energy are given by
ΦE = dΦ/dE
and ΨE = dΨ/dE
(4.2.3a+b)
These quantities are often called spectral fluence and spectral energy fluence, respectively. Fluence rate, energy fluence rate The temporal distribution of the fluence and energy fluence is generally of interest. This results in the following definitions: The fluence rate, Φ& , is the quotient of dΦ by dt, where dΦ is the increment of the fluence in the time interval dt. Φ& = dΦ /dt Unit: m−2 s−1
Landolt-Börnstein New Series VIII/4
The energy fluence rate, Ψ& , is the quotient of dΨ by dt, where dΨ is the increment of the energy fluence in the time interval dt. Ψ& = dΨ /dt Unit: J m−2
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[Ref. p. 4-27
The fluence rate has often been termed particle flux density. Because the term density mostly characterises a mass density (kg−1), it is recommended to use the term fluence rate and not particle flux density. Particle radiance, energy radiance The energy radiance, Ψ&Ω , is the quotient of d Ψ& by dΩ, where dΨ& is the energy fluence rate of particles propagating within a solid angle dΩ around a specified direction. Ψ&Ω = dΨ& / dΩ Unit: W m−2 sr−1
The particle radiance, Φ&Ω , is the quotient of d Φ& by dΩ, where d Φ& is the fluence rate of particles propagating within a solid angle dΩ around a specified direction. Φ& Ω = dΦ& / dΩ Unit: m−2 s−1 sr−1
The specification of a direction Ω requires two variables. In a spherical coordinate system with a polar angle, θ, and an azimuthal angle, ϕ, dΩ is equal to sinθ dθ dϕ. The distribution of particle radiance and energy radiance with respect to energy are given by
Φ& Ω ,E = dΦ& / dΩ dE and Ψ&Ω ,E = dΨ& / dΩ dE
(4.2.4a+b)
4.2.2 Vectorial radiation field quantities Radiometric quantities are often used to describe the flow of radiation in specific directions. This needs the definition of vectorial quantities. For example, the scalar angular differential quantities like particle radiance and energy radiance are transferred to vectorial quantities by multiplication with the unit vector Ω in a specific direction. Vectorial quantities are vectorial particle radiance, Φ& Ω
with
vectorial energy radiance, Ψ& Ω vectorial fluence rate, Φ&
with
vectorial energy fluence rate, Ψ&
with
vectorial fluence, Φ
with
vectorial energy fluence, Ψ
with
with
& = Ω ⋅ Φ& Φ Ω Ω & & ΨΩ = Ω⋅ΨΩ Φ& = Φ& dΩ
∫ & Ψ = ∫ Ψ& dΩ Φ = ∫ Φ& dt Ψ = ∫ Ψ& dt Ω
Ω
unit: m−2 s−1 sr−1, unit: W m−2 sr−1, unit: m−2 s−1, unit: W m−2, unit: m−2, unit: J m−2.
A detailed description is given in ICRU Report 60 [98I1]. The distribution of a quantity with respect to energy of the particle considered is described by an index E similar to the scalar quantities. For example, the distribution of the vectorial particle radiance is given by Φ& Ω ,E = Ω ⋅ Φ& Ω ,E unit: m−2 s−1 sr−1 MeV−1.
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4-5
4.3 Interaction coefficients and quantities Ionising radiation is either charged (e.g. electrons, positrons, protons and α-particles) or uncharged (e.g. photons and neutrons). This dominates the main interaction with matter. While charged particles (called directly ionising particles) are mainly slowed down by electromagnetic interactions with electrons of the target atoms, the uncharged particles (indirectly ionising particles) interact with matter in separated events. Indirectly ionising particles are either absorbed or its energy and direction is altered. The probabilities of specific interactions between radiation and matter are characterized by interaction coefficients. They refer to specific interaction processes, type and energy of radiation and the matter involved. The definition of those coefficients important for dosimetry and related quantities are given in this Section.
4.3.1 Cross section The cross section is the most fundamental interaction coefficient. It is defined as follows. The cross section, σ, of a target entity, for a particular interaction produced by incident particles is the quotient of P by Φ, where P is the probability of that interaction for a single target entity when subjected to the particle fluence, Φ. It is
σ = P/Φ
unit: m2.
A special unit often used for the cross section is the barn (b) with 1 b = 10−28 m2. Cross sections mostly vary with the energy of the incident radiation (notation: σ(E)). The distribution of a cross section with respect to the energy and direction of the emitted radiation is often called differential cross section (dσ/dΩ : angular differential cross section, dσ/dE: energy differential cross section, d2σ/dEdΩ : energy and angular differential cross section). The total cross section, σT, is the sum of the cross sections of all possible interaction channels for an incident particle of a given type and energy and a given target material.
4.3.2 Mass attenuation coefficient and mass energy transfer coefficient For an infinite small parallel beam of uncharged radiation, the interaction of radiation with matter results in an attenuation of the incident beam with depth in material. This can be described by the relation dN = µ ⋅ dl N
(4.3.1)
where dN/N is the fraction of particles that experience interactions in traversing a distance dl in the material. µ is the linear attenuation coefficient. The reciprocal of µ is called the mean free path λ of an uncharged particle. In first order µ is proportional to the density ρ of a material. This leads to the definition of the mass attenuation coefficient, µ/ρ . µ 1 dN = ρ ρ dl N
Landolt-Börnstein New Series VIII/4
unit: m2 kg−1
(4.3.2)
4-6
4 Radiological quantities and units
[Ref. p. 4-27
The mass attenuation coefficient is related to the total cross section by N µ NA = σ = A ρ M T M
∑σ
(4.3.3)
J
J
where NA is the Avogadro constant and M the molar mass of the material considered. σJ are the cross section related to the interaction of type J in this material. For uncharged particles the transfer of energy to charged particles in the material is of high interest in dosimetry. This is expressed by the mass energy transfer coefficient, µtr/ρ , which is defined by µ tr 1 dRtr = ρ ρ dl R
unit: m2 kg−1
(4.3.4)
where dRtr/R is the fraction of incident radiant energy that is transferred to kinetic energy of charged particles by interactions when traversing a distance dl in the material of density ρ. If incident uncharged particles of a given type and energy can produce several types of interactions in a material, µtr/ρ can be expressed in terms of the partial cross Sections, σ J, by the relation µ tr N A = ρ M
(4.3.5)
∑ f Jσ J J
where fJ is the average fraction of the incident particle energy that is transferred to kinetic energy of charged particles in an interaction of type J. The mass energy transfer coefficient is related to the mass attenuation coefficient by µ tr µ µ = f = ρ ρ ρ
∑ f Jσ J J ∑σ J
(4.3.6)
J
For µtr/ρ of a compound material the material is usually treated as consisting of independent atoms and the contributions from the different components are summed considering their partial density. A small part of the energy transferred to charged particles may not be locally absorbed in the material but further transferred to secondary photons (e.g. Bremsstrahlung). Therefore, an additional coefficient, the mass energy absorption coefficient, µen/ρ, is defined by the product of µtr/ρ and (1-g) where g is the fraction of the energy of charged particles that is lost in radiative processes in the material. Data of mass energy transfer and mass energy absorption coefficients are given by Seltzer [93Se]. For neutron radiation, the kerma coefficient K/Φ (kerma per unit neutron fluence, often called kerma factor) is mostly used instead of µtr/ρ for characterising the energy transfer (see 4.4.1). Data of kerma coefficients for biological important materials from thermal to 150 MeV neutrons are published by Chadwick et al. [99Ch] and in ICRU Report 63 [00I1].
4.3.3 Mass stopping power and linear energy transfer (LET) Charged particles passing matter loose energy by collisions with electrons, by emission of bremsstrahlung in the electric fields of nuclei or atomic electrons or by elastic Coulomb scattering and inelastic nuclear processes on atoms or nuclei. This effect is characterized by the mass stopping power S/ρ for charged particles in a material with density ρ. It is S
ρ
=
1 dE ρ dl
unit: J m2 kg−1
(4.3.7) Landolt-Börnstein New Series VIII/4
Ref. p. 4-27]
4 Radiological quantities and units
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where dE is the energy lost by a charged particle in traversing a distance dl in the material. S =dE/dl is called the linear stopping power. E may be given in eV and the unit of S/ρ may then expressed in eV m2 kg−1 or other multiples like MeV cm2 g−1, for example. The transfer of energy from the primary charged particle to secondary electrons is of specific interest in dosimetry, especially to those electrons receiving a kinetic energy less than a given value only. They will locally deposit their energy near to the track of the primary particle. This led to the definition of the quantity linear energy transfer (LET) or restricted linear electronic stopping power L∆ given by L∆ =
dE∆ dl
unit: J m−1, often used keV µm−1
(4.3.8)
where dE∆ is the energy lost by a charged particle due to electronic collisions when traversing a distance dl minus the sum of the kinetic energies of all electrons released with kinetic energies in excess of ∆. This definition given in ICRU Report 60 [98I1] differs from earlier ones [80I1] in a way that L∆ now includes the binding energies for all collisions and the threshold of the kinetic energy of the released electrons is now ∆ instead of ∆ minus the binding energy. ∆ is often given in eV and then the notation L100 means an energy cutoff of 100 eV. L∞ is often called unrestricted linear energy transfer L and is equal to Sel, the electronic stopping power due to collisions with electrons.
4.3.4 Mean energy expended in a gas per ion pair formed In dosimetry, where often charge measurements due to ionisation in gases are the basis of dose determinations, the kinetic particle energy necessary to create an ion pair is of general interest. This led to the definition of the mean energy expended in a gas per ion pair formed W. It is W=
E N
unit: J
(4.3.9)
where N is the number of ion pairs when the initial kinetic energy E of the charged particle is completely dissipated in the gas considered. This definition includes also the ions produced by secondary electrons or bremsstrahlung.
4.4 Quantities related to energy transfer 4.4.1 Stochastic quantities The energy transfer from incident particles to a target material is a stochastic process. For example, the energy deposition along a track of a charged particle is randomly distributed. The values of stochastic quantities are, therefore, subject to inherent fluctuations. They generally follow a probability distribution and mean values may be given. For example, a Poisson distribution is already uniquely determined by its mean value. Stochastic quantities are often used in microdosimetry in order to describe the energy transfer to very small volumes.
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[Ref. p. 4-27
4.4.1.1 Energy deposit and energy imparted The energy deposit ε i, is the energy deposited in a single interaction i, thus
ε i = ε in − ε out + Q
unit: J, often used eV
where ε in is the energy of the incident ionising particle (excluding rest energy), ε out the sum of energies of all ionising particles leaving the interaction (excluding rest energy) and Q the change in the rest energies of the nucleus and all particles involved in the interaction. Q > 0 means a decrease of rest energy, Q < 0 an increase. The total energy transferred to matter in a given volume is often of interest. The energy imparted ε to the matter in a given volume is the sum of all energy deposits in the volume
ε=
∑ε
unit: J, often used eV
i
The mean energy imparted ε to the matter in a given volume is a non-stochastic quantity and can be expressed in terms of the radiant energy Rin (sum of all radiant energies of the incoming particles) and Rout (sum of the radiant energies of all outgoing particles). It is
ε = Rin − Rout +
∑Q ,
unit: J, often used eV
4.4.1.2 Lineal energy and specific energy Corresponding to the non-stochastic quantity LET the stochastic quantity lineal energy y is defined by the quotient of ε s by l , where ε s is the energy imparted to the matter in a given volume by a single energy deposition event and l is the mean chord length of that volume, thus
y=
εs l
unit: J m−1, mostly used keV µm−1
This quantity is mainly used in microdosimetry, especially in measurements with low-pressure tissueequivalent proportional counters where single event distributions in terms of y are measured. The specific energy (imparted) z is the quotient of ε by m, where ε is the energy imparted to matter of mass m. It is
z=
ε m
unit: gray (Gy), 1 Gy = 1 J kg−1
The specific energy includes the energy transferred to the matter m from all events involved.
4.4.2 Non-stochastic quantities 4.4.2.1 Kerma, kerma rate The transfer of energy from uncharged particles (indirectly ionising particles, e.g. photons or neutrons) to matter is performed by the liberation and slowing down of secondary charged particles in this matter. This led to the definition of the quantity kerma. The kerma K is the quotient of dEtr by dm, where dEtr is the sum of the kinetic energies of all charged particles liberated by uncharged particles in a mass dm of material. It is given by K=
dE tr , dm
unit: gray (Gy), 1 Gy = 1 J kg-1
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Kerma is a non-stochastic quantity. For a very small mass element, however, the energy transfer dEtr underlies in principle stochastic fluctuations. In this case a non-stochastic quantity means that dEtr is seen to be the expectation value of the sum of energies of liberated charged particles. For monoenergetic uncharged particles of energy E the kerma is related to the fluence by K = Φ E (µtr/ρ)
(4.4.1)
For a given energy distribution ФE of the uncharged particles the kerma can be calculated by K = ∫ ΦE E (µtr/ρ) dE
(4.4.2)
For neutrons, the quotient of K by Φ, is called kerma coefficient (often also called kerma factor) where Φ is the neutron fluence (see 4.3.2). The kerma rate, K& , is the quotient of dK by dt, where dK is the increment of K in the time interval dt. dK , K& = dt
unit: Gy s−1.
4.4.2.2 Absorbed dose, absorbed dose rate The quantity absorbed dose is a basic quantity in radiation dosimetry and relevant to all types of ionising radiation whether directly or indirectly ionising. It is a non-stochastic quantity and defined by: D=
ε
unit: gray (Gy), 1 Gy = 1 J kg−1
dm
where ε is the mean energy imparted to the matter of mass dm. While kerma is related only to those secondary charged particles produced in dm but transferring their energy to matter partially also outside dm, absorbed dose includes all energy transferred to dm partially also from secondary charged particles produced outside but entering dm. Only under charged particle equilibrium and negligible radiation losses, however, the values of absorbed dose and kerma are equal in a homogeneous material. The absorbed dose rate, D& , is the quotient of dD by dt, where dD is the increment of the absorbed dose in the time interval dt. It is dD D& = dt
unit: Gy s−1
4.4.2.3 Exposure, exposure rate The quantity exposure is related to the production of charges in gas by ionising radiation. Historically its definition is elder than kerma or absorbed dose. Its use, however, is restricted to photons only. The exposure X is the quotient of dQ by dm, where dQ is the absolute value of the total charge of the ions of one sign produced in air when all the electrons and positrons liberated or created by photons in air of mass dm are stopped in air.
X =
dQ dm
unit: C kg−1 (former: roentgen, R)
It should be noted that in this definition the charges due to ionisation arising from the absorption of bremsstrahlung emitted by the electrons is not included in dQ. Landolt-Börnstein New Series VIII/4
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4 Radiological quantities and units
[Ref. p. 4-27
The exposure rate, X& , is the quotient of dX by dt, where dX is the increment of the exposure in the time interval dt. It is
dX X& = dt
unit: C kg−1 s−1
4.5 Dose quantities in radiation protection 4.5.1 Concept of radiation protection quantities The development of dosimetric concepts and the definition of specific quantities for use in radiation protection have a long history. An important basis for the present concepts was already provided in the 60's and 70's by both the International Commission on Radiological Protection (ICRP) and the International Commission on Radiation Units and Measurements (ICRU). In 1991 in its Publication 60 [91I1], the ICRP has published its most recent general recommendations for radiation protection including a system of quantities. The ICRP and ICRU have developed a hierarchy of quantities for radiation protection applications comprising primary limiting dose quantities (called “protection quantities”) taking account of human body properties and operational dose quantities for monitoring of external exposure. For monitoring of internal exposure other quantities than dose quantities are used. The basic idea of a primary limiting quantity is to relate the “risk” of exposure to ionising radiation (exposure by internal and external radiation sources) to a single (dose) quantity which takes account of the man as a receptor, the different radiation sensitivities of various organs and tissues and the different radiation qualities. Other influence parameters, however, e.g. the influence of dose and dose rate or sex and age of a person exposed on the biological response and the exposure risk, were not explicitly considered in the definition of these quantities. Operational quantities are dose equivalent quantities defined for use in radiation protection measurements related to external exposure (area or individual monitoring). They are needed for monitoring external exposures because • • • •
protection quantities are generally not measurable, for area monitoring a point quantity is needed, a non-isotropic human-body related quantity like effective dose is not appropriate in area monitoring, instruments for radiation monitoring need to be calibrated in terms of an operational quantity.
Operational quantities usually provide an estimate or upper limit for the value of the limiting quantities due to an exposed, or potentially exposed, person under most irradiation conditions. They are often used in practical regulations instead of the primary limiting quantities. For internal exposure, however, other methods are used and no similar dose quantities have been defined. In this case organ doses or effective dose are estimated from the information on intake or excretion of radioactive substances. Model based conversion coefficients exist for a large number of radionuclides relating the intake to organ doses and effective dose (see 4.6 and Chapter 7). Both, protection and operational quantities can be related to “radiation field quantities” (see Sect. 4.2) or air kerma (see Sect. 4.3) which are point quantities defined in any point of a radiation field and whose units are directly realised through primary standards at national standards laboratories since long time. The numerical relations (conversion coefficients) between those quantities and the protection or operational dose quantities are given in Chapter 6.
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4.5.2 Protection quantities In 1977 the ICRP [77I1] introduced the tissue (or organ) dose equivalent HT and the effective dose equivalent HE whose definition takes care of the relative variation of the tissue response with different types of radiation and different tissues or organs in the human body by introducing tissue weighting factors [73Jac]. Although in general this concept was not changed by ICRP 60 [91I1] in 1990, important modifications, however, were introduced e.g. replacing dose equivalent quantities by equivalent dose quantities. The present system of quantities is summarised in the following. 4.5.2.1 Absorbed dose and equivalent dose in a tissue or organ The absorbed dose in a tissue or organ DT is the absorbed dose averaged over the volume of a tissue or organ T (rather than at a point). While the absorbed dose at a point generally is the fundamental dose quantity, in radiation protection the mean dose in an organ becomes the basic protection quantity correlated with the exposure risk. This concept is obviously based on the linear dose-effect relationship and the additivity of doses for risk assessment as an appropriate approximation in the low dose range. The equivalent dose in a tissue or organ is defined by HT =
∑w D R
T, R
unit: sievert (Sv) (1 Sv = 1 J kg−1)
R
where DT,R is the mean organ dose in the tissue or organ T from radiation of type R incident on the human body and wR are radiation weighting factors characterising the biological effectiveness of the specific radiation R relative to photons. These factors have replaced the mean quality factors used in the concept of organ dose equivalent before [77I1]. The sum is taken over all types of radiation involved. 4.5.2.2 Radiation weighting factors For external irradiation, the values of the radiation weighting factors wR are given by the parameters of the external radiation field only (type and energy distribution of the radiation incident on the body). This means that wR is a body-averaged value representing a mean value for the relative biological effectiveness of all tissues of the body and any local variation of the radiation quality in the human body which may result from the generation of secondary radiation of different types in the body, is not explicitely considered. This effect is mainly important in the case of incident neutrons where at low energies secondary photons strongly contribute to the absorbed doses of various organs. The wR values for various types of radiation are specified in ICRP 60 in a table (see Table 4.2). For photons, electrons and muons of all energies a value of one is fixed with the exception of Auger electrons emitted from nuclei bound to DNA. For this case there exists no ICRP recommendation until now. The radiation weighting factor for neutrons depends on the neutron energy. Different wR values are given by either a step function or a continuous function as an approximation (see Fig. 4.1). In practice, neutron fields contain neutrons with a broad energy distribution. Because the use of a continuous wRfunction for effective dose estimation is more appropriate in these cases it is recommended to apply the continuous function in any case to avoid ambiguities. Then the weighting factor for neutrons ranges from 5 to 22 depending on neutron energy with its maximum value at 500 keV. All conversion coefficients for neutrons published in ICRP 74 [96I1] and ICRU 54 [98I2] are based on the continuous function only (see Chapt. 6).
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[Ref. p. 4-27
Table 4.2. Radiation weighting factors wR Radiation
wR
Photons Electrons1), muons
1 1
Neutrons: En En En En En
5 10 20 10 5
<10 keV =10 keV to 100 keV >100 keV to 2 MeV >2 MeV to 20 MeV >20 MeV
Protons: Ep
5
> 2 MeV (unless recoil protons)
α-particles, fission fragments, heavy nuclei
20
As an approximation to the step function introduced for neutrons ICRP has specified a smooth wR function: wR = 5 + 17 exp (−[ln(2 En)]2/6) with En neutron energy in MeV.
1) With the exception of Auger electrons from atoms bound to DNA
The radiation weighting factor for incident external protons with energies above 2 MeV has been set to 5. It is, however, questioned if this value is appropriate for protons of all energies above 2 MeV. There exists a general opinion that a weighting factor of about 2 seems to be more realistic for high energy protons above about 5 to 10 MeV. External protons of lower energies have a small range in tissue and contribute to the skin dose only. 4.5.2.3 Effective dose The effective dose E is the weighted sum of the equivalent doses in tissues and organs T: E=
∑w H T
T
with
T
∑w
T
=1
unit: sievert (Sv)
(4.5.1)
T
where wT are tissue weighting factors characterising the relative sensitivity of the various tissues with respect to stochastic effects resulting from ionising radiation exposure and HT is the equivalent dose in one of the 13 specified tissues and organs (see Sect. 4.5.2.4). The effective dose is a quantity which is not sex specific or dependent on age of a person. In principle, the effective dose is determined by taking the dose values in all tissues and organs of an individual person. Those data, however, are never measurable. For external exposure, therefore, always calculated conversion coefficients are used which relate the external radiation field to the doses in the tissues and organs (see Chapt. 6). Following ICRP Report 74 [96I1], the effective dose is then calculated by E = wbreast H breast, female +
∑
T ≠ breast
wT
H T, male + H T, female 2
(4.5.2)
Under a given exposure condition (radiation field, direction of radiation incidence, exposure period), therefore, all persons are given the same effective dose value independent of sex and age. Landolt-Börnstein New Series VIII/4
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4.5.2.4 Tissue Weighting Factors The definition of effective dose takes care of the different radiosensitivity of the various organs and tissues in the human body with respect to cancer induction and mortality by introducing tissue weighting factors. Twelve tissues and organs are specified with individual weighting factors wT. The values have been developed from a reference population of equal numbers of both sexes and a wide range of ages. They are applied to workers, to the whole population, and to either sex including children and the unborn child (foetus). An additional “remainder” tissue with a weighting factor of 0.05 is also defined [91I1]. Its dose is given by the mean value from ten specified tissues and organs (see Table 4.3). The upper large intestine formerly included in the remainder, is now considered as part of the colon and has been replaced by the extrathoracic airways [93I3, 94I2]. While in the calculation of conversion coefficients for the intake of radionuclides the remainder dose is obtained from the mass-weighted doses to the single tissues and organs involved, the coefficients for external exposure are calculated giving identical weights to each of the remainder tissues [96I1]. Table 4.3. Tissue weighting factors wT Organ or tissue Gonads Bone marrow (red) Colon Lung Stomach Bladder Breast Liver Oesophagus Thyroid Skin Bone surface Remainder1)
wT 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
1) “Remainder” tissues are adrenals, brain, extrathoracic airways, small intestine, kidney, muscle, pancreas, spleen, thymus and uterus. The mean value of the equivalent doses of the ten remainder organs and tissues is to be multiplied by 0.05. If in a special case a single tissue or organ has an equivalent dose higher than each of the 12 individually defined organs and tissues, then this organ or tissue should get a weighting factor of 0.025 and the other 9 remainder tissues together a weighting factor of 0.025.
4.5.2.5 Committed or collective equivalent dose and effective dose Several subsidiary dosimetric quantities have been additionally defined. After an intake of radionuclides to a body these nuclides may give rise to equivalent doses in different tissues and organs of the body spread over long time depending on the physical and biological half-life of the radionuclides and their biokinetic behaviour in the body. The time integral of the equivalent dose rate is called the committed equivalent dose HT(τ ), in a tissue or organ T, where τ is the integration time (in years) following the intake at time t0. t0 +τ
H T (τ ) =
∫ H&
T
( t )dt
(4.5.3)
t0
If τ is not specified, it is implied that its value is 50 y for workers and from intake up to age 70 years for members of the public including children. For patients in nuclear medicine, the integration may run from t0 to ∞ because the biological and physical half-life of the radionuclides applied is much less than 10 y.
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The same specification holds for the quantity committed effective dose E(τ ) defined by the weighted sum of HT(τ ) over all specified tissues and organs T. E( τ ) =
∑w
T
⋅ H T (τ )
(4.5.4)
T
All dosimetric quantities referred before are related to a single tissue or organ of a single individual. Often it may be of interest to quantify the total dose a number of people received from one source or one release of radioactive material. The relevant quantity is called collective equivalent dose ST in a tissue or organ T and is defined by ∞
∫
ST = H T ⋅ 0
dN dH T dH T
unit: man sievert (man Sv)
(4.5.5)
where (dN/dHT)dHT is the number of individuals receiving an equivalent dose between HT and HT+dHT or by ST =
∑H
T, i
(4.5.6)
Ni
i
where Ni is the number of individuals in a subgroup i receiving a mean tissue equivalent dose HT,i . The summed effective doses of all members of a group or population is called the collective effective dose S defined in a similar way by ∞
∫
S = E⋅ 0
dN dE dE
or
S=
∑E ⋅N i
i
unit: man Sv
(4.5.7a+b)
i
where Ni is a subgroup i receiving a mean equivalent dose Ei .
4.5.3 Operational quantities 4.5.3.1 Dose equivalent and quality factor The radiation protection quantity dose equivalent H is defined by unit: Sv (1 Sv = 1 J kg−1)
H=QD
where D is the absorbed dose at the point of interest and Q a quality factor weighting the relative biological effectiveness of radiation. Q is defined as a function of the linear energy transfer L of a charged particle in water [77I1]. In principle, this concept of Q has not been changed by ICRP 60 [91I1], but the dose equivalent is now restricted to the definition of operational radiation protection quantities and the quality factor function Q(L) was modified in 1991 according to the following equation:
Q(L) =
1 0.32 L – 2.2 300/√L
for L < 10 keV/µm for 10 ≤ L ≤ 100 keV/µm for L > 100 keV/µm
(4.5.8)
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The quality factor Q at a point in tissue is then given by [86I1]: ∞
Q=
1 Q( L )DL dL D L=0
∫
(4.5.9)
where DL is the distribution of D in L at the point of interest. This function is most important for neutrons because various types of secondary charged particles are produced in tissue in this case. 4.5.3.2 The concept of operational quantities The basic concept of the operational quantities is described in the ICRU Reports 39 and 43 [85I1, 88I1]. They have been introduced linking the external irradiation to the effective dose and the equivalent dose of the skin and eye lens in order to control their limits. The present definitions are given in ICRU Report 51 [93I1]. The operational quantities for radiation protection are dose equivalent quantities defined either for strongly penetrating or for weakly penetrating radiation incident on the human body (sometimes also the expressions penetrating and low penetrating are used instead of strongly and weakly penetrating radiation). The radiation is characterised as either weakly- or strongly penetrating depending on which dose (effective dose or skin equivalent dose) is closer to the corresponding limit. Weakly penetrating radiations are α-particles, β-particles with energies below 2 MeV and photons with mean energies below about 12 keV. Photons above this energy, electrons above 2 MeV and all neutrons are strongly penetrating radiation. Due to the different tasks in radiation protection monitoring − area monitoring for controlling the radiation at work places and definition of controlled or forbidden areas or individual monitoring for the control and limitation of individual exposures − different operational quantities were defined. While measurements with an area monitor are mostly performed free in air, an individual dosemeter is usually worn on the front of the body. As a consequence, in a given situation, the radiation field “seen” by an area monitor free in air differs from that “seen” by an individual dosemeter worn on a body where the radiation field is strongly influenced by the backscatter and absorption of radiation in the body. The operational quantities allows for this effect. They may be presented as follows:
Radiation type
Operational quantities for area monitoring individual monitoring
Strongly penetrating radiation Weakly penetrating radiation
ambient dose equivalent directional dose equivalent
personal dose equivalent personal dose equivalent
4.5.3.3 Operational quantities for area monitoring ICRU sphere phantom For all types of radiation the operational quantities for area monitoring are defined on the basis of a dose equivalent value at a point in a simple phantom, the ICRU sphere. It is a sphere of tissue-equivalent material (30 cm in diameter, density: 1 g cm−3, mass composition: 76.2 % oxygen, 11.1 % carbon, 10.1 % hydrogen and 2.6 % nitrogen). It adequately approximates the human body as regards the scattering and attenuation of the radiation fields under consideration.
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Aligned and expanded radiation field The operational quantities for area monitoring defined in the ICRU sphere should retain their character of a point quantity and the property of additivity. This is achieved by introducing the terms “expanded” and “aligned” radiation field in the definition of these quantities (see Fig. 4.1).
Φ
Φ ÄUHDOILHOG³
Φ Φ
Φ
ÄDOLJQHGILHOG³
ΦΦΦΦ Φ
ÄDOLJQHGDQGH[SDQGHGILHOG³
Fig. 4.1. Aligned and expanded field concept.
An expanded radiation field is a radiation field in which the spectral and the angular fluence have the same values in all points of a sufficiently large volume equal to the values in the actual field at the point of interest. The expansion of the radiation field ensures that the whole ICRU sphere is thought to be exposed to a homogeneous radiation field with the same fluence, energy distribution and directional distribution as in the point of interest of the real radiation field. If all radiation is (thought to be) aligned in the expanded radiation field so that it is opposed to a radius vector Ω specified for the ICRU sphere, the aligned and expanded radiation field is obtained. In this fictitious radiation field, the ICRU sphere is homogeneously irradiated from one direction, and the fluence of the field is the integral of the angular differential fluence at the point of interest in the real radiation field over all directions. In the expanded and aligned radiation field, the value of the dose equivalent at any point in the ICRU sphere is independent of the directional distribution of the radiation of the real radiation field. Ambient dose equivalent H*(d ) For area monitoring of strongly penetrating radiation the operational quantity is the ambient dose equivalent H*(10) defined by:
Landolt-Börnstein New Series VIII/4
Ref. p. 4-27]
4 Radiological quantities and units
4-17
The ambient dose equivalent H*(d) at a point of interest in the real radiation field, is the dose equivalent that would be produced by the corresponding aligned and expanded radiation field, in the ICRU sphere at a depth d, on the radius vector opposing the direction of radiation incidence. For strongly penetrating radiation it is d = 10 mm and H*(d) is written H*(10). While this definition with the parameter d is given in ICRU [93I1] and ICRP [91I1] the most recent ICRU Report [01I1] dealing also with operational quantities defines ambient dose equivalent by H*(10), thus restricting its definition to strongly penetrating radiation only. In practice, however, this has already been realised because other values have never been used. As a result of the imaginary alignment and expansion of the radiation field, the contributions of radiation from all directions add up. The value of H*(10) is therefore independent of the directional distribution of the radiation in the actual field. This means that the reading of an area dosemeter for the measurement of H*(10) should be independent of the directional distribution of the radiation − an ideal detector should have an isotropic fluence response. Directional dose equivalent, H'(d ,Ω ) For area monitoring of weakly penetrating radiation the operational quantity is the directional dose equivalent H'(0.07,Ω ) or, in rare cases, H'(3,Ω ) defined by. The directional dose equivalent H'(d,Ω ) at a point of interest in the actual radiation field, is the dose equivalent that would be produced by the corresponding expanded radiation field, in the ICRU sphere at a depth d, on a radius in a specified direction Ω. For weakly penetrating radiation it is d = 0.07 mm and H'(d,Ω ) is written H'(0.07,Ω ). In case of monitoring the dose to the eye lens H'(3,Ω ) with d = 3 mm may be chosen. In practice H'(0.07,Ω ) is almost exclusively used in area monitoring for weakly penetrating radiation. For unidirectional radiation incidence the quantity may be written H'(0.07,α), where α is the angle between the direction Ω and the direction opposite to radiation incidence. The value of the directional dose equivalent can strongly depend on the direction Ω. The same is true for instruments for measuring weakly penetrating radiation − e.g. beta- or alpha-particle radiation – the reading of which can strongly depend on the orientation in space. In radiation protection practice, however, it is always the maximum value of H'(0.07,Ω ) at the point of interest which is of importance. It is usually obtained by rotating the dose rate meter during the measurement and looking for the maximum reading. 4.5.3.4 Operational quantities for individual monitoring Individual monitoring is usually performed with individual dosemeters worn on the body and the operational quantity defined for this application takes into account this situation. The true value of the operational quantity is determined by the irradiation situation near the point where the dosemeter is worn. For individual monitoring the operational quantity is the personal dose equivalent Hp(d). The personal dose equivalent Hp(d) is the dose equivalent in ICRU tissue at a depth d in a human body below the position where an individual dosemeter is worn. For strongly penetrating radiation a depth d = 10 mm is recommended. For weakly penetrating radiation a depth d = 0.07 mm is recommended. In special cases of monitoring the dose to the eye lens a depth d = 3 mm may be appropriate. Landolt-Börnstein New Series VIII/4
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4 Radiological quantities and units
[Ref. p. 4-27
The operational quantities for individual monitoring meet several criteria. They are defined for all types of radiation, additive with respect to various directions of radiation incidence, take into account the backscattering from the body and can be measured with a dosemeter worn on the body. The personal dose equivalent quantities, Hp(10) and Hp(0.07), are defined in the person, in the actually existing radiation field, and are measured directly on the person. Other requirements the quantities should satisfy can, however, be fulfilled only with additional specifications. An operational quantity for individual monitoring should allow the effective dose to be assessed or should provide a conservative estimate under nearly all irradiation conditions. This, however, requires that the personal dosemeter must be worn at a position on the body which is representative with respect to the exposure. For the usual dosemeter position in front of the trunk the quantity Hp(10) mostly furnishes a conservative estimate of E even in cases of lateral or isotropic radiation incidence on the body. In cases of exposure from the back, however, a dosemeter worn at the front side and correctly measuring Hp(10), will not provide a conservative estimate of E. A further requirement for an operational quantity is that it allows dosemeters to be calibrated under reference conditions in terms of that quantity. The personal dose equivalent is defined in the individual human body and obviously individual dosemeters cannot be calibrated in front of a real human body. For calibration, the human body must therefore be replaced by an appropriate phantom. Three standard phantoms have been defined by ISO for this purpose and the definition of Hp(10) and Hp(0.07) is extended to be defined not only in the human body but also in three phantoms of ICRU tissue (see Fig. 4.2) – a slab phantom (30 cm × 30 cm × 15 cm), a wrist phantom (a cylinder of 73 mm in diameter and 300 mm in length) and a finger phantom (a cylinder of 19 mm in diameter and 300 mm in length). In reference radiation fields used for calibration, the values of the quantities in these phantoms, Hp,slab(10) and Hp,slab(0.07) etc., are defined as the true values of the corresponding Hp-quantities (see also Sect. 6.1.2 and Sect. 10.2.1).
Fig. 4.2. Phantoms of ICRU tissue for the definition of Hp-quantities for calibration of individual dosemeters. a) slab phantom b) wrist phantom c) finger phantom
4.6 Radioactivity quantities The decay of a radionuclide is a stochastic process which means that the number of decays within a fixed time interval is described by a probability distribution. The expectation value of the number of decays is determined by the decay constant which is specific for each radionuclide and energy state (mostly the decay constant for the ground state is given). The decay constant λ of a radionuclide in a particular energy state is the quotient of dP by dt, where dP is the probability that a given nucleus undergoes a spontaneous transition from that energy state in the time interval dt. It is
λ=
dP dt
unit: s−1 Landolt-Börnstein New Series VIII/4
Ref. p. 4-27]
4 Radiological quantities and units
4-19
The half-life T1/2 of a radionuclide in a particular energy state is the mean time of the radionuclide in that state to decrease to one half of their initial number of nuclei. It is T1/2 = (ln 2)/λ.
4.6.1 Activity, specific activity, activity concentration, activity per area The activity A of an amount of a radionuclide in a particular energy state at a given time is the quotient of dN by dt, where dN is the expectation value of the number of spontaneous nuclear transitions from that energy state in the time interval dt. It is A=
dN dt
unit: becquerel (Bq), 1 Bq = 1 s−1
Radionuclides are mostly included in other solid, liquid or gaseous material and the amount is quantified by the quantities specific activity and activity concentration. The specific activity as is given by the quotient of the activity A by the mass m, where A is the activity of the radionuclide in the mass m. as =
A m
unit: Bq kg−1
The activity concentration cnuclide is given by the quotient of the activity A by the volume V, where A is the activity of the radionuclide in the volume V. c nuclide =
A, V
unit: Bq m−3
For the determination of contaminations the distribution of radionuclides on surfaces is of interest. The related quantity is the activity per unit area aa defined by the quotient of the activity A by the area F, where A is the activity of a radionuclide distributed on the surface area F. aa =
A F
unit: Bq m−2, often Bq cm−2
For decontamination of a surface from deposited radionuclides it is usually important if the radionuclides are removable or if they are diffused into the surface region of the material and are fixed near the surface in the material. If an aa-value is given it should be specified if this value is related to the removable part only or to the total activity at the surface.
4.6.2 Specific quantities for radon, thoron and their progeny Radon (222Rn) and thoron (220Rn) are gaseous radionuclides in the U- and Th-decay chain, respectively, occurring naturally (see 3.4.3). Their decay products are also radionuclides but metallic. While for radon the short-lived progeny 218Po, 214Pb, 214Bi and 214Po (see Table 4.4) are important in radiation protection, the important thoron progeny are 216Po, 212Pb, 212Bi and 212Po (see Table 4.5). In air there is usually a mixture of radon/thoron and short-lived radon/thoron progeny. These progeny are mostly attached to aerosols. A few percentages of them, however, are non-attached. The progeny may be deposited in the lung where its decay by alpha-particle emission is seen to be most important for lung cancer induction. Specific quantities have been defined taking care of this situation. Landolt-Börnstein New Series VIII/4
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4 Radiological quantities and units
[Ref. p. 4-27
Table 4.4. Data of radon (222Ra) progeny (nuclear data are from [NN98]) potential alpha energy number of Radionuclide half-life atoms per Bq per atom ε per Bq ε /λ A/λ i [MeV] [10−12 J] [MeV] T1/2 [10−12 J] 218
Po Pb 214 Bi 214 Po 214
1) 2)
1 2 3 4
3.10 min 26.8 min 19.9 min
268 2 320 1 723
164 µs
1)
13.69 7.69 7.69 7.69
2.19 1.23 1.23 1.23
3 670 17 800 13 100
589 2 860 2 100
0.106 0.513 0.381
2×10−3
2.9×10−4
6×10−8
no number is given because all atoms decay in less than 1 s and a calculated number would be much less than 1. factor k is defined in Eq. (4.6.1)
Table 4.5. Data of thoron (220Ra) progeny (nuclear data are from [NN98]) potential alpha energy number of Radionuclide half-life atoms per Bq per atom ε per Bq ε /λ A/λ i [MeV] [10−12 J] [MeV] [10−12 J] T1/2 216
Po
1
0.15 s
1)
Pb Bi 212 Po
2 3 4
10.6 h 60.6 min 304 ns
55 056 5 246
212
212
1) 2) 3)
k(2)
1)
14.6 7.8 7.82) 8.8
2.34
3.32
0.51
1.25 1.25 1.25
429 000 40 900
68 710 6 554
3.8×10−6
6×10−7
k(3)
6×10−6 0.913 0.087 8×10−12
no number is given because all atoms decay in less than 1 s and a calculated number would be much less than 1. mean value from decay of 212Bi and 212Po by α-particle emission. factor k is defined in Eq. (4.6.1)
Potential alpha energy The potential alpha energy εi of an atom i in the decay chain of radon or thoron is the total energy of alpha-particles emitted during the decay of this atom to the long-living 210Pb or stable 208Pb, respectively. The potential alpha energy of N atoms of type i is N·εi. The number of atoms N per Bq is equal to A/λ, where A is the activity of this radionuclide and λ its decay constant. The potential alpha energy per Bq is then given by ε /λ (unit: J Bq−1, often used MeV Bq−1). Concentration in air The potential alpha energy concentration cp,i of a short-lived radon (or thoron) progeny in air is the sum of the potential alpha energy εi of all atoms of this progeny present in a volume V divided by this volume. It is cp,i =
N iε i ε = ci i V λi
unit: J m−3 often MeV m−3
where Ni is the number of atoms of this progeny in the volume V, ci the corresponding activity concentration and λi the decay constant. The units are related by 1 J m−3 = 6.242 × 1012 MeV m−3. The potential alpha energy concentration (PAEC) cp of any mixture of short-lived radon (or thoron) progeny in air is the sum of the potential alpha energy concentrations of all progeny in the volume considered. Landolt-Börnstein New Series VIII/4
Ref. p. 4-27] cp =
4 Radiological quantities and units
∑ c = ∑ε ⋅ c p,i
i
i
i
λi
4-21
unit: J m−3 often MeV m−3
i
Historically, for the potential alpha energy concentration the unit working level (WL) has widely been used. While originally defined as the potential alpha energy concentration associated with the radon progeny in equilibrium with 100 pCi l−1, 1 WL is now accurately fixed equal to 1.300 × 108 MeV m−3 which equals 2.08 × 10−5 J m−3. Equilibrium equivalent concentration, equilibrium factor In case of radioactive equilibrium the activity concentration of radon cRn and of its progeny are equal. This, however, is usually not the case in air. For a non-equilibrium mixture a quantity equilibrium equivalent concentration ce has been defined. The equilibrium equivalent concentration (EEC) ce corresponding to a non-equilibrium mixture of progeny in air is the fictitious activity concentration of radon in radioactive equilibrium with its shortlived progeny that has the same potential alpha energy concentration cp as the actual non-equilibrium mixture. It is always ce ≤ cRn. The SI-unit for both quantities, ce and cRn, is Bq m−3. In order to avoid confusion, the values of ce are often marked Bq m−3 (EEC). The equilibrium equivalent concentration ce can be calculated from the activity concentrations of the progeny by the equation ce =
∑k ⋅c i
with ki = (ε i λi )
i
i
∑ (ε
i
λi )
(4.6.1)
The factors ki are given in Tables 4 and 5 and it is for radon progeny: and for thoron progenies:
ce = 0.106 cPo-218 + 0.513 cPb-214 + 0.381 cBi-214 + 6 × 10−8 cPo-214 −6
ce = 7 × 10 cPo-216 + 0.913 cPb-212 + 0.087 cBi-212 + 8 × 10
−12
cPo-212
(4.6.2) (4.6.3)
Obviously, the radionuclides 216Po, 214Po and 212Po can be ignored when calculating ce because of their very low ki–values. The equilibrium factor F is defined as the quotient of the equilibrium equivalent concentration and the activity concentration of the parent nuclide, radon, in air. F = ce / cRn
(4.6.4)
The value of F ranges from 0 to 1 and is a measure to what extent radioactive equilibrium between radon and its progeny is obtained. Mostly this is not the case and often a mean value of 0.4 is convenient for the situation in homes. The unattached progeny in air which are not attached to aerosols is also of special interest. The unattached fraction fp is defined by the relative fraction of the total potential alpha energy concentration which stems from progeny in air which are not attached to aerosols. It is fp =
cpf cp
=
cpf cpa + cpf
(4.6.5)
where c pa is the potential alpha energy concentration of the progeny attached to aerosols, c pf is that of the unattached fraction and cp is the sum of both parts. Landolt-Börnstein New Series VIII/4
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4 Radiological quantities and units
[Ref. p. 4-27
Inhalation exposure of individuals The exposure of an individual to radon progeny Pp is defined as the time integral of the potential alpha energy concentration cp in air to which the individual is exposed. T
Pp (T ) = cp (t ) dt
∫
(4.6.6)
0
where T is the period of the exposure. A similar integral is given if the equilibrium concentration ce(t) is taken for integration. It is then called the equilibrium equivalent exposure Pe(T): T
Pe (T ) = ce (t ) dt
∫
(4.6.7)
0
The potential alpha energy exposure Pp is often expressed in terms of working level month (WLM), even if not recommended for further use. This quantity has been introduced especially for specifying occupational exposure and a fixed time period T of 170 hours has therefore been chosen equal to a mean monthly working time. The relation to SI-units (see Table 4.1) is given by 1 WLM = 3.54 × 10−3 J h m-3 = 2.21 × 1010 MeV h m−3.
4.7 Quantities for internal dosimetry Internal exposure means an exposure by ionising radiation emitted from radionuclides incorporated and distributed in the body. A direct measurement of doses in a human body is not possible. For internal exposure there are, therefore, no specific operational dose quantities defined. In contrast to external monitoring usually the committed tissue or organ equivalent doses and committed effective dose are determined (period of 50 y for workers and a period up to the 70th year of life for members of the public including children) and complex compartment models are used to describe the long term biokinetic behaviour of the radionuclides in the human body (see Chapt. 7). The committed tissue and organ equivalent doses of an individual are usually determined by external measurements, e.g. activity concentration of specific radionuclides in the air, specific activity of food and water or contamination of the skin, and the application of calculated dose conversion coefficients (often called dose coefficients) which have been published for inhalation, ingestion and intake through the skin for a large number of radionuclides (see Chapter 7). Measured excretion data are usually used to estimate the intake of radionuclides subsequently and then conversion coefficients are applied to evaluate doses. The intake of radionuclides by inhalation, ingestion or through intact or wounded skin or the excretion by exhalation, urine, faeces etc, is determined in terms of measurable quantities. These are often activity concentrations in air , cnuclide (in terms of Bq m-3), and inhalation frequencies or inhaled activities, the specific activity of solids and liquids, as (in terms of Bq kg-1), the amount of ingested radioactive substances or their specific activity in excretions. Further details are given in Chapter 7. In cases where radionuclides emit high energy γ-rays their distribution in the body may be determined by external measurements of γ-rays with a whole-body counter (large γ-detectors well shielded against radiation from the environment) in combination with computer codes simulating the photon absorption in the body.
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Ref. p. 4-27]
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4.8 Limits, constraints, action levels The system of radiation protection as recommended by the ICRP [91I1] is based on the following principles (see also Chapter 1): a) No practice involving exposure to radiation should be adopted unless it produces sufficient benefit to the exposed individuals or to society to offset the radiation detriment it causes (principle of justification). b) In relation to any particular source within a practice, the magnitude of individual doses, the number of people exposed, and the likelihood of incurring exposures where these are not certain to be received should all be kept as low as reasonably achievable, economic and social factors being taken into account. This procedure should be constrained by restrictions on the doses to individuals (dose constraints), or the risks to individuals in the case of potential exposures (risk constraints), so as to limit the inequity likely to result from the inherent economic and social judgements (the optimisation of protection). c) The exposure of individuals resulting from the combination of all the relevant practices should be subject to dose limits, or to some control of risk in the case of potential exposures. These are aimed at ensuring that no individual is exposed to radiation risks that are judged to be unacceptable from these practices in any normal circumstances. Not all sources are susceptible of control by action at the source and it is necessary to specify the sources to be included as relevant before selecting a dose limit (individual dose and risk limits). Generally, radiation protection takes care of exposure situations and doses which are relevant for the health of the persons involved or may not be ignored compared to the normal exposure from natural radiation sources. This means that there should exist a dose level below which exposures from radionuclides or other radiation sources may not be taken care of and where no regulations are necessary, independent of the fact that, in principle, any radiation may induce cancer or genetic defects. The ICRP sees such a dose range below a few tens of µSv (committed dose or dose per year) for a single individual which is about 1/100 of the normal exposure from natural sources in the environment. Often an upper boundary of 10 µSv (committed dose or dose per year) is used to decide if further investigations or actions are necessary. Usually the human exposures are classified in three different categories. The first is called occupational exposure which means any exposure incurred at work and principally as a result of situations which can be reasonably regarded as being in the responsibility of the operating management. It also includes potential exposures where the probability of a future exposure due to planned work forces may be estimated [97I1]. Medical exposures describe the exposure of patients during diagnostic and treatment. While medical exposures are intended to provide a direct benefit to the patient, the practice should be justified and optimized with respect to applied doses and the medical benefit. Public exposures are all exposures other than occupational or medical exposures. Public exposures include environmental exposures due to natural sources in the environment, e.g. natural actinides, radon, potassium-40 and cosmic radiation, but also those exposures due to artificial sources where the target group is the general population (details are given in Chapter 11). Examples are the broadly distributed radionuclides from the nuclear bomb test in 1950 to 1970 and the contamination due to the Chernobyl accident.
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[Ref. p. 4-27
Dose limits Dose limits have been recommended by the ICRP [91I1] for occupational exposure and for exposures to the public. They are given in terms of effective dose and few organ equivalent doses, always summed over a given period. The limits apply to the sum of the relevant doses from external exposure in the specified period (often one year) and from intakes of radionuclides in the same period. The corresponding internal dose is the 50-year committed dose (for occupational exposure) or the committed dose up to the age of 70 years (for members of the public). For public exposures the scope of these limits are restricted to doses incurred as the result of practices. Doses incurred in situations where the only protective action takes the form of an intervention are not included in system of dose limits (see action levels). Radon, thoron and their progeny in open air or in houses, natural radionuclides already in the environment and cosmic radiation on ground, are examples of those situations. While these dose limits are called primary limits, for practical reasons further limits (secondary or derived limits) are specified which are given in terms of operational or other quantities and derived from the primary limits. They are applied, for example, to define control or prohibited areas or annual limits of intake (ALI) of radionuclides. The ALI values which are specific for each radionuclide considered are also based on the committed dose for the same periods as mentioned above. The primary dose limits internationally recommended by ICRP [91I1] and the IAEA [96IA] are given below. Many, but not all countries have transferred these values into their national legislation and regulations. Some countries have either less or more restrictive regulations. As a consequence, the legal dose limits may, therefore, be different in different countries. For occupational exposure the effective dose is limited to 20 mSv per year averaged over 5 years (100 mSv in 5 years) with the further provision that the effective dose should not exceed 50 mSv within any single year. This limit avoids any deterministic effects of exposures and limits the stochastic effects to a risk level of about 10−3. For internal exposure the committed dose limit is restricted to 20 mSv in each year and the annual limit of intake (ALI) is related to this value. For women, when pregnancy has been declared, the embryo and foetus should be protected by applying for external exposure an additional equivalent dose limit to the surface of its abdomen of 2 mSv for the remaining period of the pregnancy and limiting the intake of radionuclides to 1/20 of the annual limit of intake. The detriment due to external weakly penetrating radiation mainly concerns the skin or the eye lens. In order to avoid deterministic effects the skin dose is, therefore, additionally limited to 500 mSv per year (averaged over any 1 cm2, regardless of the area exposed) and the dose to the eye lens to 150 mSv per year. For the same reason, the annual equivalent dose to the extremities (hands, feet) is also limited to 500 mSv. The approach for choosing dose limits to the public may be either based on the same ideas as for occupational exposure considering, however, the fact of the large number of persons involved or on the judgement on the existing dose level from natural radiation sources and its variation in different places where no influence on the health detriment of the population has been observed. For public exposure from sources given in practices, the ICRP has recommended a limitation of the effective dose to 1 mSv per year. In special circumstances, however, a higher value may be allowed in a single year if the average over 5 years does not exceed 1 mSv per year. The ICRP has also defined additional annual limits for the skin and the eye lenses which are 1/10 of the value recommended for workers (50 mSv averaged over any 1 cm2 of the skin and 15 mSv for the lens of the eye).
Dose constraints The control of public exposure in normal situations is usually performed by the application of controls at the different sources applying procedures of constrained optimisation and the use of prescriptive limits. A dose constraint, which is a value of individual dose from a defined source, should be used in the optimisation of protection to exclude protection options that would result in individual doses exceeding the constraint. Dose constraints are an integral part of the optimisation of protection and are thus prospectively. They are not, however, limits to be applied retrospectively.
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Ref. p. 4-27]
4 Radiological quantities and units
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The above mentioned annual limit for occupational exposure means implicitly that the dose constraint for optimisation should not exceed 20 mSv per year. It is often convenient to define a homogeneous group of persons – a critical group – which are assumed to be most highly exposed by the single source considered and to apply the dose constraint to the mean dose in that critical group. For medical exposures no dose limits have been recommended because a radiotherapy or diagnostic treatment should always provide a direct benefit to the patient. The choice of the practice and its performance should be optimized with respect to applied doses and the medical benefit. In order to characterize good medical practice and to enable quality assurance programs for use in these cases, it is helpful to define constraints or reference values based on the actual state of the art of the various investigations and procedures. Such values are especially given in the various practices of diagnostics with X-rays and radionuclides.
Action levels While for the situation of occupational exposures generally dose limits are defined, there are other situations where the only protective action takes the form of an intervention, e.g. in cases of public exposure in areas of high level of natural radiation or in areas contaminated because of former human activities or accidents like e.g. nuclear bomb testing or the Chernobyl accident. For intervention situations action levels may be defined which specify dose levels or an activity concentration in air or the specific activity in materials of the environment which are of concern with respect to public exposure. If such a level is exceeded, this should initiate measures for a reduction of the exposure and different action levels may define different measures characterising the strength of the necessary intervention. While for public exposure from sources given in practices, a dose limit of 1 mSv per year is recommended which is in the order of the natural background exposure excluding radon, action levels for initiating protection measures to the public are mostly higher, depending on the strength of the recommended measures. A typical case for the definition of action levels is the exposure by radon and its progenies. Radon is always present in the environment and may appear in higher concentrations at specific work places or in homes. For radon, however, the specification of a dose level for actions is relatively complex because the dose coefficient relating a mean radon concentration at a place to an effective dose value depends on the mean equilibrium factor F (see 4.6.2) and the mean annual time people stay at this place. Furthermore, the coefficient is mainly based on modelling and includes a large uncertainty (see Chapt. 7). Mostly the radon action levels are given in terms of Bq m−3 or Bq m−3 h. For occupational exposure with the assumption of 2000 h at work and a mean equilibrium factor of 0.4 a conversion factor of 156 Bq m−3 mSv−1 or 62 Bq m−3(EEC)mSv−1, respectively, is given [94I1]. Actions for reducing the radon concentration at a work place may be performed if the mean annual radon concentration is in the range from 500 Bq m−3 to 1500 Bq m−3 (1000 Bq m−3 corresponds to about 6 mSv per year for 2000 h at work and F = 0.4). A mean annual concentration of 3000 Bq m−3 corresponds to about 20 mSv per year which is equal to the dose limit for occupational exposure. For public exposure in dwellings a conversion factor of 58 Bq m−3 mSv−1 or 23 Bq m−3(EEC)mSv−1, respectively, is given [94I1] under the assumption that a person stays 7000 h per year in the house (with F = 0.4 in the house) and the other time in free air (with a low radon concentration). Action levels are recommended by the ICRP also for this case. They are based on the following ideas. There exists a range of normal mean radon concentrations in dwellings where no actions are necessary or useful. For existing houses with higher mean concentrations actions for reducing such values should be considered. Future houses should be designed to stay within the normal range. Because of the very different situation in the various regions regarding the natural radon concentration in the ground and hence in houses, local recommendations or regulations may differ strongly in the different countries. The ICRP has specified an upper boundary of the normal mean radon concentration in dwellings with 200-400 Bq m−3 depending on the regional situation. Landolt-Börnstein New Series VIII/4
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4 Radiological quantities and units
[Ref. p. 4-27
Another situations which may occur are emergencies where people or the government needs advice what type of actions are necessary under given or expected exposures, e.g. staying at home, avoiding to eat fresh vegetables or drink fresh milk or leaving a defined area for some time. Such a situation is also a type of intervention where action levels may be defined in national regulations like those mentioned above. Principles and more detailed information are given in ICRP Publication 63 [93I4]. For immediate emergency situations there may also dose values be given for fireman and other rescue personnel in order to restrict their risks due to an exposure. Dose levels may be defined which should not be exceeded in one case or annually or in life time.
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4-27
4.9 References 73Jac 77I1 80I1 85I1 86I1 88I1 91BI 91I1 93I1 93I2 93I3 93I4 93Se 94I1 94I2 96IA 96I1 97I1 98I1 98I2 98NN 99Ch
00I1 01I1
Jacobi, W.: The concept of effective dose - a proposal for the combination of organ doses. Radiat. Environ. Biophys. 12 (1975) 101. ICRP: Recommendations of the international commission on radiological protection. ICRP Publication 26, Ann. ICRP 1 (3) (1977). ICRU: Radiation quantities and units. ICRU Report 33, Washington, 1980. ICRU: Determination of dose equivalents resulting from external radiation sources. ICRU Report 39. Bethesda, MD: ICRU Publications, 1985. ICRU: The quality factor in radiation protection. ICRU Report 40. Bethesda, MD: ICRU Publications, 1986. ICRU: Measurement of dose equivalents from external radiation sources, Part 2. ICRU Report 43. Bethesda, MD: ICRU Publications, 1988. Bureau International des Poids et Mesures: Le Système International d´Unités (SI). 6th edition. Pavillon de Breteuil, Sevres, 1991. ICRP: Recommendations of the international commission on radiological protection. ICRP Publication 60, Ann. ICRP 21 (1-3) (1991). ICRU: Quantities and units in radiation protection dosimetry. ICRU Report 51. Bethesda, MD: ICRU Publications, 1993. International Organisation for Standardisation: ISO Standards Handbook, Quantities and Units, 3rd edition, Geneva: International Organisation for Standardisation, 1993. ICRP: Age-dependent doses to members of the public from intake of radionuclides: Part 2 Ingestion Dose Coefficients. ICRP Publication 67, Ann. ICRP 23 (4) (1993). ICRP: Principles for intervention for protection of the public in a radiological emergency. ICRP Publication 63, Edited by ICRP, 1993. Seltzer, S.M.: Calculation of photon mass energy-transfer and mass energy-absorption coefficients. Radiat. Res. 136 (1993) 147. ICRP: Protection against Radon-222 at home and at work. ICRP Publication 65, Ann. ICRP 23 (2) (1994). ICRP: Dose coefficients for intakes of radionuclides by workers. ICRP Publication 68, Ann. ICRP 24 (4) (1994). IAEA: International basic safety standards for protection against ionizing radiation an for the safety of radiation sources. Safety Series No. 115, International Atomic Agency, Vienna, 1996. ICRP: Conversion coefficients for use in radiological protection against external radiation. ICRP Publication 74, Ann. ICRP 26 (3-4) (1996). ICRP: General principles for the radiation protection of workers. ICRP Publication 75, Ann. ICRP 27 (1) (1997). ICRU: Fundamental quantities and units. ICRU Report 60. Bethesda, MD: ICRU Publications, 1998. ICRU: Conversion coefficients for use in radiological protection against external radiation. ICRU Report 57. Bethesda, MD: ICRU Publications, 1998. NNDC: Nuclear Data, Decay Radiations. National Nuclear Data Center, Brookhaven National Laboratory, Upton, NY, 1998. Chadwick, M.B., Barshall, H.H., Caswell, R.S., DeLuca, P.M., Hale, G.M., Jones, D.T.L., MacFarlane, R.E., Meulders, J.P., Schuhmacher, H., Schrewe, U.J., Wambersie, A., Young, P.G. A.: Consistent set of neutron kerma coefficients from thermal to 150 MeV for biologically important materials. Med. Phys. 26 (6) (1999) 974. ICRU: Nuclear data for neutron and proton radiotherapy and for radiation protection. ICRU Report 63. Bethesda, MD: ICRU Publications, 2000. ICRU: Determination of operational dose equivalent quantities for neutrons. ICRU Report 66, J. ICRU 1 (3) (2001).
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5 Shielding against ionizing radiation
In the shielding calculation for the radiation facility, simple dose estimation methods by using the shielding calculation constants are effective and widely used. These shielding calculation constants depend on the dose quantity to be estimated. Chapter 5 presents simple calculation methods and related constants for charged particles, photons and neutrons.
5.1 Introduction Shielding is an essential element of practical radiation protection. The necessary precautions depend especially on the kind of radiation. Charged particles as electrons and alpha particles can be absorbed in matter totally, as they have a maximum penetration depth, depending on their charge, mass, energy and on the properties of the absorbing media. A shielding slab will attenuate photons and neutrons just to a fraction of their primary intensity on the other hand, without enabling total absorption. Thus shielding calculations for charged particles mainly concern the evaluation of maximum penetration depths. As to photons an analytical calculation method has proved successful for simple geometries. It is based on the exponential law of attenuation of the unscattered photon component and a build-up portion for the scattered radiation component. The treatment of neutrons is more complicated than photon calculations. Simple methods to estimate dose rates can only be described for typical neutron sources and shielding materials, using results of more powerful shielding codes.
5.2 Stopping power and range To calculate the penetration of charged particles in matter, it is necessary to have information on the basic interactions that govern the passage through the shield. The predominant effects of protons and alpha particles are the elastic and inelastic collisions with electrons. While elastic collisions are resulting in a change of direction for the incident particle, inelastic collisions lead to energy loss and production of secondary radiation. Electrons traversing some distance in matter lose energy in numerous inelastic collisions with bound atomic electrons along their track. Furthermore Bremsstrahlung production becomes important in electron transport especially for high-Z media and high energies. The basic quantity for shielding purposes of charged particles is the stopping power, which is defined as the average energy loss per unit path length. It can be separated into the components collision stopping power Scol due to Coulomb collisions and radiative stopping power Srad due to Bremsstrahlung production. The range of charged particles is usually estimated on the basis of the continuous–slowing– down−approximation (csda). In this approximation particles are assumed to loose their energy continuously in the course of slowing down, with a fixed energy loss per path length given by the stopping power. That means energy-loss fluctuations are neglected. Integrating the reciprocal of the total
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stopping power with respect to kinetic energy E gives the csda-range r0, which is a rather good approximation of the mean path length by a particle on its course to rest. Generally the csda-range of a particle of initial energy E0 slowing down in matter to the rest energy Er is evaluated from the expression E0
r0 =
∫ [S (E )
col
+ S ( E ) rad ]−1 dE
(5.2.1)
Er
Because of the numerous scattering processes with angular deflections the csda-range is in most cases much larger than its projection on the initial direction of the particle track or the penetration depth. Tables of stopping powers and csda-ranges for protons and alpha particles are given in ICRU Report 49 [93ICR]. Corresponding values for electrons are presented in ICRU Report 35 [84ICR1], ICRU Report 37 [84ICR2], and ICRU Report 56 [97ICR]. An early review is given by Knop and Paul [64Kno].
5.3 Penetration depths of charged particles 5.3.1 Heavy charged particles Protons and alpha particles keep their initial direction rather far on their way, except near the end of the track. Because of this deviation from linear tracks, caused by multiple scattering, the average penetration depth R is a more useful quantity than the csda-range. It gives the expectation value of the distance in the initial direction of motion to that point, where the particle has slowed down to rest. The deviation from linearity of the particle track is described by the detour factor. It is practically equivalent to R/r0. The detour factor increases with increasing energy and decreasing atomic number. For this reason detour factors become important for low energies. The values are about 0.8 and 0.65 for protons and alpha particles of 1 MeV in lead, respectively. Fig. 5.3.1 gives average penetration depths R for common used materials of density ρ, estimated on the basis of the csda-range and detour factor tables of ICRU Report 49 [93ICR]. Table A5.3.1 gives the corresponding numerical values for some further materials as well (CD-ROM). As collision stopping powers at a given particle velocity are the same for all particles with the same charge number, proton ranges can be used as well for the estimation of the ranges of deuterons and positively charged pions and muons. At a given velocity the kinetic energy E of a particle with mass m is related to the kinetic energy Ep of a particle with mass mp by E=
m Ep mp
(5.3.1)
The relation between the range R(E) of particles with charge number z and the proton-range Rp(Ep) is given by R( E ) = R p (m p / m ⋅ E )
m 1 Fcorr mp z2
(5.3.2)
The correction factor Fcorr can be assumed to be unity for deuterons and positively charged pions and muons. It takes into account above all uncertainties of the particle charge. It is near unity for light ions at high particle energies. As slow ions can capture and lose electrons the effective charge may become much smaller than the nominal charge, resulting in a reduced stopping power. For alpha particles Fcorr approaches unity above 1 MeV and may increase to about 2 at lower energies. Fig. 5.3.2 gives further average penetration depths for protons and alpha particles in air on the basis of the csda-approximation. The average penetration depths for deuterons in air are calculated by Eq. (5.3.2). Table A5.3.2 gives the corresponding numerical values (CD-ROM). Landolt-Börnstein New Series VIII/4
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1
10
Average penetration depth Rρ [g cm−2]
0
10
lead -1 10 iron water -2 10 aluminum -3 10
-4 10 0.1
Fig. 5.3.1. Average penetration depth Rρ of protons in water, aluminium, iron and lead. 1
10
100
Proton energy E [MeV]
Fig. 5.3.2. Average penetration depth R of protons, deuterons and alpha particles in air. (ρ = 1.205×10−3 g/cm3).
Sample problem Average penetration depth of 3 MeV tritons in air. From Eq. (5.3.2) and Fig. 5.3.2 results: R(3 MeV) = Rp(1/3 · 3 MeV) 3/1 · 1/12 = 3 Rp(1 MeV) = 3 · 2.35 cm = 7.05 cm
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5.3.2 Electrons and positrons The range of electrons depends on the total (collision plus the radiative) stopping power. Monoenergetic electrons and beta rays suffer similar energy and angle straggling as heavy charged particles while slowing down. The deflections are however comparatively large because of their small mass. Thus the range resulting from the continuous-slowing-down-approximation (csda) is always larger than the projection of the particle track on the initial direction. The detour factor is near unit for low-Z media but reaches values up to 4 for lead. Different range definitions are used for practical purposes. In shielding calculations the maximum range Rmax is the most adequate quantity. It is defined as the depth at which the extrapolation of the tail of the transmission curve (versus electron beam axis depth) meets the Bremsstrahlung background. A frequently used numerical expression for the maximum range Rmax of beta rays of maximum beta energy Emax in matter of density ρ is given by Rmax ρ = −0.11 + 0.0121 + ( Emax / 1.92) 2
(5.3.3)
where Emax is in MeV and Rmax ρ in g cm−2 [97ICR]. The relation is useful in the energy range between 0.05 MeV and 5 MeV, for monoenergetic electrons as well. Fig. 5.3.3 gives maximum ranges for six common used materials, calculated by Eq. (5.3.3). Table A5.3.3 gives the corresponding numerical values (CD-ROM). 2
10
1
Maximum range Rmax [cm]
10
air*
0
10
glass
water
Al -1
Fe
10
Pb -2
10
-3
10
0.1
1
10
Fig. 5.3.3. Electron ranges Rmax in air (1.205×10−3), water (1.0), glass (2.23), Al (2.7), Fe (7.87) and Pb (11.35). Values in parentheses: densities in g/cm3. *Maximum range Rmax in m.
Electron energy E [MeV]
A rather simple empirical expression for the maximum range of electrons of energy E in matter of density ρ is given by Rmaxρ = E/2
(5.3.4)
where E is in MeV and Rmax ρ in g cm−2. Maximum ranges evaluated from Eq. (5.3.3) and (5.3.4) are shown in Fig. 5.3.4 together with csda-ranges for water and lead [84ICR2]. Eq. (5.3.4) turns out to be a cautious overall approximation of the maximum range overestimating for energies below 0.3 MeV and above 20 MeV. Eq. (5.3.3) provides considerable overestimation only in the energy range above 20 MeV. Table A5.3.4 gives the numerical values of the csda-ranges of air, water, Be, Al, Fe, Pb (CD-ROM). Landolt-Börnstein New Series VIII/4
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10
Electron range Rρ [gcm−2]
water(csda) lead (csda)
1
Eq. (5.3.4)
0.1
Eq. (5.3.3) 0.01 0.1
1
10
100
Fig. 5.3.4. Electron ranges Rρ evaluated from Eq. (5.3.3), Eq. (5.3.4) and by csda-approximation for water and lead.
Electron energy E [MeV]
Positrons undergo the same interactions in matter as electrons. Because of differences in collision and radiative stopping powers the positron csda-range varies between shortening at low energies and prolongation at high energies. The shortening of the positron csda-range (with respect to the corresponding electron csda-range) will amount up to 7 % at 0.1 MeV. At energies near 100 MeV the prolongation will be up to 11 % in lead and up to 2 % in water. On the assumption of positrons being annihilated before being slowed down to rest a further shortening of the positron range of up to 4 % in lead has to be considered [84ICR2]. In shielding of beta sources the range curves will yield a sufficient estimate of the necessary slab thickness. With high source activities Bremsstrahlung resulting from the deceleration of the beta particles in the material may need to be shielded as well.
5.4 Photons 5.4.1 Basic shielding concept For most gamma shielding studies photon energies of 10 keV to 10 MeV are important. In this energy range, the photoelectric effect, pair production and Compton scattering mechanisms of interaction predominate over all others. Of these three interactions, the photoelectric effect predominates at the lower photon energies; pair production is important only for higher-energy photons, while Compton scattering predominates at intermediate energies. In a few cases the shielding analyst may need to account also for coherent (Rayleigh) scattering, annihilation and fluorescence radiation. Most shielding analysis involves a study of the fluence field at pertinent locations with respect to the outside or inside shield. The purpose of such analysis is to predict the corresponding responses of some type of detector, and therefore the field information must be converted into the detector responses. These relate to the fluence by a multiplier called the detector response function. The fluence Φ of photons is the quotient of ∆N by ∆a, where ∆N is the number of photons, which enter a sphere of cross-sectional area ∆a. Detector responses for photons of interest are exposure, air kerma, absorbed dose and ambient dose equivalent.
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In this chapter the ambient dose equivalent is used in accordance with ICRP Report 74 [96ICR]. The ambient dose equivalent rates are obtained by a shielding code that satisfies the Boltzmann transport equation. Using these calculated results, a simple method to estimate the ambient dose equivalent rate for typical radioactive sources and typical shielding media is introduced.
5.4.2 Attenuation data of radioactive sources in shielding materials The radionuclides dealt with in this chapter are shown in Table 5.4.1. The value of the ambient dose equivalent rate depends on the photon energy, the geometric configuration of the source, the nature and the thickness of the shielding material. The photon fluences emitted from a radioactive source with ordinary concrete (ρ = 2.10 g/cm3), iron, lead and water as shielding materials are calculated using the BERMUDA code [92Suz, 93Suz]. Densities and elemental weight fractions of ordinary concrete (ρ = 2.10 g/cm3 and 2.35 g/cm3) are shown in Table A5.4.1 (CD-ROM). The mass attenuation coefficients used were taken from the photon library PHOTX [88DLC], which are shown in Table A5.4.2 (CD-ROM). * (t ) , where t is the thickness In this text the ambient dose equivalent rate H& * (10) is represented as H& 10 of the shield layer. The purpose of shielding analysis is to predict the corresponding responses of certain types of detectors as defined by the symbol R, and to relate R to the fluence Φ (E, t) or to the flux density φ (E, t ) of photons of energy E by a multiplier to be called the conversion coefficient; where the flux density φ (E, t ) is defined as quotient of the incremental fluence ∆Φ that occurs at a specified position and the time interval ∆ t’. The conversion coefficients of the exposure dose rate ( X& / φ ) E , of the air kerma rate ( K& / φ ) and of the ambient dose equivalent rate ( H& * / φ ) for the flux density [cm−2 s−1] are given a
E
10
E
in Table A5.4.3 (CD-ROM). While the conversion coefficients of exposure rate ( X& / φ ) E [µR h−1 cm2 s] are taken from the third column of Table A.1 in ICRU Report 47 [92ICR], the ones of the air kerma rate * ( K& a / φ ) E [nGy h−1 cm2 s] and the ambient dose equivalent rate ( H& 10 / φ ) E [nSv h−1 cm2 s] are taken from the fourth and fifth column of Table A.21 in ICRP Report 74 [96ICR]. Here, the three conversion coefficients are given in special units for convenience’ sake of calculation. Consider a point source of activity A [Bq] and a point detector P located at a distance r + t [m] from the source, as illustrated in Fig. 5.4.1, where a shield layer of thickness t [m] is placed between the source and the detector. The energy spectrum of the photon flux density at point P is represented by φ (E, t ). Then the air kerma rate and ambient dose equivalent rate at point P are represented by the following formulas.
∫ (K& φ ) ⋅ φ (E ,t ) dE (t ) = ∫ (H& φ ) ⋅ φ (E ,t ) dE
K& a (t ) = * H& 10
a
(5.4.1)
E
* 10
(5.4.2)
E
By introducing the new constant Γ10* , the ambient dose equivalent transmission factor T(t), and effective conversion coefficient f * ( E ,t ) , the ambient dose equivalent rate H& * (t ) is simply obtained. 10
0
10
* 5.4.2.1 Simple method of calculating the ambient dose equivalent rate H& 10 ( t ) for radionuclides listed in Table 5.4.1
The flux density φi of photon energy group ‘i’ emitted from a radioactive point source of activity A [Bq] is represented by the following formula, where the detector P is located at distance r [m] from the source.
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φi =
A 4π r 2
5 Shielding against ionizing radiation
5-7
(5.4.3)
Ii
Ii represents the transition yield (number of gamma rays per decay). Then, the ambient dose equivalent rate is obtained by the following summation. H& 10* =
∑ ( H&
/ φ ) i ⋅ φi
* 10
(5.4.4)
i
Constant: Γ10* With regard to Eq. (5.4.3) and Eq. (5.4.4), the constant Γ10* [µSv m2 MBq−1 h−1] is defined as the sum of components Γ10* ,i for specified photon energy groups ‘i’ of a gamma-emitting nuclide, as given in Eq. (5.4.5).
Γ10* =
∑Γ
* 10,i
=
i
1
∑ 4π I ⋅ ( H& i
* 10
/ φ ) i ⋅ 0.1
(5.4.5)
i
That means, the value of Γ10* represents the ambient dose equivalent rate H& 10* [µSv h−1] for A = 1 MBq and r = 1 m of a gamma-emitting nuclide. The values of Γ10* for 33 radionuclides are given in Table 5.4.1 [01JRIA]. With the constant Γ10* the ambient dose equivalent rate of a point source of a gamma-emitting radionuclide of activity A at distance r becomes A H& 10* = 2 Γ10* r
(5.4.6)
Ambient dose equivalent transmission factor: T(t) The ambient dose equivalent transmission factor T(t) is the quotient of the ambient dose equivalent rate * H& 10 (t ) by the ambient dose equivalent rate in the absence of shielding material H& 10* (0) . T (t ) = H& 10* (t ) / H& 10* (0)
(5.4.7)
The values of the ambient dose equivalent transmission factor T(t) for 33 radionuclides are given in Table A5.4.4 through A5.4.36 (CD-ROM) for four shielding materials, namely iron, lead, concrete (ρ = 2.10 g/cm3) and water [01Sak]. Furthermore, the transmission factors T(t) are presented in Fig. 5.4.2 through 5.38, as a function of the thickness t of shield layers [01Sak]. Using the constant Γ10* and the ambient dose equivalent transmission factor T(t), the ambient dose equivalent rate H& * ( t ) at a point detector P located a distance r + t [m] from the source, as illustrated in 10
Fig.5.5, is obtained by Eq.(5.4.8). H& 10* (t ) =
A ⋅ Γ10* ⋅ T (t ) ( r + t )2
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* 5.4.2.2 Simple calculation of ambient dose equivalent rate H& 10 ( t ) for radionuclides unlisted in Table 5.4.1
Effective (averaged) conversion coefficient: f10* ( E0 ,t ) When f10* ( E ) defines the quotient of conversion coefficient ( H& 10* / φ ) E by ( K& a / φ ) E for a specified energy E, the ambient dose equivalent rate in Eq. (5.4.2) is given by the following formula:
∫f
* H& 10 (t ) =
* 10
(E ) ⋅ (K& a
φ )E ⋅ φ (E ,t ) dE
(5.4.9)
Equation (5.4.9) averaged by air absorbed dose rate spectrum at the point P in Fig. 5.4.1 gives Eq. (5.4.10). This formula was introduced as the effective conversion coefficient f10* = f10* ( E0 , t ) for photons of primary energy E0 and for shield layer thickness t is represented by Tanaka and Suzuki [91Tan]:
f
* 10
∫f =
* 10
(E ) ⋅ (K& a
φ )E ⋅ φ (E , t ) dE
(5.4.10)
∫ (K& φ ) ⋅ φ (E ,t ) dE a
E
Then, the relationship between the ambient dose equivalent rate and the air kerma rate becomes * H& 10 ( t ) = f 10* ⋅ K& a (t )
(5.4.11)
Effective conversion coefficients f10* = f10* ( E0 , t ) for iron, lead, concrete and water are given in Table A5.4.37 through A5.4.40 for photon energies from 15 keV to 10 MeV and shield layers up to 40 mfp (CD-ROM) [01Sak].
* 5.4.2.3 Calculation method of ambient dose equivalent rate H& 10 ( t ) using exposure dose rate and effective conversion coefficient
Step 1: The flux density of uncollided photons of energy E0 for source intensity S = A · I at distance r + t behind a shield layer of thickness t is
φ 0 (E0 ) =
S exp(− µt ) 2 4π (r + t )
(5.4.12)
The mass attenuation coefficients µ are listed in Table A5.4.2 (CD-ROM). Step 2: The exposure dose rate for uncollided photons is
(
X& 0 (E0 ) = φ0 (E0 ) ⋅ X& φ
)
E0
(5.4.13)
The conversion coefficients ( X& / φ )E are listed in Table A5.4.3 (CD-ROM). Step 3: The total exposure dose rate including collided photons is X& = B ⋅ X& 0
(5.4.14)
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Exposure buildup factors B = B(E0, t ) for a point isotropic source in iron, lead, concrete (ρ = 2.35 g/cm3) and water are given in Table A5.4.41 through A5.4.44 for photon energies from 15 keV to 10 MeV and shield layers up to 40 mfp (CD-ROM), which are taken from ANSI/ANS6-4-3 [91ANS]. Step 4: The air kerma rate obtained from the total exposure dose rate is K& a = 8.764 ⋅ 10 −3 ⋅ X&
(5.4.15)
where 8.764·10–3 is the value of a typical conversion coefficient from exposure to air kerma, X& is in R h−1 and in Gy h−1. Step 5: Using the effective conversion coefficient f10* the ambient dose equivalent rate follows from Eq. (5.4.11). Values for not tabulated energies are obtained by interpolation.
5.4.3 An example of the calculation of an ambient dose equivalent rate Sample problem: 60Co source of A = 3.7⋅1013 Bq (1,000 Ci) is treated in a room enclosed by concrete wall of 100 cm thickness. Calculate the ambient dose equivalent rate at a point P on the outside wall. The calculation model is illustrated in Fig. 5.4.1, where the distance r between the source and the front of the concrete wall is 500 cm.
5.4.3.1 The method using the constant Γ10* and ambient dose equivalent transmission factor T(t ) Using the value Γ10* of 60Co in Table 5.4.1 and the one of T(t) at concrete 100 cm in Table A5.4.11 (CD-ROM), the ambient dose equivalent rate in Eq.(5.4.8) is obtained, as follows. 3.7 ⋅ 10 7 Γ10* T (t ) * (t ) = = 1.028 ⋅ 0.354 ⋅ 1.55⋅10−4 Sv h−1 = 56.4 µSv h−1 Eq. (5.4.8): H& 10 2 (5 + 1)
Values for not tabulated thicknesses of the shield layer are obtained by linear interpolation of log T(t ) – t [cm].
5.4.3.2 The method using the effective conversion coefficient f10* and the exposure dose rate conversion coefficient ( X& / φ ) Simplifying assumption: 60Co emits two gamma rays of E0 = 1.25 MeV per disintegration. Step 1: The photon flux density for the uncollided photons is: Eq. (5.4.12):
φ0 (E0 ) =
3.7 ⋅ 1013 ⋅ 2 ⋅ exp(− 12.02 ) cm−2 s−1 = 98.56 cm−2 s−1 4 ⋅ 3.14 ⋅ 6002
The linear attenuation coefficient of concrete (ρ = 2.10 g/cm3) for 1.25 MeV photons is µ = 0.1202 cm−1, which is obtained by the linear interpolation of log µ – log E, using the concrete mass attenuation coefficients of E0 = 1 MeV and 1.5 MeV given in Table A5.4.2 (CD-ROM). With t = 100 cm results µ t = 12.02.
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Step 2: Calculation of the exposure dose rate of the uncollided photons:
(
X& 0 (E0 ) = φ0 (E0 ) ⋅ X& φ
Eq. (5.4.13):
)
E0
= 98.56 ⋅ 2.182 ⋅ 10−6 R h−1 = 2.151⋅10−4 R h−1
The conversion factor ( X& / φ ) E for 1.25 MeV is obtained by linear interpolation of log ( X& / φ ) E – log E, using ( X& / φ ) for 1 MeV and 1.5 MeV given in Table A5.4.3 (CD-ROM). E
Step 3: Calculation of total exposure dose rate including collided photons: X& = B ⋅ X& 0 = 21.84 ⋅ 2.151 ⋅ 10−4 R h−1 4.698⋅10−3 R h−1
Eq. (5.4.14):
The exposure buildup factor for E0 = 1.25 MeV in infinite concrete material is obtained by the following procedure. At first, concrete buildup factors B = 27.01 and 18.36 are obtained for E0 = 1 MeV and 1.5 MeV for µ t = 12.02 mfp. They result from the linear interpolation of log B(E) – µt [mfp], using buildup factors for µ t = 10 and 15 mfp from Table A5.4.43 (CD-ROM). Next, buildup factor B = 21.84 is obtained for E0 = 1.25 MeV by the linear interpolation of log B(E) – log E, using buildup factors of E0 = 1 and 1.5 MeV at µ t = 12.02. Step 4: Conversion from exposure dose rate X& to air kerma rate K& a : K& a = 8.764 ⋅ 10 −3 ⋅ 4.698 ⋅ 10 −3 Gy h−1 = 4.117⋅10−5 Gy h−1
Eq. (5.4.15):
* Step 5: Conversion from air kerma rate K& a to ambient dose equivalent H& 10 (t ):
The effective conversion coefficient f 10* in Eq. (5.4.11) is obtained by the following procedure. At first, the effective conversion coefficients of concrete f 10* = 1.308 and 1.256 are obtained for E0 = 1 MeV and 1.5 MeV and for µ t = 12.02. They result from the linear interpolation of log f 10* – µt [mfp], using f 10* for
µ t = 10 and 15 mfp from Table A5.4.39 (CD-ROM). Next, f 10* = 1.279 is obtained for E0 = 1.25 MeV by the linear interpolation of log f 10* – log E, using effective conversion coefficients of E0 = 1 and 1.5 MeV at µ t = 12.02. H& 10* (t ) = f 10* ⋅ K& a = 1.279 ⋅ 4.117 ⋅ 10−5 Sv h−1 = 52.7 µSv h−1
Eq. (5.4.11):
* The values of H& 10 (t ) obtained by the methods in 5.4.3.1 and 5.4.3.2 agree within 7 %.
Table 5.4.1 Constant Γ10* [µSv m2 MBq−1 h−1] Radionuclide 18
F Na 51 Cr 54 Mn 59 Fe 56 Co 57 Co 60 Co 64 Cu 24
Γ10*
Radionuclide
0.166 0.492 0.00547 0.130 0.171 0.492 0.0206 0.354 0.0307
65
Zn Ga 68 Ge* 75 Se 81 Rb* 85 Kr 85 Sr 99 Mo* 99m Tc 67
Γ10*
Radionuclide
0.0847 0.0268 0.158 0.0660 0.104 0.00037 0.0826 0.0444 0.0214
103
Pd* Ag 111 In 124 Sb 123 I 125 I 131 I 133 Xe 137 Cs* 110m
Γ10*
Radionuclide
Γ10*
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5.5 Neutrons 5.5.1 Basic shielding concepts In passing through shielding material, neutrons attenuate by elastic scattering, inelastic scattering and absorption. For high-energy neutrons over 10 MeV, inelastic scattering reaction is effective to decrease energy. Therefore iron is a suitable material because of its relatively big inelastic cross section. For neutrons with energy lower than 10 MeV, materials that contain hydrogen are used to utilize its elastic scattering reaction and absorption process. Dose calculations of neutrons are more complicated than photons because secondary gamma-ray dose by neutron capture and inelastic scattering should be taken into account. Moreover, reaction type and cross section depends strongly on the neutron energy and the shielding material. Therefore, dose evaluation is generally done by using a shielding code that solves the Boltzmann transport equation. In the present chapter a simple method to estimate dose rates from typical neutron sources and shielding materials is described using results of the shielding code ANISN [73Eng].
5.5.2 Attenuation data of various neutron sources in shield material Fig. 5.5.1 shows the geometry of the transmission calculations by ANISN. A neutron point isotropic source with source intensity S = 1 s−1 is surrounded by a thick spherical shield. The JENDL3.2 [90Shi] cross-section library is used in this calculation. The elemental composition and the density of each shield material are shown in Table 5.5.1. Dose rates were estimated using the calculated neutron and secondary gamma-ray flux and the flux-to-dose conversion factor of ambient dose equivalent H ∗ (10) in ICRP Report 74 [96ICR]. Table 5.5.3 - 5.5.6 give the calculated neutron and secondary gamma-ray ambient dose rates for 252Cf, Am-Be, D-D and D-T sources [01Sak]. The first column shows the distance r [cm] from the centre, the second column shows the equivalent dose rate at distance r when no shield material is present: H& 0 [µSv h−1], the third to fifth columns show the neutron, secondary gamma-ray and total equivalent dose rates in water at distance r: H& n [µSv h−1], H& g [µSv h−1] and H& t [µSv h−1], and the sixth to eighth columns show the neutron, secondary gamma-ray, and total transmission factors: Fn (= H& n / H& 0 ), Fg (= H& g / H& 0 ) and Ft (= H& t / H& 0 ) for water. Fn, Fg, Ft values for polyethylene, ordinary concrete and heavy concrete are also described. As for the simple shielding estimation, the constant Γ for various sources is shown in Table 5.5.2, representing the bare ambient dose equivalent rate at 1 m for unit source intensity. Similar tables for the effective dose rates for AP (anterior - posterior) exposure geometry are shown in Table A5.5.1 - A5.5.4 [01Sak] of the accompanying CD-ROM. Neutron and secondary gamma-ray ambient dose equivalent rates and effective dose rates are calculated by the following equation.
Landolt-Börnstein New Series VIII/4
Ref. p. 5-32]
5 Shielding against ionizing radiation
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S H& = Γ ⋅ F (t ) 2 d
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Fig. 5.5.2 - 5.5.5 show the neutron and secondary gamma-ray dose transmission curves for a 252Cf source in water, polyethylene, ordinary concrete, and heavy concrete. As shown in these figures, secondary gamma-rays play dominant role from about 50 cm in water and polyethylene, and about 150 cm in ordinary concrete. The total transmission factor for heavy concrete is lower than that for the other materials, because it suppresses secondary gamma-rays efficiently. Fig. 5.5.6 - 5.5.9 show the transmission curves for Am-Be source in each shield material. The neutron source spectrum from (α, n)-sources depends on the grain radius and mixing ratio. This transmission factor was calculated using the source spectrum measured by Greiss [68Gre]. Transmission factors for Am-Be are similar to these for 252Cf source. Fig. 5.5.10 - 5.5.13 show the transmission curves for D-D source in each shield material. The transmission factor is low because of the low source neutron energy (2.45 MeV). D(d, p)T and D(d, n)3He reactions occur with almost the same probabilities. Therefore the ordinary D-D source produces a small amount of D-T neutrons by accumulated tritium. The D-T neutrons must be considered simultaneously. Fig. 5.5.14 - 5.5.17 show the transmission curves for D-T source in each shield material. The transmission factor is very high for D-T neutrons because of the neutron energy being high (14.1MeV). Similar figures for the effective dose rate for AP (anterior - posterior) exposure geometry are shown in Fig. 5.5.20 - 5.5.35 of the accompanying CD-ROM. These attenuation calculations were done in sufficiently thick material, called infinite geometry. This means, calculated dose rates at every point contain the backscattered components. This causes unrealistic results, so that dose attenuation factor exceeds one. Backscattered dose contributions depend on the shield material and thickness. Attention has to be paid as well to a more realistic geometry, which usually shows a certain distance between the source and the shield wall. This effect increases the transmission factor to a certain extent. Considering these effects, overestimation results at 0.5 m to about 1.4 for water and to about 1.2 for ordinary concrete. Although the attenuation factor is conservative, it is sufficient for easy evaluation. Table 5.5.1 Elemental composition and densities of shield material. *) Type 02-a concrete from ANL5800, p.660 (1963); **) From JAERI-M 6928, p.36 (1977). Material −3
Density [g cm ] Element H C O Mg Al Si Ca Fe Landolt-Börnstein New Series VIII/4
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Heavy concrete** 3.715
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Table 5.5.2 Constant Γ (ambient dose equivalent) for various neutron sources Source 252
Cf Am-Be D-D D-T
Γ [µSv h−1 m2 s] 1.11·10–5 1.13·10–5 1.19·10–5 1.49·10–5
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Landolt-Börnstein New Series VIII/4
Ref. p. 5-32]
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(1) Dose calculation at the surface of storage container Neutron dose rate The shielding effect by the 0.1 cm thick iron layer is small, therefore the iron is neglected, and polyethylene is the only shielding material to be considered. With Γ from Table 5.5.2 and the neutron transmission factor Fn(40) for polyethylene from Table 5.5.3 (t → r = 40 cm) the neutron ambient dose equivalent rate is 4.32 ⋅ 10 6 Eq. (5.5.1): H& n = 1.11·10−5 · 5.74·10−3 · µSv h−1 = 1.10 µSv h−1 0.52 Secondary gamma-ray dose rate The shielding effect by the 0.1 cm thick iron layer is small, therefore the iron is neglected, and polyethylene is the only shielding material to be considered. With Γ from Table 5.5.2 and the secondary gamma-ray transmission factor Fg(40) for polyethylene from Table 5.5.3 (t → r = 40 cm) the secondary gamma ambient dose equivalent rate is 4.32 ⋅ 10 6 Eq. (5.5.1): H& g = 1.11·10−5 · 1.26·10−2 · µSv h−1 = 2.42 µSv h−1 0.52
Also, the primary gamma-rays of 252Cf source should be considered. (2) Dose calculation outside the exposure room Neutron dose rate With Γ from Table 5.5.2 and the neutron transmission factor Fn(50) for ordinary concrete from Table 5.5.3 (t → r = 50 cm) the neutron ambient dose equivalent rate is 4.32 ⋅ 10 6 µSv h−1 = 0.299 µSv h−1 Eq. (5.5.1): H& n = 1.11·10−5 · 1.56·10−1 · 5.0 2
Secondary gamma-ray dose rate With Γ from Table 5.5.2 and the secondary gamma-ray transmission factor Fg(50) for ordinary concrete from Table 5.5.3 (t → r = 50 cm) the secondary gamma-ray ambient dose equivalent rate is 4.32 ⋅ 10 6 µSv h−1 = 0.0313 µSv h−1 Eq. (5.5.1): H& g = 1.11·10−5 · 1.63·10−2 · 5.0 2
5.5.4 Induced activity Structure materials, air, coolant waters etc. are activated in neutron fields. The induced activity has to be considered in radiation protection design of nuclear reactors, fusion experimental reactors and high energy accelerators. Examples of well-known activation reactions, half-lives, and gamma-ray energies of produced nuclides – typical for nuclear reactors − are listed in Table 5.5.7. These reaction cross sections strongly depend on neutron energy. The induced activity and gamma-ray dose rate can be estimated by the following three steps: 1) Estimation of the neutron energy spectrum with a computer code such as ANISN [73Eng]. 2) Estimation of the induced activity and gamma-ray source strength for a given irradiation and decay time with a computer code such as ORIGEN [73Bel], using the neutron energy spectrum data 3) Estimation of the gamma-ray dose at a given point with a computer code such as ANISN.
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( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+W ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)J ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)W ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
3RO\HWK\OHQH )J )W
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)Q
Table 5.5.3 Neutron and secondary gamma-ray ambient dose equivalent rates for 252Cf
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
2UGLQDU\&RQFUHWH )Q )J )W ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+HDY\&RQFUHWH )J )W
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)Q
5-20 5 Shielding against ionizing radiation [Ref. p. 5-32
Landolt-Börnstein New Series VIII/4
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
U>FP@
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+Q
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+J
)Q
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
:DWHU
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+W ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)J ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)W ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
3RO\HWK\OHQH )J )W
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)Q
Table 5.5.4 Neutron and secondary gamma-ray ambient dose equivalent rates for Am-Be
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
2UGLQDU\&RQFUHWH )Q )J )W ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+HDY\&RQFUHWH )J )W
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)Q
Ref. p. 5-32]
Landolt-Börnstein New Series VIII/4
5 Shielding against ionizing radiation 5-21
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
U>FP@
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+Q
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+J
)Q
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
:DWHU
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+W ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)J ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)W ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
3RO\HWK\OHQH )J )W
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)Q
Table 5.5.5 Neutron and secondary gamma-ray ambient dose equivalent rates for D-D
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
2UGLQDU\&RQFUHWH )Q )J )W ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+HDY\&RQFUHWH )J )W
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)Q
5-22 5 Shielding against ionizing radiation [Ref. p. 5-32
Landolt-Börnstein New Series VIII/4
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
U>FP@
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+Q
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+J
)Q
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
:DWHU
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+W ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)J ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)W ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
3RO\HWK\OHQH )J )W
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)Q
Table 5.5.6 Neutron and secondary gamma-ray ambient dose equivalent rates for D-T
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
2UGLQDU\&RQFUHWH )Q )J )W ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
+HDY\&RQFUHWH )J )W
( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( ( (
)Q
Ref. p. 5-32]
Landolt-Börnstein New Series VIII/4
5 Shielding against ionizing radiation 5-23
5-24
5 Shielding against ionizing radiation
[Ref. p. 5-32
Table 5.5.7 Typical activation reactions and cross section data, half-lives, gamma-ray energies of produced radionuclides [81Mug], [88Mcl], [02Nak]. Material Stainless steel Water
Reaction Ni(n, p)58Co
Half life 70.8 d
γ-ray energy [MeV] 0.811, 0.511
54
312 d
0.835
59
Co(n, γ)60Co 16 O(n, p)16N
5.27 y
1.17, 1.33
8.2·10−2 (Fission spectrum averaged) 3.7·101 (at 0.0253 eV)
7.13 s
6.13
2.0·10−5 (Fission spectrum averaged)
40
1.83 h
1.29
6.6·10−1 (at 0.0253 eV)
58
Fe(n, p)54Mn
Ar(n, γ)41Ar
Air 10
Cross section [barn] 1.1·10−1 (Fission spectrum averaged)
10 252
252
Cf/water
Cf/Polyethylene
–1
–1
10
–3
10
–3
10
Transmission factor
Transmission factor
10
–5
10
–7
10
Fn Fg Ft
–9
10
–7
10
Fn Fg Ft
–9
10
–11
10
–5
10
–11
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.2. Transmission factor of water for 252Cf source.
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.3. Transmission factor of polyethylene for 252Cf source.
10
10 252
252
Cf/ordinary concrete
1
Cf/heavy concrete
1 –1
10
Transmission factor
Transmission factor
–2
–2
10
–3
10
–4
10
–5
Fn Fg Ft
10
–6
10
–4
10
–6
10
Fn Fg Ft
–8
10
–7
10
10
–10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.4. Transmission factor of ordinary concrete for 252 Cf source.
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.5. Transmission factor of heavy concrete for 252 Cf source.
Landolt-Börnstein New Series VIII/4
Ref. p. 5-32]
5 Shielding against ionizing radiation 10
10
Am-Be/water
Am-Be/Polyethylene –1
1
10
–3
Transmission factor
Transmission factor
–2
10
–4
10
–6
10
Fn Fg Ft
–8
10
10
–5
10
–7
10
Fn Fg Ft
–9
10
–10
10
5-25
–11
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.6. Transmission factor of water for Am-Be source.
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.7. Transmission factor of polyethylene for Am-Be source.
10
10
Am-Be/heavy concrete
Am-Be/ordinary concrete 1
1 –1 –2
Transmission factor
Transmission factor
10
–2
10
–3
10
–4
10
Fn Fg Ft
–5
10
–4
10
–6
10
Fn Fg Ft
–8
10
–6
10
10
–10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.8. Transmission factor of ordinary concrete for Am-Be source.
Landolt-Börnstein New Series VIII/4
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.9. Transmission factor of heavy concrete for Am-Be source.
5-26
5 Shielding against ionizing radiation
10
[Ref. p. 5-32
1
D-D/water
–1
10
D-D/Polyethylene
–2
10
–4
10
–3
10
–6
10 10
Transmission factor
Transmission factor
–5 –7
10
–8
10
–10
10
–9
–12
10
10
–14
10
–11
10
–16
10
Fn Fg Ft
–13
10
–15
10
10
–20
10
–17
10
Fn Fg Ft
–18
–22
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.10. Transmission factor of water for D-D source.
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.11. Transmission factor of polyethylene for D-D source.
10
10
D-D/heavy concrete
D-D/ordinary concrete
1
1 –1
10
Transmission factor
Transmission factor
–2
–2
10
–3
10
–4
10
–5
Fn Fg Ft
10
–6
10
–4
10
–6
10
Fn Fg Ft
–8
10
–7
10
10
–10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.12. Transmission factor of ordinary concrete for D-D source.
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.13. Transmission factor of heavy concrete for D-D source.
Landolt-Börnstein New Series VIII/4
Ref. p. 5-32]
5 Shielding against ionizing radiation
5-27
10
D-T/water
1
D-T/Polyethylene
1
–1
10
Transmission factor
Transmission factor
–2
–2
10
–3
10
–4
10
–5
–6
10
Fn Fg Ft
10
–8
10
–7
10
–4
10
–6
Fn Fg Ft
10
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.14. Transmission factor of water for D-T source.
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.15. Transmission factor of polyethylene for D-T source.
1
10
D-T/heavy concrete
D-T/ordinary concrete 1
–1
10 –1
Transmission factor
Transmission factor
10
–2
10
–3
10
–3
10
–5
10
–4
10
Fn Fg Ft
–5
10
–9
10
–6
10
Fn Fg Ft
–7
10
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.16. Transmission factor of ordinary concrete for D-T source.
Landolt-Börnstein New Series VIII/4
0
50
100 150 Radius [cm]
200
250
Fig. 5.5.17. Transmission factor of heavy concrete for D-T source.
5-28
5 Shielding against ionizing radiation
[Ref. p. 5-32
5.6 Computer codes and online nuclear data services Table 5.6.1 gives a collection of typical computer codes for transport calculations of gamma-rays, neutrons and charged particles with regard to shielding problems. The programs are collected and distributed by different Data Centres, as for example NEA or NNDC. The necessary data in specified formats for program testing and evaluation are provided by the centres as well. Some Online Nuclear Data Services for basic nuclear data and evaluated nuclear data are listed in Table 5.6.2. Table 5.6.1 Computer codes for shielding and source calculations Computer Codes ANISN
Engle Jr., W.W.: A Users Manual for ANISN, A One Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering. K-1693, 1973
BERMUDA
Suzuki, T. et al.: Development of: A Radiation Transport Code System Part I. Neutron Transport Codes, JAERI 1327, JAERI, 1992 A Radiation Transport Code System Part II. Gamma Rays Transport Codes. JAERI-M 93-143, JAERI, 1993
DOORS
DOORS includes: TORT Three-dimensional neutron/photon transport DORT Two-dimensional neutron/photon transport ANISN One-dimensional neutron/photon transport Rhoades, W.A., D.B. Simpson: The TORT Three-Dimensional Discrete Ordinates Neutron/Photon Transport Code. ORNL/TM-13221, 1997
DOT-4.2
Rhoades, W.A., D.B. Simpson, R.L. Childs, W.W. Engle Jr.: The DOT-4 Two Dimensional, Discrete-Ordinates Transport Code with Space-Dependent Mesh and Quadrature. ORNL/TM-6529, 1978
DUCT-III
Tayama, R., H. Nakano, H. Handa, K. Hayashi, H. Hirayama, K. Shin, F. Masukawa, H. Nakashima, N. Sasamoto: DUCT-III, A Simple Design Code for Duct-Streaming Radiations. KEK Internal 2001-8, 2001
EGS4
Nelson, W.R., H. Hirayama, W.O. Rogers: The EGS4 Code System. SLAC265, 1985 (Electron Photon Shower Simulation by Monte-Carlo)
ETRAN
Berger, M.J., S.M. Seltzer: Electron and Photon Transport Programs – 1. Introduction and Notes on Program DATAPAC-4. NBS 9836, 1968; 2. Notes on Program ETRAN-15. NBS 9837, 1968 (Monte Carlo Code System for Electron and Photon Transport Through Extended Media)
ISO-PC
Revision of ISOSHLD Engle, R.L., J. Greenborg, N.M. Hendrickson: ISOSHLD - A Computer Code for General Purpose Isotope Shielding Analysis. BNWL-236, 1966 (X-Ray, Gamma-Ray, Bremsstrahlung Dose-Rates)
Landolt-Börnstein New Series VIII/4
Ref. p. 5-32]
5 Shielding against ionizing radiation
5-29
Computer Codes MCNP-4C
Briesmeister, J. F.: MCNP - A General Monte Carlo N-Particle Transport Code. Version 4C, LA-13709-M, 2000 (Coupled Neutron, Electron, Gamma 3-D Time-Dependent Monte Carlo Transport Calculation Code)
MORSE-CGA
Emmett, M.B.: MORSE-CGA, A Monte Carlo Radiation Transport Code with Array Geometry Capability. ORNL-6174, 1985
NAC
Weinstein, Suzanne T.: NAC - Neutron Activation Code. NASA TM X-5260, 1968
ORIGEN2
Croff, A.G.: A User’s Manual for the ORIGEN2 Computer Code. ORNL/TM-7175, 1980
PALLAS
Takeuchi, K., S. Tanaka: PALLAS-1D(VII): A Code for Direct Integration of Transport Equation in One-Dimensional Plane and Spherical Geometries. JAERI-M 84-214, 1984
PENELOPE2001
Salvat, F., J.M. Fernandez-Varea, E. Acosta, J. Sempau: PENELOPE, A Code System for Monte Carlo Simulation of Electron and Photon Transport. Proceedings of a Workshop/Training Course, OECD/NEA 5-7 November, 2001, NEA/NSC/DOC, 2001
PUTZ
Ingersoll, D.T.: User's Manual for PUTZ - A Point-kernel Photon Shielding Code. ORNL/TM-9803, 1986
QAD-CGGP
Sakamoto, Y., S. Tanaka: QAD-CGGP2 and G33-GP2- Revised Versions of QAD-CGGP and G33-GP Codes with Conversion Factors from Exposure to Ambient and Maximum Dose Equivalents. JAERI-M 90-110, 1990
RAID
Moore, J.A., J.B. Eggen, F.O. Leopard: Monte Carlo Procedure for Analysis of Radiation in Ducts (RAID). AFWL-TR 67-9, 1967 (Gamma, Neutron Scattering in Cylindrical or Multibend Ducts)
SAM-CE
Steinberg, H.A. et al.: SAM-CE - A Monte Carlo Code for Three Dimensional Neutron, Gamma Ray and Electron Transport (Revision 5). MR-7052-5, 1977
SKYSHINE
Lampley, C.M., M.C. Andrew, M.B. Wells: The SKYSHINE-III Procedure: Calculation of the Effects of a Structure Design on Neutron, Primary Gamma-Ray and Secondary Gamma-Ray Dose Rates in Air. RRA-T8209A, 1988
SRNA-2KG
Ilic, R.D.: SRNA, Protons Transport Simulation by Monte Carlo Techniques User's Guide. Version 2KG, 2001
Landolt-Börnstein New Series VIII/4
5-30
5 Shielding against ionizing radiation
[Ref. p. 5-32
Table 5.6.2 Online Nuclear Data Services and Code Services Address
Information
Atomic Mass Data Center (AMDC) http://csnwww.in2p3.fr/AMDC/
Atomic Mass Data, Q-Values
Gesellschaft für Schwerionenforschung (GSI) http://www.gsi.de/
Links to Data Banks, Literature-Research
International Atomic Energy Agency (IAEA) Nuclear Data Center http://www-nds.iaea.or.at
Nuclear Structure and Decay Data, Cross Section Data (Photons, Neutrons, charged Particles) (ENSDF, ENDF), Nuclear Science References
Nuclear Data Center Japan Atomic Energy Research Institute http://wwwndc.tokai.jaeri.go.jp/
Chart of Nuclides, Evaluated Nuclear Data Library, Tables of Nuclear Data
Korea Atomic Energy Research Institute (KAERI) Nuclear Structure and Decay Data, Nuclear Data Evaluation Lab Photon Cross Section Data http://atom.kaeri.re.kr Lawrence Berkeley National Laboratory (LBNL) Isotopes Project http://isotopes.lbl.gov
Nuclide-Table, Nuclide-Chart, Isotope Explorer, Internet Isotope Explorer
Los Alamos National Laboratory (LANL) T-2 Nuclear Information Service http://t2.lanl.gov/data/decayd.html
Nuclear Structure and Decay Data, Cross Section Data (Photons, charged Particles, thermal Neutrons) (ENSDF, ENDF) Nuclide-Chart, Nuclear Data Viewer
Lunds Universitet LUND Nuclear Data Service http://nucleardata.nuclear.lu.se/nucleardata
Nuclear Structure and Decay Data (ENSDF), Literature, References, Isotope Explorer, Internet Isotope Explorer
www-tech.mit.edu/Chemicool
Periodical System of Elements
Nuclear Energy Agency (NEA) http://www.nea.fr/html/dbdata/
Nuclear Data, Computer Codes, Experimental Nuclear Reaction Data Retrievals, Evaluated Nuclear Data Retrievals, Bibliographical Research
National Institute of Standards and Technology (NIST) http://physics.nist.gov/PhysRefData/contents.html
Physical Constants, X-ray and gamma ray data, X-Ray Attenuation and Absorption for Materials of Dosimetric Interest, XCOM: Photon Cross Sections Database, Stopping-Power and Range Tables for Electrons, Protons, and Helium Ions, Radionuclide Half-life Measurements Made at NIST, Atomic Weights and Isotopic Compositions
Landolt-Börnstein New Series VIII/4
Ref. p. 5-32]
5 Shielding against ionizing radiation
5-31
Address
Information
National Nuclear Data Center (NNDC) http://www.nndc.bnl.gov
Nuclear Structure and Decay Data, Neutron Cross Sections, Online-Service, Literature, References, Documentation of Data Banks and Computer Codes (ENSDF, ENDF,...), Nuclear Science References
PhysicsWeb http://physicsweb.org/TIPTOP/paw/
Information, Links
Radiation Safety Information Computational Center (RSICC) http://www-rsicc.ornl.gov/rsicc.html
Codes and Data, Newsletter, Workshops
Triangle Universities Nuclear Laboratory (TUNL) http://www.tunl.duke.edu/NuclData
Nuclear Data for light Nuclides (A = 3 to 20)
Department of Computer Science University of Columbia (UBC) http://www.cs.ubc.ca/elements/periodic-table
Periodical System of Elements
US Nuclear Data Program (USNDP) http://www.nndc.bnl.gov/usndp/
Links to Nuclear Data Banks
WebElements http://www.webelements.com/
Periodical System of Elements
Landolt-Börnstein New Series VIII/4
5-32
5 Shielding against ionizing radiation
5.7 References 64Kno 68Gre 73Bel 73Eng 74Jae 81Mug 84ICR1 84ICR2 84Mug 88DLC 88Mcl 90Shi 91ANS 91Tan 92ICR 92Suz 93Suz 93ICR 95Hub
96ICR
Knop, G., Paul,W.: Interaction of electrons and α-particles with matter; Alpha-, beta- and gamma-Ray spectroscopy, Vol. 1, Siegbahn, K. (ed.), North-Holland Publishing Company, 1964. Greiss, H.B.: Nukleonik 10 (1968) 283. Bell, M.J.: ORIGEN-The ORNL isotope generation and depletion code, ORNL-4628, 1973. Engle, jr., W.W.: A user manual for ANISN; A one dimensional discrete ordinates transport code, ORNL-TM-4280, 1973. Jaeger, R.G., Hübner, W.: Dosimetrie und Strahlenschutz, G. Thieme Verlag, 1974, p. 378. Mughabghab, S.F., Divadeenam, M., Holden, N.E.: Neutron cross sections, Vol 1, Neutron Resonance Parameters and Thermal Cross Sections, Part A, Z=1-60. New York: Academic Press, 1981; NNDC: Online-Datenbank: http://www.nndc.bnl.gov/nndc/ensdf/ensdfindex.html International Commission on Radiation Units and Measurements, ICRU Report 35, Radiation dosimetry: Electron beams with energies between 1 and 50 MeV, ICRU Publications, 1984. International Commission on Radiation Units and Measurements, ICRU Report 37, Stopping powers for electrons and positrons, ICRU Publications, 1984. Mughabghab, S.F.: Neutron cross sections, Vol 1, Neutron resonance parameters and thermal cross sections, Part B, Z=61-100. New York: Academic Press, 1984; NNDC: OnlineDatenbank: http://www.nndc.bnl.gov/nndc/ensdf/ensdfindex.html Radiation Shielding Information Center Data Package DLC-136/PHOTX, Photon Interaction Cross Section Library, contributed by National Institute of Standards and Technology, 1988. McLane, V., Dunford, C.L., Rose, P.F.: Neutron Cross Sections, Vol. 2. Boston: Academic Press, 1988. Shibata, K., Nakagawa, T., Asami, T., Fukahori, T., Narita, T., Chiba, S., Mizumoto, M., Hasegawa, , Kikuchi, Y., Nakajima, Y., Igarasi, S.: “Japanese Evaluated Nuclear Data Library, Version-3, -JENDL3-, JAERI1319” Japan Atomic Energy Research Institute, 1990. American National Standard for Gamma-Ray Attenuation Coefficients and Buildup Factors for Engineering Materials, ANSI/ANS-6.4.3-1991, 1991. Tanaka, T., Suzuki, T.: A calculation method of photon dose equivalent based on the revised technical standards of radiological protection, ORNL/TR-90/29, Oak Ridge National Laboratory, 1991. International Commission on Radiation Units and Measurements, ICRU Report 47, Measurements of dose equivalents from external photon and electron radiations, ICRU Publications, 1992. Suzuki, T., Hasegawa, A., Tanaka, S., Nakashima, H.: Development of BERMUDA: A radiation transport code system, Part I: Neutron transport codes, JAERI 1327, Japan Atomic Energy Research Institute, 1992. Suzuki, T., Hasegawa, A., Tanaka, S., Nakashima, H.: Development of BERMUDA: A radiation transport code system, Part II: Gamma rays transport codes, JAERI-M 93-143, Japan Atomic Energy Research Institute, 1993. International Commission on Radiation Units and Measurements, ICRU Report 49, Stopping powers and ranges for protons and alpha particles, ICRU Publications, 1993. Hubbell, J.H., Seltzer, S.M.: Tables of X-ray mass attenuation coefficients and mass energyabsorption coefficients from 1 keV to 20 MeV for Elements Z=1 to 92 and 48 additional substances of dosimetric interest. Online: http://physics.nist.gov/PhysRefData/Xcom/Text/XCOM.html International Commission on Radiological Protection, Publication 74: Conversion coefficients for use in radiological protection against external radiation, Ann. ICRP 26, No.3/4 (1996).
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International Commission on Radiation Units and Measurements, ICRU Report 56, Dosimetry of external Beta-rays for radiation protection, ICRU Publications, 1997. The Japan Radioisotope Association, Radioisotope Pocket Data Book (Revised Edition 10), 2001 (in Japanese). Sakamoto, Y., Endo, A., Tsuda, S., Takahashi, F., Yamaguchi, Y.: Shielding calculation constants for use in effective dose evaluation for photons, neutrons and bremsstrahlung from Beta-ray, JAERI-Data/Code 2000-044, Japan Atomic Energy Research Institute, 2001, (in Japanese). Nakagawa, T, Kawasaki, H., Shibata, K. (Eds.):Curves and tables of neutron cross sections in JENDL-3.3, Part I (Z = 1 - 50) and Part II (Z = 51 - 100), JAERI-Data/Code 2002-020, Japan Atomic Energy Research Institute, 2002.
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6 External dosimetry
This chapter introduces the main protection and operational quantities in external dosimetry and describes the anthropomorphic models used for their calculations. Conversion coefficients i.e. mean organ equivalent doses normalised to the measurable quantity “air kerma free-in-air” are given for idealized geometries representing occupational exposures and for environmental source geometries.
6.1 Protection and operational quantities 6.1.1 Protection quantities The International Commission on Radiological Protection (ICRP) for more than 50 years supports a system for radiological protection, based on concepts, quantities, and basic recommendations. The concept of radiation protection is based on the justification, optimisation and limitation of radiation exposure. This concept includes a dose limitation system for occupational and man-made environmental radiation exposures to ensure that the radiation risk would not exceed reasonable limits. The most recent set of protection quantities recommended in ICRP60 [91ICR] includes the organ or tissue equivalent doses HT and the effective dose E (see Chap. 4). These quantities are not measurable but can be calculated if the exposure conditions are known. The quantity to be limited in radiation protection of occupationally exposed persons and members of the public is the effective dose E which is the weighted mean of equivalent doses of several organs and tissues of the body that are considered to be most sensitive. E=
∑w H = ∑w ∑D T
T
T
T
T
T ,R
wR
R
where HT is the mean organ equivalent dose and wT is the tissue weighting factor (with Σ wT = 1) which takes into account the differences in the stochastic radiation risk of the different organs. It is derived from the mean organ absorbed dose DT, i.e. the total amount of energy deposited in an organ (or tissue) T per mass of the organ, by multiplying with a radiation weighting factor wR reflecting the relative biological effectiveness of the radiation incident on the body or emitted from radionuclides in the body. The sensitive organs and tissues together with their respective tissue weighting factors wT were defined in ICRP Publication 60 [91ICR] (see Chap. 4).
6.1.2 Operational Quantities The International Commission on Radiation Units and Measurements (ICRU) has defined a set of operational quantities for area and individual monitoring [85ICR, 92ICR1, 93ICR] in response to the recommendations of the International Commission on Radiological Protection [77ICR] which were Landolt-Börnstein New Series VIII/4
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designed to provide an estimate of the protection quantities defined by ICRP and to serve as calibration quantities for dosimeters used in monitoring. For area monitoring, the appropriate operational quantities are the ambient dose equivalent H*(d), and the directional dose equivalent, H'(d,Ω), both defined at a depth d, on the principal axis of the 30 cm diameter ICRU sphere. The recommended value of d for strongly penetrating radiation is 10 mm and for weakly penetrating radiation it is 0.07 mm (see Sect. 4.5.3.3). For individual monitoring, the quantity “personal dose equivalent” Hp(d) was proposed, which is the dose equivalent in soft tissue, at an appropriate depth d below a specified point on the body [92ICR1, 93ICR]. For weakly penetrating radiation, depths of 0.07 mm for the skin and 3 mm for the eye lens are used, denoted by Hp(0.07) and Hp(3), respectively; for strongly penetrating radiation, a depth of 10 mm is currently recommended by the ICRU, denoted by Hp(10) (see Sect. 4.5.3.4). Personal dose equivalent is defined in the human body and may, therefore, vary between individuals; furthermore, the depth d is specified but the position of the point below which it is defined is not fixed but only correlated to the position of the dosemeter worn on the body. Consequently, the personal dose equivalent can be expected to vary also between locations on any given individual and is, hence, anticipated to be a multi-valued quantity [96ICR, 98ICR, 99Zan]. To make this quantity single-valued in a given exposure situation, both a particular location on the human body and a particular phantom of the body need to be specified for evaluation. For calibration purposes, “surrogate quantities” for Hp(d) have been introduced: it is recommended that personal dosimeters normally worn on the trunk are calibrated on an ICRU tissue slab or PMMA (polymethylmethacrylate) slab with dimensions 30 × 30 × 15 cm3 [92ICR1]. Conversion coefficients for personal dose equivalent at the relevant depths d in the ICRU tissue slab Hp,slab(d) have been calculated for calibration purposes [91Gro, 95ISO, 95Til] and have been recommended for use [96ICR, 98ICR, 98Cla]. The operational quantities used in measurement were designed to provide a reasonable estimate of the appropriate protection quantity. For external exposures of the body in a given field, it is desirable that the ratio of the value of the appropriate protection quantity to the value of the corresponding operational quantity is less than unity, i.e. the operational quantity should always provide a “conservative” estimate of the protection quantity. More about the definitions of the operational quantities can be found in Chap. 4.
6.2 Dosimetric models 6.2.1 Models and phantoms of the human body To estimate the protection quantities organ and tissue equivalent doses HT there are two approaches, an experimental and a theoretical one. The experimental determination is very difficult whereas the mathematical modelling of an exposure has been proved to be extremely flexible and powerful. For this purpose, a series of computer models of the human body were designed in the past, together with computer codes simulating the radiation transport and energy deposition in the body. The computer models used for the representation of the human body in dose calculations can range from simple geometric forms such as spheres, cylinders or slabs to complex representations of detailed anatomical features. Such complex models, used since 1966 for the estimation of organ doses are the socalled mathematical phantoms, which are models whose body organs and tissues are described by mathematical expressions representing planes or cylindrical, conical, elliptical or spherical surfaces. The mostly used model was the “MIRD” one, named after the initials of the Medical Internal Radiation Dose Committee of the US Society of Nuclear Medicine where it was initially developed [69Sny, 78Sny]. From this, several paediatric models were derived to represent infants and children of various ages, for example those from Cristy [80Cri]. As an improvement to these hermaphrodite models, separate male and female
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adult mathematical models have been introduced by Kramer et al. [82Kra] called Adam and Eva. For these models, the organ masses and volumes are in accordance with the ICRP data on Reference Man [75ICR]; in addition to a separation of the gender-specific organs, the phantom Eva is smaller than Adam, according to the difference in size of the male and female Reference Man. The oesophagus, an organ which had not originally been defined in these models, was incorporated in the form of an elliptical cylinder ranging in height from within the neck down to the top of the stomach and lying in front of the spine, slightly shifted to the left side [92Zan]. Adam and Eva contain all organs and tissues relevant for the evaluation of effective dose, with only few exceptions: since there is no specific representation of the “bone surface”, the skeleton is modelled as a homogeneous mixture of all skeletal constituents, i.e., hard bone, bone marrow and certain peri-articular tissues. Commonly, the dose to this representation of the entire skeleton is taken to represent the dose to the bone surface. Although there may be certain differences, these are usually considered to be small in view of the small weighting factor (wT=0.01) assigned to this tissue. The muscles were represented by that part of the body volume not attributed to any other organ or tissue of the models. More recently, four models representing the adult female, non-pregnant and at 3 stages of pregnancy were elaborated by Stabin et al [99Sta]. A comprehensive review of models and phantoms of the human body can be found in ICRU Report 48 [92ICR2]. The term “model” refers to computational models, whereas “phantom” implies either a physical phantom or a computational one. Spherical and slab phantoms are convenient and simple approximations of the human body. A spherical model of 30 cm diameter made of ICRU tissue-equivalent material (see Sect. 4.5.3.3) is used for the definition of the operational quantities. Various tissue substitutes are available for fabrication of corresponding physical phantoms, including tissue-equivalent material, water and perspex. For calibration purposes, slab tissue-substitute phantoms of 30 × 30 × 15 cm3 are used. Recently a new generation of computational phantoms has become available which offer the prospect of increased realism and accuracy in dose calculations. These models use computed (CT) or magnetic resonance (MR) tomographic data of real persons to provide three-dimensional representations of the human body and comprise a large number of volume elements (voxels) all of the same size but with differing composition according to the organ to which they belong. The GSF-National Research Centre in Germany started since the mid eighties the development of voxel models covering various ages and anatomical statures [88Zan, 01Zan, 02Pet]. Due to their anatomical realism, such models have been the subject of increasing interest and acceptance, and others have been developed also elsewhere [94Zub, 96Dim, 00Xu]. Both MIRD-type and voxel models incorporate different densities and atomic compositions for the various body tissues. The number of organs simulated varies from model to model, however, the latest versions include all organs defined to be important. Fig. 1 shows views of selected organs of the mathematical model Eva [82Kra] and the adult female voxel model Donna, developed at GSF [02Pet].
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Fig. 6.1. View of selected organs of the mathematical model Eva [82Kra] and the adult female voxel model Donna, developed at GSF [02Pet].
6.2.2 Idealized geometries representing occupational exposures To simulate occupational exposure conditions, whole-body irradiation with idealised geometries are conventionally taken into account. These include broad parallel beams and fully isotropic radiation incidence. The directions of incidence for the parallel beams considered are: anterior-posterior (AP), posterior-anterior (PA), left lateral (LLAT), right lateral (RLAT) and a full 360° rotation around the phantoms’ longitudinal axis (ROT). Although these geometries are idealised, they may be taken as acceptable approximations to actual conditions of exposure. The AP, PA and both lateral geometries are supposed to approximate radiation fields from single sources and particular body orientations. The ROT geometry approximates the exposure of a person who moves randomly in the field of a single source irradiating at right angles to the longitudinal axis of the body. The fully isotropic (ISO) source simulates the geometry of a body suspended in a large cloud of radioactive gas.
6. 2. 3 Environmental source geometries For external exposures to environmental sources the dosimetric quantities of interest are the radiation doses received by the radiosensitive organs and tissues of the body due to photons and electrons emitted by radionuclides distributed in soil and air. The radiation dose depends strongly on the temporal and spatial distribution of the radionuclide to which a human is exposed. The situation of radioactive release in water is more rare and is not covered here. The kinds of radiation of concern are those sufficiently Landolt-Börnstein New Series VIII/4
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penetrating to traverse the overlying tissues of the body and which deposit their energy in organs and tissues of the body. Penetrating radiations are limited to photons and electrons. Neutrons from cosmic radiation are not dealt here. For simulating the exposure to environmental gamma-rays, the following three typical cases of environmental sources are considered here: (1) semi-infinite volume source in the air; (2) infinite plane source in the ground; (3) semi-infinite volume source in the ground. The first source configuration models the gaseous radioactive release into the atmosphere at locations which are not too near to the release point, by assuming a homogeneous contamination of the air up to a height of 1000 m above a smooth airground interface. The second source simulates the deposition of radionuclides in the ground, by assuming an infinite plane source in the soil. The third source simulates the natural radioactivity in the ground (the dominant radionuclides of the 238U series, the 232Th series and 40K) being homogeneously distributed to a depth of 1 m in the soil.
6.2.4 Methods of calculating protection quantities in computational models Today the predominant method for assessment of absorbed doses in the body is the application of Monte Carlo methods to simulate the transport of radiation in the body. The organ and tissue doses are then estimated in the form of conversion coefficients giving organ doses per unit of a measurable quantity. The Monte Carlo method is a computational model in which physical quantities are calculated by simulating the transport of particles. In the computer program, single particles are followed through their histories of inelastic and elastic scattering or absorption within the anthropomorphic model. Depending on their energy and on the material they are passing through, the particles interact differently and each mode of interaction has a certain probability of occurring, which can be selected by appropriate use of random numbers and probability distributions. Individual particles have different energies, directions and path lengths modelled randomly from probability distributions. By averaging over large numbers of random paths, good estimates of the quantities of interest can be made. Basic elements of Monte Carlo simulation include the choice of random number generator which provides the method of sampling the cross-section data and coordinate transformations from probability distributions. The mean absorbed dose in a defined volume of material is computed from the incident and emerging energies from the volume by dividing the energy imparted by the mass of the volume material. Most codes dealing mainly with photon interaction assume that electrons generated through different interactions are absorbed “on the spot”. The energy transferred at a point of inelastic photon interaction is then modelled as being deposited at that point, without considering the energy transport by secondary charged particles (“kerma approximation”). This approach is valid as long as there is approximate secondary charged particle equilibrium, which can be supposed in most cases due to the macroscopic approach considering mean organ and tissue doses. This is acceptable for photon energies up to 3 MeV. For the skeleton, however, the boundary effects do have an impact on the tissue dose and corrections for secondary electron effects in the skeleton have to be applied. For neutrons up to 20 MeV the kerma approximation introduces no significant error due to the short range of the recoil protons and heavier charged particles. For the estimation of absorbed doses distributions in the body, several transport codes are used and their description is beyond the scope of this book. General Monte Carlo codes are available, such as EGS4 [85Nel, 94Hir], ITS [92Hal] and MCNP4 [91Bri]. Various other research institutes have also developed their Monte Carlo codes like the GSF-National Research Center code [82Kra, 89Vei], PTB/PG [86Gro, 94Gro], etc. The anthropomorphic models used for the collection of the data shown below were the mathematical Adam and Eva and the mathematical MIRD hermaphrodite for photons and neutrons; for electrons, the ICRU tissue sphere and slab as well as Adam and Eva were employed.
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A joint task Group of ICRP and ICRU reviewed the conversion coefficients reported by different researchers. Conversion coefficients for external photon, neutron and electron exposures and the idealised geometries AP, PA, LLAT, RLAT and ROT were evaluated and compiled to be used as “reference data”; they can be found at the ICRP Report 74 [96ICR] and ICRU Report 57 [98ICR].
6.3 Conversion coefficients for photons 6.3.1 Occupational This Section provides organ doses in the form of conversion coefficients, i.e. mean organ equivalent doses normalised to “air kerma free-in-air” which is a measurable quantity; the conversion coefficients are given in the unit Sv·Gy−1. No location for a measurement of the normalisation quantity has to be specified since for the parallel and ideally isotropic geometries the photon fluence is invariant throughout the field. The organ equivalent dose conversion coefficients were calculated for the male Adam and female Eva models [82Kra] separately. Average organ equivalent dose conversion coefficients were computed as the arithmetic mean of those for the male and female models. The gonad equivalent dose conversion coefficients are the arithmetic mean values of the respective coefficients for testes and ovaries. For these calculations, the GSF Monte Carlo code was used which computes the dose deposited by photons from an external or internal source in various sections of a different media model of the body. The code is based on the fractional photon technique and uses the kerma approximation. The latter is valid only when charged particle equilibrium is established which can be supposed in most cases due to the moderate differences of the photon cross sections for the tissues in the human body and the macroscopic approach of energy deposition. There are two exceptions where boundary effects do have an impact on the tissue dose. One is the red bone marrow, where a moderate increase in energy deposited in the marrow cavities is expected from increased photoelectron emission from the surrounding bone. For photon exposures this effect was accounted for by applying appropriate correction factors [69Spi, 97Zan] to the energy deposited to the red bone marrow calculated using the kerma approximation. The other tissue where boundary effects could be of consequence is the bone surface, a very thin soft tissue layer enveloping the bones. Here secondary electron equilibrium is not valid for energies below approximately 300 keV as there the bone cross section values are considerably higher than those for soft tissues, resulting in an increased production of secondary electrons in the bones and, consequently, a dose enhancement at the interface between the bones and the adjacent soft tissues compared to the dose to tissue beyond the range of these secondary electrons. This enhanced dose to the tissue adjacent to bones is, however, not as high as the mean dose to the homogeneous mixture of skeletal tissues [68Dre]. Consequently this can be taken as a conservative estimate of the dose to the bone surface in this photon energy range. Above 300 keV, the cross sections of bone and soft tissue per mass density have a similar magnitude, and approximate secondary electron equilibrium is established. To the calculated values of the conversion coefficients for monoenergetic photons, a fitting procedure using cubic spline functions was applied. With these fitting functions, values were also evaluated for 200 photon energies distributed equidistantly on a logarithmic scale between 10 or 15 keV and 10 MeV. Figures 6.2-6.8 show the conversion coefficients as a function of photon energy for 8 selected organs and tissues and for AP, PA, LLAT, RLAT, LAT, ROT and ISO geometries, respectively. These are the average values evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model. The complete sets of organ equivalent doses for the male, female and average can be found in Zankl et. al. [97Zan] shown graphically as well as in tabular form. The average values of the male and female models denoted as “adult” are also adopted as reference values for conversion coefficients and are presented in detail in ICRP Report 74 [96ICR] and ICRU Report 57 [98ICR] for those specific organs for which the ICRP recommends tissue weighting factors (see Chap. 4, Table 3).
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Organ equivalent dose per unit air kerma [Sv/Gy]
The energy dependence of the conversion coefficients for single organs is determined by the photon interaction cross sections in tissues, the location of the organ in the body and the irradiation geometry. The cross sections decrease with increasing photon energy and the conversion coefficients correspondingly increase due to the increasing range of photons in the body. With further increasing energy and range of the photons, the conversion coefficients decrease. This leads to more or less pronounced peak in the energy range between 80-100 keV. The more pronounced peak of the conversion coefficients of skeleton is due to the large values of the ratio of the attenuation coefficients of bone and air respectively. Table 6.1 shows the effective dose per unit air kerma as a function of energy for the six irradiation geometries, calculated for adults using the models Adam and Eva. The different forms of the energy dependence of the conversion coefficients for effective dose with irradiation geometry result from the different locations of the organs relative to the incoming photon beam and the value of their tissue weighting factors. As it can be seen from Table 6.1, the conversion coefficients of E for AP irradiation are always higher than the corresponding ones for other irradiation geometries. This is due to the fact that for AP photon incidence most organs with large tissue weighting factors are anteriorly located.
gonads
Parallel AP
lungs RBM stomach liver thyroid skin skeleton
2
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.2. Organ equivalent doses per unit air kerma free-in-air for selected organs in AP irradiation as a function of photon energy, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model;[97Zan].
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Organ equivalent dose per unit air kerma [Sv/Gy]
3 gonads
Parallel PA
lungs RBM stomach liver thyroid skin skeleton
2
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.3. Organ equivalent doses per unit air kerma free-in-air for selected organs in PA irradiation as a function of photon energy, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
Organ equivalent dose per unit air kerma [Sv/Gy]
2 gonads lungs RMB stomach liver thyroid skin skeleton
Parallel LLAT
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.4. Organ equivalent doses per unit air kerma free-in-air for selected organs in LLAT irradiation as a function of photon energy, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
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Organ equivalent dose per unit air kerma [Sv/Gy]
2 gonads
lungs RBM stomach liver thyroid skin
Parallel RLAT
skeleton
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.5. Organ equivalent doses per unit air kerma free-in-air for selected organs in RLAT irradiation as a function of photon energy, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
Organ equivalent dose per unit air kerma [Sv/Gy]
2 gonads
lungs RBM stomach liver thyroid skin
Parallel LAT
skeleton
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.6. Organ equivalent doses per unit air kerma free-in-air for selected organs in LAT irradiation as a function of photon energy, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
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gonads
Parallel ROT
lungs RBM stomach liver thyroid skin
2
skeleton
1
0 0.01
0.1
Photon energy [MeV]
1
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Organ equivalent dose per unit air kerma [Sv/Gy]
Fig. 6.7. Organ equivalent doses per unit air kerma free-in-air for selected organs in ROT irradiation as a function of photon energy, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
gonads
Parallel ISO
lungs RBM stomach liver thyroid skin
2
skeleton
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.8. Organ equivalent doses per unit air kerma free-in-air for selected organs in ISO irradiation as a function of photon energy, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
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Table 6.1. Effective dose E per unit air kerma free-in-air Ka for monoenergetic photons and various irradiation geometries. Data are from Zankl et. al. [97Zan], calculated using the male (Adam) and female (Eva) model. Photon energy [MeV] 0.010 0.015 0.020 0.030 0.040 0.050 0.060 0.070 0.080 0.100 0.150 0.200 0.300 0.400 0.500 0.600 0.800 1.000 2.000 4.000 6.000 8.000 10.000
E/Ka [Sv·Gy−1] AP 0.00654 0.0402 0.122 0.416 0.787 1.104 1.306 1.405 1.431 1.392 1.255 1.172 1.091 1.055 1.035 1.024 1.010 1.002 0.992 0.992 0.993 0.991 0.990
PA 0.00248 0.00590 0.0183 0.129 0.372 0.641 0.847 0.968 1.020 1.031 0.960 0.916 0.881 0.872 0.870 0.871 0.875 0.881 0.901 0.918 0.924 0.928 0.929
LLAT 0.00173 0.00550 0.0156 0.0907 0.242 0.406 0.529 0.599 0.630 0.643 0.622 0.616 0.616 0.624 0.636 0.648 0.671 0.692 0.758 0.813 0.836 0.850 0.860
RLAT 0.00172 0.00551 0.0151 0.0782 0.204 0.344 0.454 0.521 0.553 0.570 0.551 0.548 0.556 0.570 0.585 0.600 0.627 0.651 0.728 0.796 0.827 0.846 0.860
ROT 0.00326 0.0154 0.0463 0.191 0.427 0.661 0.828 0.924 0.961 0.960 0.893 0.854 0.824 0.814 0.812 0.814 0.821 0.831 0.871 0.909 0.925 0.934 0.941
ISO 0.00271 0.0123 0.0362 0.144 0.326 0.511 0.642 0.720 0.749 0.748 0.700 0.679 0.664 0.667 0.675 0.685 0.703 0.719 0.774 0.824 0.846 0.859 0.868
Figures 6.8 and 6.9 demonstrate the age dependence for effective dose, by showing the effective dose for adults as well as for 0-, 1-, 5-, 10-, and 15-year old children for AP and ROT geometry respectively. The data of this figure stem from Yamagushi [94Yam] who used hermaphrodite paediatric and adult mathematical models of different body sizes developed by Cristy [80Cri]. It can be seen that the smaller body size results in higher organ dose conversion coefficients, and consequently higher effective doses, particularly at low photon energies. The largest variation of effective dose with age was found for the LAT and ISO geometries. Similarly, Zankl et. al. [97Zan] have shown that the conversion coefficients for the female model Eva are approximately 2 % to 20 % higher than those for the male model, depending on photon energy, due to the slightly smaller body size of the female model. For AP irradiation, the lung dose conversion coefficients for the female model are between 5 % and 20 % lower than those for the male, as for this geometry the lungs of the female model are partially shielded by the breast. Furthermore, some differences were observed for the organ conversion coefficients calculated by different authors and are mainly due to the different human models used in the calculations and occur at low photon energies: the female model Eva, used for the calculations of Zankl et. al. has a smaller body size than the hermaphrodite model used by Yamaguchi, resulting in higher organ conversion coefficients particularly for low energies. Consequently, differences (up to 20 %) between the effective dose coefficients from adults were observed. For energies above 70 keV there was general agreement within the statistical uncertainties.
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Parallel AP
1
0 year 1 year 5 years 10 years 15 years adult
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.9. Effective dose per unit air kerma free-in-air for AP irradiation geometry calculated for MIRD-type hermaphrodite phantoms of various ages; [94Yam].
Effective dose per unit air kerma [Sv/Gy]
Parallel ROT 1
0 year 1 year 5 years 10 years 15 years adult
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.10. Effective dose per unit air kerma free-in-air for ROT irradiation geometry calculated for MIRD-type hermaphrodite phantoms of various ages; [94Yam].
Conversion coefficients for the operational quantities ambient dose equivalent and directional dose equivalent are shown in Table 6.2. These are values recommended by the ICRP [92ICR1] and stem from calculations by several groups using Monte Carlo methods on the ICRU sphere assuming electronic equilibrium. Landolt-Börnstein New Series VIII/4
Ref. p. 6-42]
6 External dosimetry
6-13
Table 6.2. Conversion coefficients for air kerma free-in-air Ka, directional dose equivalent H´(0.07,0°), and ambient dose equivalent H*(10), per unit fluence of monoenergetic photons; [92ICR1]. Photon energy Ka/Φ H´(0.07,0°)/Φ H* (10)/Φ [MeV] [pSv cm2] [pGy cm2] [pSv cm2] 0.010 7.60 7.20 0.061 0.015 3.21 3.19 0.83 0.020 1.73 1.81 1.05 0.030 0.739 0.90 0.81 0.040 0.438 0.62 0.64 0.050 0.328 0.50 0.55 a 0.060 0.292 0.51 0.080 0.308 0.53 0.100 0.372 0.61 0.150 0.600 0.89 0.200 0.856 1.20 0.300 1.38 1.80 0.400 1.89 2.38 0.500 2.38 2.93 0.600 2.84 3.44 0.800 3.69 4.38 1.0 4.47 5.20 1.5 6.12 6.90 2.0 7.51 8.60 3.0 9.89 11.1 4.0 12.0 13.4 5.0 13.9 15.5 6.0 15.8 17.6 8.0 19.5 21.6 10.0 23.2 25.6 a
H´(0.07,0°) is not accurately determined at energies above 60 keV since there is no electronic equilibrium. Comparing tables 6.1 and 6.2, it can be seen that for photons with energies up to 10 MeV and irradiation geometries AP, PA and ROT, the operational quantity H*(10) always overestimates the protection quantity E, i.e. E/H*(10) < 1.
6.3.2 Conversion coefficients for environmental gamma ray fields 6.3.2.1 Calculation of doses for monoenergetic photons As already mentioned above, for simulating the exposure to environmental gamma-rays, the following three typical cases of environmental sources are considered to be representative: (a) semi-infinite volume source in the air; (b) infinite plane source in the ground; (c) semi-infinite volume source in the ground. The first source configuration models the gaseous radioactive release into the atmosphere at locations which are not too near to the release point, by assuming a homogeneous contamination of the air up to a height of 1000 m above a smooth air-ground interface. The second source simulates the deposition of radionuclides in the ground, by assuming an infinite plane source in the soil. The source is shielded by a soil slab of 0.5 g cm−2, allowing for some surface roughness and initial migration into soil with precipitation. The third source simulates the natural radioactivity in the ground (e.g. radionuclides of the 238 U series, the 232Th series and 40K) being homogeneously distributed to a depth of 1 m in the soil. Landolt-Börnstein New Series VIII/4
6-14
6 External dosimetry
[Ref. p. 6-42
In source (a), the dominant gamma rays come almost isotropically from the upper 2π directions, while only a small amount of scattered gamma-rays comes from the lower 2π directions. Source (c) shows the inverse tendency: the angular distribution is nearly uniform in the lower 2π directions with small components of scattered gamma-rays stemming from the upper 2π directions. In source (b), quite a large portion of the gamma-rays comes from horizontal directions. When the source distribution in the environment varies from the three typical source distributions, the angular and energy distributions also change. To estimate the organ doses from environmental photon sources presented in this book, a three-step procedure [91Pet] was followed: (1) Calculation of the gamma-ray transport in the environment (monoenergetic gamma-rays and natural radionuclides); (2) Simulation of a secondary source around the phantom; (3) Calculation of organ doses due to the secondary sources. The result of this procedure, is a set of dose conversion coefficients for monoenergetic photons. Using those, and considering the energies and intensities of the radiations emitted during nuclear transformations of these nuclides, conversion coefficients for specific radionuclides can be computed to relate a measurable quantity i.e. activity concentration or air kerma to the non-measurable quantities of organ dose. The photon transport in the environment was simulated with the Monte Carlo code YURI [85Sai], a code specially developed for environmental problems. Compton scattering, photoelectric absorption and pair production were considered as photon interaction processes. Air and ground were assumed to contact each other with an infinite plane. The cross Sections used were from Storm and Israel [70Sto]. Air was assumed to have a constant density of 1.2×10−3 g cm−3, corresponding to a temperature of 20 ºC and an air pressure of 0.1 MPa and to consist of N2, O2 and Ar having weight fractions of 75.5 %, 23.2 % and 1.3 %, respectively. Soil was taken to consist of SiO2, Al2O3, Fe2O3 and H2O with weight fractions of 58.3 %, 16.7 %, 8.3 % and 16.7 %, respectively. A soil density of 1 g cm−3 has been assumed in the calculations, since this value represents reasonably well the upper 2 cm of soil. It should be noted that the environmental transport calculations were performed without the presence of the phantom; however, the perturbation caused by the human body was investigated and found to be insignificant. From the transport calculations in the environment, double differential fluences, currents (i.e. fluences multiplied by the cosine of the angle of incidence) and air kerma values are obtained for points from 0 to 2 m above ground in steps of 20 cm. Table 6.3 shows calculated values of the air kerma rate free-in-air at 1 m height above the ground per unit activity concentration for a semi-infinite volume source in air and per unit activity per area for an infinite plane source in the ground; Table 6.4 shows the air kerma free-inair at 1 m height above the ground per disintegration/kg for the semi-infinite volume source in the ground due to the natural radionuclides. Table 6.3. next page Table 6.4. Calculated air kerma at 1 m height above the ground per disintegration/kg for a semi-infinite volume source in the ground due to natural radionuclides; [95Sai]. Radionuclides Air kerma per unit source intensity [Gy / (disintegration/kg)] 238 U series 1.29×10−13 232 Th series 1.68×10−13 40 K 1.16×10−14
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42]
6 External dosimetry
6-15
Table 6.3. Calculated air kerma rate at height 1 m above the ground per unit of activity concentration for a semi-infinite volume source in air and per unit of the activity per unit area for an infinite plane source in the ground; [95Sai]. Volume source in air Plane source in ground Energy Air kerma rate per unit Air kerma rate per unit [MeV] of activity concentration of activity per unit area [(Gy s-1)/ (Bq m-3)] [(Gy s-1)/ (Bq m-2)] −15 0.015 1.47×10 8.06×10−19 −15 0.020 7.72×10−18 1.71×10 −15 0.030 2.12×10 2.63×10−17 −15 0.040 2.40×10 3.59×10−17 −15 0.050 2.81×10 4.14×10−17 −15 0.060 3.31×10 4.65×10−17 −15 0.070 3.79×10 5.35×10−17 0.080 4.36×10−15 5.98×10−17 −15 0.100 5.55×10 7.54×10−17 −15 0.150 8.68×10 1.21×10−16 −14 0.200 1.20×10 1.68×10−16 −14 0.300 1.87×10 2.61×10−16 −14 0.500 3.21×10 4.34×10−16 −14 0.700 4.56×10 5.90×10−16 −14 1.000 6.58×10 8.09×10−16 −14 1.500 9.08×10 1.13×10−15 −13 2.000 1.32×10 1.41×10−15 −13 3.000 1.96×10 1.91×10−15 −13 6.000 3.85×10 3.19×10−15 −13 10.000 6.26×10 4.81×10−15 These height-dependent double differential (with respect to angle of incidence and photon energy) gamma ray fields were then incorporated into the organ dose calculation with anthropomorphic models, by establishing a secondary cylindrical source around the model to simulate the gamma-ray fields after the results of the transport calculation in the environment (step 2 of the procedure mentioned above) [91Pet]. The anthropomorphical models are standing on the soil modelled as a planar air/ground interface. Scatter and absorption of the radiation in both air and ground was considered in the calculation. The Monte Carlo code used for the transport calculation in the body was the GSF code mentioned above. The interactions considered were photoelectric absorption, Compton scattering and pair production and the cross section data were taken from ORNL [83Rou]. Dose conversion coefficients for Adam and Eva (see previous Section) and several organs, including the critical ones, were estimated and can be found in Zankl et. al. [97Zan]. From the dose conversion coefficients of the sex-specific models Adam and Eva, conversion coefficients for an “adult” were derived as arithmetic average. In Fig. 6.10 dose conversion coefficients are given for some selected organs for submersion in a radioactive cloud (volume source in air) and in Fig. 6.11 for surface contamination. The conversion coefficients are for an adult and are expressed as equivalent doses normalised to air kerma free-in-air at height 1 m above the ground in Sv Gy−1 as a function of photon energy. For the volume source in the ground, conversion coefficients for the photon energy distributions corresponding to the natural radionuclides of the decay series of 238U, 232Th and 40K are tabulated in Table 6.5.
Landolt-Börnstein New Series VIII/4
6-16
6 External dosimetry
[Ref. p. 6-42
Table 6.5. Organ equivalent dose conversion coefficients for the natural radionuclides for an adult, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan]. Organ
Adrenals Bladder Brain Colon Eye lenses Gonads Kidneys Liver Lungs Muscle Oesophagus Pancreas Red bone marrow Skeleton Skin Small intestine Spleen Stomach Thymus Thyroid Effective dose
Organ equivalent dose per unit air kerma free-in-air at 1 m above ground [Sv Gy−1] 238 232 40 U series Th series K 0.589 0.617 0.634 0.648 0.681 0.692 0.689 0.715 0.727 0.627 0.655 0.659 0.872 0.876 0.947 0.682 0.681 0.738 0.674 0.700 0.700 0.658 0.684 0.692 0.709 0.732 0.740 0.737 0.761 0.767 0.608 0.635 0.638 0.600 0.627 0.662 0.656 0.680 0.687 0.770 0.792 0.764 0.849 0.863 0.861 0.620 0.651 0.652 0.646 0.699 0.703 0.660 0.671 0.684 0.675 0.753 0.733 0.659 0.776 0.731 0.672 0.695 0.709
For the environmental irradiation geometries, the dependence of the organ equivalent dose conversion coefficients on photon energy is much more uniform than for the unidirectional geometries considered for occupational radiation exposures, and depends less on the position of the organ in the body. As the radiation comes from all directions, every organ is quasi deep-lying relative to at least a considerable part of the incoming photons. The conversion coefficients for the female model were found to be up to 5 % higher than those for the male model, due to the slightly smaller body size of the female model. Considering the two different source types, it can be seen that the equivalent dose conversion coefficients for the volume source in air are generally lower than those for the plane source in the ground. This results from the different angular distribution of the radiation impinging on the body: the gamma-ray field from a source in the air is nearly isotropic with respect to directions from the upper hemisphere, while the incident directions of the gamma rays from a plane source have strong horizontal bias, and most photons come from horizontal directions. Since the human body standing vertically has a reduced shielding effect for photons coming from horizontal directions, this leads to the higher doses resulting from this geometry. However, in most cases, the differences in the conversion coefficients were found to be less than 20 %. Saito et. al. [98Sai] investigated the variation of effective dose for environmental gamma-rays for source distributions other than these three typical ones and for a lying posture further to the standing one. The change of posture of a human body and the biases of environmental sources were found to affect the effective dose by some tens percent. A similar trend is anticipated for the individual organ doses. Therefore, it could be concluded that the conversion coefficients for the three typical environmental sources can be used as a reference set of values to derive the organ doses and effective doses of adults from air kerma or source activity obtained by measurement for a variety of environmental source configurations.
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42]
6 External dosimetry
6-17
Organ equivalent dose per unit air kerma [Sv/Gy]
2 gonads
lungs RBM stomach liver thyroid skin
Volume source in air
skeleton effective dose
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.11. Organ equivalent doses per unit air kerma at 1 m above the ground for some selected organs of an adult for a volume source in air, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
Organ equivalent dose per unit air kerma [Sv/Gy]
2
gonads
lungs RBM stomach liver thyroid skin
Plane source in ground
skeleton effective dose
1
0 0.01
0.1
Photon energy [MeV]
1
10
Fig. 6.12. Organ equivalent doses per unit air kerma at 1 m above the ground for some selected organs of an adult for plane source in ground, evaluated as the arithmetic mean of those for the male (Adam) and female (Eva) model; [97Zan].
Landolt-Börnstein New Series VIII/4
6-18
6 External dosimetry
[Ref. p. 6-42
6.3.2.2 Calculation of doses for radionuclides Kerma rates in air and equivalent dose rates in organs for radionuclides are obtained from dose conversion coefficients for the monoenergetic sources hT,i by multiplication with the yield yiN (in number of photons Bq−1) of photons with energy i per disintegration and summing over the photon energies of the emission spectrum of radionuclide N: gN =
∑y
N i
⋅ hT,i
i
Kerma rates in air calculated with Monte Carlo methods in the environment due to monoenergetic photon sources distributed exponentially in the soil or homogeneously in the air are given in Tables 6.3 and 6.4 respectively. Radionuclide-specific results are given in Table 6.6. Table 6.6. Kerma-rates in air at 1 m above ground per unit activity per unit area (nGy h-1 per kBq m-2) and per activity concentration in air; [94ICR]. Radionuclide Source in soil Volume Radionuclide Source in soil Volume source in air source in air at a depth of at a depth of 0.5 g cm−2 0.5 g cm−2 Kerma rate Kerma rate Kerma rate Kerma rate −1 −1 −1 [(nGy h )/ [(nGy h−1)/ [(nGy h )/ [(nGy h )/ −3 (kBq m )] (kBq m−3)] (kBq m−2)] (kBq m−2)] Be-7 Nb-93m 1.67 × 10−1 1.29 × 10−2 1.10 × 10−2 6.19 × 10−4 0 −1 0 Na-22 Nb-95 6.97 × 10 2.49 × 10 5.11 × 10 1.80 × 10−1 1 −1 −1 Na-24 Nb-95m 1.10 × 10 2.40 × 10 9.72 × 10 1.59 × 10−2 −1 −2 0 K-40 Nb-97 4.62 × 10 2.21 × 10 3.78 × 10 1.52 × 10−1 −1 −2 −2 K-42 Mo-93 8.53 × 10 7.58 × 10 6.52 × 10 3.54 × 10−3 0 −1 −1 Sc-46 Mo-99 6.31 × 10 4.85 × 10 4.75 × 10 3.43 × 10−2 −1 −3 −1 Cr-51 Tc-99m 1.09 × 10 3.93 × 10 7.06 × 10 2.64 × 10−2 0 1 0 Mn-54 Ru-103 2.69 × 10 1.64 × 10 1.97 × 10 1.08 × 10−1 0 −1 0 Mn-56 Ru-105 5.05 × 10 2.45 × 10 3.96 × 10 1.82 × 101 0 −1 −2 Fe-59 Rh-103m 3.64 × 10 8.82 × 10 2.81 × 10 5.00 × 104 0 −1 −1 Co-56 Rh-105 9.86 × 10 2.62 × 10 8.39 × 10 1.73 × 10−2 Co-57 Rh-106 3.88 × 10−1 6.90 × 10−1 2.53 × 10−2 4.75 × 10−2 0 −1 0 Co-58 Ag-110m 3.17 × 10 8.76 × 10 2.28 × 10 6.14 × 10−1 0 −1 −2 Co-60 Ag-111 7.59 × 10 8.91 × 10 5.90 × 10 5.90 × 10−3 0 −1 −1 Ni-65 Sn-117m 1.65 × 10 5.52 × 10 1.31 × 10 3.41 × 10−2 0 −1 −1 Zn-65 Sn-126 1.82 × 10 1.92 × 10 1.39 × 10 1.18 × 10−2 0 −2 0 Zn-69m Sb-124 1.41 × 10 5.67 × 10 9.58 × 10 4.25 × 10−1 0 −2 0 Se-75 Sb-125 1.30 × 10 1.48 × 10 8.68 × 10 9.97 × 102 0 −1 0 Br-84 Sb-126 4.90 × 10 9.07 × 10 4.18 × 10 6.66 × 101 −1 −2 0 Rb-86 Sb-127 2.94 × 10 2.20 × 10 2.25 × 10 1.61 × 10−1 0 −1 1 Sr-92 Sb-128 4.01 × 10 1.01 × 10 3.14 × 10 7.20 × 101 0 −1 0 Y-90m Sb-129 2.10 × 10 4.58 × 10 1.42 × 10 3.36 × 10−1 −2 −4 1 Y-91 Sb-130 1.10 × 10 1.04 × 10 8.50 × 10 7.63 × 10−1 0 −1 −1 Y-91m Te-123m 1.77 × 10 4.96 × 10 1.23 × 10 3.16 × 10−2 −1 −2 −1 Y-92 Te-125m 7.91 × 10 1.87 × 10 5.94 × 10 9.07 × 10−3 −1 −2 −2 Y-93 Te-127 2.69 × 10 1.65 × 10 2.06 × 10 1.11 × 10−3 Zr-95 Te-127m 2.40 × 100 5.91 × 10−2 1.73 × 10−1 2.78 × 10−3 −1 −2 −1 Te-129 Zr-97 2.07 × 10 6.06 × 10 4.14 × 10 1.38 × 10−2 Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide
Te-129m Te-131m Te-132 Te-133m Te-134 I-129 I-130 I-131 I-132 I-133 I-134 I-135 Cs-134 Cs-134m Cs-136 Cs-138 Ba-137m Ba-139 Ba-140 La-140 La-141 La-142 Ce-141 Ce-143 Ce-144 Pr-145 Nd-147 Pm-148 Pm-148m Pm-149 Pm-151 Eu-152 Eu-152m Eu-154 Eu-155 Eu-156 Hf-181
Landolt-Börnstein New Series VIII/4
6 External dosimetry Source in soil at a depth of 0.5 g cm−2 Kerma rate [(nGy h−1)/ (kBq m−2)] 1.65 × 10−1 4.51 × 100 8.05 × 10−1 5.79 × 100 2.84 × 100 1.14 × 10−1 7.05 × 100 1.29 × 100 7.35 × 100 2.01 × 100 8.26 × 100 4.79 × 100 5.09 × 100 1.08 × 10−1 6.75 × 100 6.96 × 100 1.98 × 100 1.45 × 10−1 6.21 × 10−1 6.93 × 100 1.27 × 10−1 6.57 × 100 2.44 × 10−1 9.28 × 10−1 6.37 × 10−2 8.32 × 10−2 4.71 × 10−1 1.78 × 100 6.55 × 100 1.23 × 10−3 1.09 × 100 3.53 × 100 9.47 × 10−1 3.85 × 100 1.88 × 10−1 3.98 × 100 1.79 × 100
Volume source in air Kerma rate [(nGy h−1)/ (kBq m−3)] 9.04 × 103 3.22 × 10−1 5.15 × 10−2 5.36 × 10−1 2.03 × 10−1 6.01 × 10−3 4.93 × 101 8.68 × 10−2 5.26 × 10−1 1.40 × 101 6.05 × 10−1 3.67 × 10−1 3.64 × 10−1 5.90 × 10−3 5.08 × 10−1 5.44 × 10−1 1.40 × 10−1 9.18 × 10−3 4.14 × 10−2 5.44 × 10−1 9.97 × 10−3 6.26 × 10−1 1.59 × 10−2 6.34 × 10−2 4.36 × 10−3 3.11 × 10−3 3.11 × 10−2 1.35 × 10−1 4.64 × 10−1 2.36 × 10−3 6.88 × 10−2 1.82 × 10−1 6.91 × 10−2 2.89 × 10−1 1.22 × 10−2 3.09 × 10−1 1.25 × 10−1
Radionuclide
Ta-182 W-187 Pb-210 Pb-212 Bi-212 Ra-224 Ra-226 Ac-228 Th-228 Th-231 Th-232 Th-234 Pa-233 U-232 U-234 U-235 U-236 U-237 U-238 Np-237 Np-238 Np-239 Pu-236 Pu-238 Pu-239 Pu-240 Pu-242 Am-241 Am-242 Am-242m Am-243 Cm-242 Cm-243 Cm-244 Cm-245 Cm-247
6-19 Source in soil at a depth of 0.5 g cm−2 Kerma rate [(nGy h−1)/ (kBq m−2)] 4.00 × 100 1.70 × 100 2.71 × 10−2 4.74 × 10−1 3.38 × 10−1 3.33 × 10−2 2.19 × 10−2 3.05 × 100 1.43 × 10−2 1.31 × 10−1 7.89 × 10−3 3.33 × 10−2 7.37 × 10−1 1.06 × 10−2 9.47 × 10−3 5.55 × 10−1 7.90 × 10−3 4.80 × 10−1 8.33 × 10−3 1.42 × 10−1 2.05 × 100 6.02 × 10−1 1.08 × 10−2 1.06 × 10−2 4.38 × 10−3 9.25 × 10−3 8.56 × 10−3 1.12 × 10−1 6.76 × 10−2 3.25 × 10−2 1.81 × 10−1 9.38 × 10−3 4.47 × 10−1 8.87 × 10−3 3.10 × 10−1 1.03 × 100
Volume source in air Kerma rate [(nGy h−1)/ (kBq m−3)] 2.99 × 10−1 1.09 × 10−1 1.51 × 10−3 3.24 × 10−2 4.36 × 10−2 2.17 × 10−3 1.46 × 10−3 2.19 × 10−1 8.75 × 10−4 7.02 × 10−3 4.54 × 104 2.07 × 10−3 4.54 × 10−2 7.24 × 10−4 5.80 × 10−4 3.34 × 10−2 5.44 × 10−4 3.11 × 10−2 6.88 × 10−4 8.42 × 10−3 1.31 × 10−1 3.78 × 10−2 7.06 × 10−4 6.19 × 10−4 2.41 × 10−4 5.83 × 10−4 5.04 × 10−4 7.96 × 10−3 4.39 × 10−3 1.61 × 10−3 1.14 × 10−2 6.05 × 10−4 2.97 × 10−2 5.62 × 10−4 2.07 × 10−2 7.16 × 10−2
6-20
6 External dosimetry
[Ref. p. 6-42
6.4 Conversion coefficients for neutrons Significant radiation exposures to neutrons occur primarily at workplaces and not in the environment. In the natural environment neutrons are found mainly in secondary cosmic ray fields [00Pel, 02Roe]; accidental exposures to neutron emitting radionuclides in clouds or on the soil are extremely unlikely and therefore not considered here. Neutron fields are in practice mixed radiation fields of wide neutron energy range, almost always associated with photons. To obtain the conversion coefficients for such fields, appropriate averaging of coefficients over the relevant spectra should be performed. The calculation of deposition of energy at any point in a body resulting from external exposure in mixed fields is a complex process of summation over all primary and secondary particle deposition. Several authors calculated for incident neutrons the protection quantities organ absorbed dose and effective dose using anthropomorphic phantoms such as the hermaphrodite MIRD-5 phantom or the sexspecific MIRD-type phantoms Adam and Eva (see Sect. 6.2 and 6.3). For the operational quantities, the ICRU sphere and slab phantoms were used. Several Monte Carlo codes were applied such as the MCNP [91Bri], SAM-CE [79Lic], MORSE-CG [75Emm], the JAERI code [Yam93], the PTB-code [90Hol] etc. Extensive tables of organ dose conversion coefficients data are given in ICRP Report 74 [96ICR] and ICRU Report 57 [978ICR] derived as sets of “best estimates” of the data of various authors. Table 6.7 lists the effective dose per unit of neutron fluence for idealized whole-body irradiation geometries and for energies ranging from thermal up to 180 MeV. In the same table, the coefficients for ambient dose equivalent are also given. Table 6.7. Effective dose per unit neutron fluence E/Φ for monoenergetic neutrons incident in various geometries on an adult anthropomorphic computational model. The last column of the table shows the coefficients for ambient dose equivalent; [978ICR]. Energy E/Φ [pSv cm2] H*(10)/Φ [MeV] AP PA RLAT LLAT ROT ISO [pSv cm2] 5.24 3.52 1.36 1.68 2.99 2.99 6.60 1.0 × 10−9 6.55 4.39 1.70 2.04 3.72 2.89 9.00 1.0 × 10−8 7.60 5.16 1.99 2.31 4.40 3.30 10.6 2.5 × 10−8 9.95 6.77 2.58 2.86 5.75 4.13 12.9 1.0 × 10−7 11.2 7.63 2.92 3.21 6.43 4.59 13.5 2.0 × 10−7 12.8 8.76 3.35 3.72 7.27 5.20 13.6 5.0 × 10−7 13.8 9.55 3.67 4.12 7.84 5.63 13.3 1.0 × 10−6 14.5 10.2 3.89 4.39 8.31 5.96 12.9 2.0 × 10−6 15.0 10.7 4.08 4.66 8.72 6.28 12.0 5.0 × 10−6 15.1 11.0 4.16 4.80 8.90 6.44 11.3 1.0 × 10−5 15.1 11.1 4.20 4.89 8.92 6.51 10.6 2.0 × 10−5 −5 14.8 11.1 4.19 4.95 8.82 6.51 9.90 5.0 × 10 −4 14.6 11.0 4.15 4.95 8.69 6.45 9.40 1.0 × 10 −4 14.4 10.9 4.10 4.92 8.56 6.32 8.90 2.0 × 10 −4 14.2 10.7 4.03 4.86 8.40 6.14 8.30 5.0 × 10 14.2 10.7 4.00 4.84 8.34 6.04 7.90 1.0 × 10−3 14.4 10.8 4.00 4.87 8.39 6.05 7.70 2.0 × 10−3 15.7 11.6 4.29 5.25 9.06 6.52 8.00 5.0 × 10−3 18.3 13.5 5.02 6.14 10.6 7.70 10.5 1.0 × 10−2 23.8 17.3 6.48 7.95 13.8 10.2 16.6 2.0 × 10−2 29.0 21.0 7.93 9.74 16.9 12.7 23.7 3.0 × 10−2 38.5 27.6 10.6 13.1 22.7 17.3 41.1 5.0 × 10−2 47.2 33.5 13.1 16.1 27.8 21.5 60.0 7.0 × 10−2
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42]
6 External dosimetry
Energy [MeV] 1.0 × 10−1 1.5 × 10−1 2.0 × 10−1 3.0 × 10−1 5.0 × 10−1 7.0 × 10−1 9.0 × 10−1 1.0 × 100 1.2 × 100 2.0 × 100 3.0 × 100 4.0 × 100 5.0 × 100 6.0 × 100 7.0 × 100 8.0 × 100 9.0 × 100 1.0 × 101 1.2 × 101 1.4 × 101 1.5 × 101 1.6 × 101 1.8 × 101 2.0 × 101 3.0 × 101 5.0 × 101 7.5 × 101 1.0 × 102 1.3 × 102 1.5 × 102 1.8 × 102 2.0 × 102 a Not available
E/Φ [pSv cm2] RLAT LLAT 16.4 20.1 21.2 25.5 25.6 30.3 33.4 38.6 46.8 53.2 58.3 66.6 69.1 79.6 74.5 86.0 85.8 99.8 129 153 171 195 198 224 217 244 232 261 244 274 253 285 261 294 268 302 278 315 286 324 290 328 293 331 299 335 305 338 324 naa 358 naa 397 naa 433 naa 467 naa 501 naa 542 naa
AP 59.8 80.2 99.0 133 188 231 267 282 310 383 432 458 474 483 490 494 497 499 499 496 494 491 486 480 458 437 429 429 432 438 445
PA 41.3 52.2 61.5 77.1 103 124 144 154 175 247 308 345 366 380 391 399 406 412 422 429 431 433 435 436 437 444 459 477 495 514 535
6-21
ROT 34.8 45.4 54.8 71.6 99.4 123 144 154 173 234 283 315 335 348 358 366 373 378 385 390 391 393 394 395 395 404 422 443 465 489 517
ISO 27.2 35.2 42.4 54.7 75.0 92.8 108 116 130 178 220 250 272 282 290 297 303 309 322 333 338 342 345 343 naa naa naa naa naa naa naa
H*(10)/Φ [pSv cm2] 88.0 132 170 233 322 375 400 416 425 420 412 408 405 400 405 409 420 440 480 520 540 555 570 600 515 400 300 285 260 245 250 260
6.5 Conversion coefficients for electrons 6.5.1 Occupational exposure Unshielded whole body irradiation by monoenergetic electrons does not represent a practical situation in occupational exposures and evaluated absorbed doses for electron beams are still sparse. However, irradiation of the skin, the lens of the eye and other superficial organs are of concern in radiological protection for electron energies below 10 MeV because the electron range is small, varying from 50 µm to about 5 cm for electron energies from 60 keV to 10 MeV. Table 6.8 shows some conversion coefficients for organ absorbed doses determined with the MCNP-4 code for the MIRD-type phantoms Adam and Eva, for monoenergetic electrons in the energy range of 100 keV to 10 MeV, incident in the AP geometry Landolt-Börnstein New Series VIII/4
6-22
6 External dosimetry
[Ref. p. 6-42
(Schultz and Zoetelief, data from [96ICR]). ICRU Report 43 contains dose distributions in anthropomorphic phantoms resulting from irradiation by electrons of energies between 5 and 46 MeV [88ICR]. Various workers performed Monte Carlo calculations with different codes (EGS4 [85Nel], MCNP-4 [91Bri], MCNP-BO code [94Gua1, 94Gua2], PTB-BG code [86Gro] etc.) enabling them to derive fluence-to-dose-equivalent conversion coefficients for parallel electron beams of energies between 60 keV and 10 MeV. The conversion coefficients for H’(0.07,α), H’(3,α), Hp,slab(0.07,α), Hp,slab(3,α) and Hp,slab(10,α) were determined with the ICRU sphere or the 4-element ICRU tissue slab phantom, respectively. A compilation of data can be found in [96Cha], in ICRP 74 and ICRU 57 [96ICR, 98ICR]. By appropriately averaging of these data, reference fluence-to-dose-equivalent conversion coefficients were derived as a function of energy for normally incident electrons. Table 6.9 shows these data for depths of 0.07 mm and 3 mm. Table 6.8. Organ absorbed dose per unit fluence DT/Φ and effective dose per unit fluence E/Φ for monoenergetic electrons incident in the AP geometry on an adult anthropomorphic computational model (Schultz and Zoetelief, data from [96ICR]). Energy [MeV]
0.1
0.4
0.6
Skin Testes Bone marrow Stomach Breast Liver Thyroid Effective dose
8
98
171 0 0 0
0.1
1
1.5
1.0 1.5 2.0 2 DT/Ф and E/Ф [pGy cm ] 164 158 153 1 14 37 1 5 11 0 14 43 75
2.7
5.9
0 11
4.0
10.0
150 214 28 3 200 0 121 44
165 345 52 184 325 97 297 131
Table 6.9. Reference conversion coefficients from fluence to directional dose equivalent for monoenergetic electrons and normal incidence Energy H´(0.07,0°)/Φ Energy H´(0.07,0°)/Φ H´(3,0°)/Φ H´(3,0°)/Φ [MeV] [MeV] [nSv cm2] [nSv cm2] [nSv cm2] [nSv cm2] 1.00 0.312 0.301 0.07 0.221 1.25 0.296 0.486 0.08 1.056 1.50 0.287 0.524 0.09 1.527 1.75 0.282 0.512 0.10 1.661 2.00 0.279 0.481 0.1125 1.627 2.50 0.278 0.417 0.125 1.513 3.00 0.276 0.373 0.15 1.229 3.50 0.274 0.351 0.20 0.834 4.00 0.272 0.334 0.30 0.542 5.00 0.271 0.317 0.40 0.455 6.00 0.271 0.309 0.50 0.403 7.00 0.271 0.306 0.60 0.366 8.00 0.271 0.305 0.70 0.344 0.000 10.0 0.275 0.303 0.80 0.329 0.045
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42]
6 External dosimetry
6-23
6.5.2 Environmental exposure Due to the short ranges of electrons emitted by radionuclides, electrons contribute only to the dose to skin. Skin dose coefficients for a series of monoenergetic electron sources were calculated by Eckerman and Ryman [93Eck1] using the code DOSFACTER of Kocher [88DOE]. The results are shown in Figs. 6.13 and 6.14 for submersion in contaminated air and for exposure to contaminated soil respectively. These data can be then convoluted to the spectra of the various radionuclides, using the energy and intensity of beta and electron emissions of radionuclides to obtain radionuclide specific conversion coefficients. 100
10-1 Contaminated air
Contaminated soil
10-2 -1 -3 h(skin) [pSv s per Bq m ]
-1 -3 h(skin) [pSv s per Bq m ]
10-1 10-2 10-3 10-4
10-3 10-4 10-5 10-6 10-7 Infinite volume Surface
10-8
10-5 10-9 10-6 10-2
10-1
100
101
Electron Energy [MeV]
Fig. 6.13. Electron skin dose coefficient for submersion in air; [93Eck1].
10-10 10-2
10-1
100
Electron Energy [MeV]
Fig. 6.14. Electron skin dose coefficient for exposure to contaminated soil on the surface and in the volume; [93Eck1].
6.6 Doses from external exposure of radionuclides in the environment Only photons, including bremsstrahlung, and electrons emitted by the radionuclides are sufficiently penetrating to traverse the overlying tissues of the body and contribute to the dose to tissues and organs of the body. The energy spectra of emitted radiation are either discrete, as in the case of photons, or continuous, as in the case of beta particles and bremsstrahlung. The dose coefficient H TS for tissue T for any exposure mode S can be expressed as H TS =
∞
∑ [∑ y j ( E i ) H TS, j ( Ei ) + ∫ 0
j =e ,γ
y j ( E ) H TS, j ( E )dE ]
i
where y j ( Ei ) is the yield of radiations of type j and discrete energy Ei and y j (E ) denotes the yield of radiations per nuclear transformation with continuous energy between E and E + dE. The other summation is over all electron and photon radiations. The contribution of the radiations to the dose in tissue or organ T is defined by the quantity H TS which is estimated by means of Monte Carlo calculations and is given as a function of energy for tissue and organ T for each exposure mode [93Eck1]. Landolt-Börnstein New Series VIII/4
6-24
6 External dosimetry
[Ref. p. 6-42
By using the dose conversion coefficients for monoenergetic sources of photon and electron radiation and by scaling them to the emissions of the radionuclides of interest, dose coefficients from radionuclides in the environment can be derived. The following tables contain data from the American Federal Guidance Report No. 12, based on Monte Carlo radiation transport calculations and data obtained from Eckerman [02Eck]. Tables 6.10 and 6.11 give the skin dose and effective dose coefficients for several radionuclides for exposure to contaminated ground surface to a depth of 5 cm and for air submersion respectively. The nuclear decay data used are from Eckerman et. al. [93Eck2] and are based on the ICRP Publication 38 [83ICR] on radionuclide transformations. Table 6.10. Effective dose and skin dose coefficients for exposure depth of 5 cm ; [93Eck1] and [02Eck]. Effective dose Radionuclide Skin dose Radionuclide [Sv/(Bq s m−2)] [Sv/(Bq s m−2)] He-3 0.00 0.00 Ti-45 Be-7 Sc-46 1.04·10−18 8.55·10−19 Be-10 Ca-47 4.21·10−21 3.02·10−20 C-11 2.23·10−17 1.77·10−17 Sc-47 N-13 2.34·10−17 V-47 1.77·10−17 C-14 1.21·10−22 Cr-48 5.50·10−23 O-15 2.72·10−17 Sc-48 1.78·10−17 F-18 V-48 2.16·10−17 1.77·10−17 Ne-19 Ca-49 3.14·10−17 1.79·10−17 −17 −17 Na-22 4.48·10 3.76·10 Cr-49 Na-24 8.02·10−17 Sc-49 6.80·10−17 Al-26 5.52·10−17 V-49 4.54·10−17 Cr-51 Al-28 5.00·10−17 3.01·10−17 Mg-28 Mn-51 2.74·10−17 2.31·10−17 P-30 Fe-52 4.10·10−17 1.81·10−17 Si-31 Mn-52m 3.59·10−18 7.88·10−20 P-32 Mn-52 5.18·10−18 9.05·10−20 Si-32 Mn-53 2.67·10−22 1.43·10−22 Mn-54 P-33 4.11·10−22 2.36·10−22 Co-55 S-35 1.30·10−22 6.08·10−23 Fe-55 Cl-36 1.76·10−19 9.72·10−21 Co-56 Ar-37 0.00 0.00 Mn-56 Cl-38 5.04·10−17 2.53·10−17 Ni-56 K-38 7.71·10−17 5.38·10−17 Co-57 Ar-39 2.29·10−20 3.37·10−21 Ni-57 Cl-39 3.63·10−17 2.46·10−17 Co-58m K-40 6.11·10−18 2.70·10−18 Co-58 Ar-41 2.75·10−17 2.19·10−17 Fe-59 Ca-41 0.00 0.00 K-42 Ni-59 2.52·10−17 5.05·10−18 Co-60m K-43 2.08·10−17 1.68·10−17 −17 −17 Co-60 Sc-43 2.38·10 1.90·10 −17 −17 Cu-60 K-44 6.38·10 3.81·10 −18 −18 Fe-60 Sc-44m 5.84·10 4.79·10 −17 −17 Co-61 Sc-44 4.76·10 3.68·10 −18 −18 Cu-61 Ti-44 1.90·10 1.47·10 −22 −22 Co-62m Ca-45 4.30·10 2.50·10 −17 −17 Cu-62 K-45 4.66·10 3.14·10
to contaminated ground surface to a Skin dose [Sv/(Bq s m−2)] 1.94·10−17 4.09·10−17 2.27·10−17 2.13·10−18 2.80·10−17 8.80·10−18 6.79·10−17 5.92·10−17 6.47·10−17 2.55·10−17 7.49·10−18 0.00 6.64·10−19 3.05·10−17 1.56·10−17 6.24·10−17 7.01·10−17 0.00 1.72·10−17 4.33·10−17 0.00 7.09·10−17 4.24·10−17 3.53·10−17 2.21·10−18 3.85·10−17 4.58·10−23 2.02·10−17 2.39·10−17 0.00 9.16·10−20 5.02·10−17 8.68·10−17 1.08·10−22 3.26·10−18 1.86·10−17 6.56·10−17 3.77·10−17
Effective dose [Sv/(Bq s m−2)] 1.51·10−17 3.45·10−17 1.81·10−17 1.74·10−18 1.74·10−17 7.19·10−18 5.73·10−17 4.99·10−17 5.08·10−17 1.80·10−17 1.47·10−19 0.00 5.43·10−19 1.74·10−17 1.26·10−17 4.15·10−17 5.92·10−17 0.00 1.44·10−17 3.44·10−17 0.00 6.03·10−17 2.88·10−17 2.94·10−17 1.81·10−18 3.26·10−17 9.55·10−24 1.68·10−17 2.03·10−17 0.00 6.82·10−20 4.27·10−17 6.62·10−17 4.60·10−23 1.17·10−18 1.43·10−17 4.60·10−17 1.77·10−17 Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Zn-62 Ni-63 Zn-63 Cu-64 Ga-65 Ni-65 Zn-65 Cu-66 Ga-66 Ge-66 Ni-66 Cu-67 Ga-67 Ge-67 Ga-68 Ge-68 As-69 Ge-69 Zn-69m Zn-69 As-70 Ga-70 Se-70 As-71 Ge-71 Zn-71m As-72 Ga-72 Zn-72 As-73 Ga-73 Se-73m Se-73 As-74 Br-74m Br-74 Kr-74 Br-75 Ge-75 Se-75 As-76 Br-76 Kr-76 As-77 Br-77 Ge-77 Kr-77 Se-77m Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 9.03·10−18 0.00 3.27·10−17 4.00·10−18 3.21·10−17 1.62·10−17 1.18·10−17 1.40·10−17 6.18·10−17 1.42·10−17 3.11·10−22 2.24·10−18 3.01·10−18 4.62·10−17 2.65·10−17 4.59·10−22 3.76·10−17 1.87·10−17 8.79·10−18 4.90·10−19 9.17·10−17 4.65·10−18 2.34·10−17 1.19·10−17 4.65·10−22 3.56·10−17 4.93·10−17 5.69·10−17 2.81·10−18 6.27·10−20 8.13·10−18 6.34·10−18 2.42·10−17 1.68·10−17 1.00·10−16 1.00·10−16 3.16·10−17 2.93·10−17 2.11·10−18 7.90·10−18 2.15·10−17 6.09·10−17 8.86·10−18 2.64·10−19 6.59·10−18 2.75·10−17 2.62·10−17 1.67·10−18
Effective dose [Sv/(Bq s m−2)] 7.41·10−18 0.00 1.92·10−17 3.28·10−18 2.02·10−17 9.42·10−18 9.96·10−18 1.67·10−18 4.13·10−17 1.15·10−17 1.73·10−22 1.82·10−18 2.45·10−18 2.44·10−17 1.66·10−17 6.94·10−24 1.78·10−17 1.50·10−17 7.21·10−18 1.28·10−20 7.02·10−17 2.08·10−19 1.70·10−17 9.70·10−18 7.02·10−24 2.69·10−17 3.11·10−17 4.58·10−17 2.29·10−18 4.03·10−20 5.31·10−18 4.22·10−18 1.84·10−17 1.31·10−17 6.92·10−17 7.53·10−17 2.00·10−17 2.10·10−17 6.11·10−19 6.46·10−18 7.64·10−18 4.46·10−17 7.23·10−18 1.54·10−19 5.42·10−18 1.87·10−17 1.74·10−17 1.36·10−18
Radionuclide As-78 Ge-78 Kr-79 Rb-79 Se-79 Br-80m Br-80 Rb-80 Sr-80 Kr-81m Kr-81 Rb-81m Rb-81 Se-81m Se-81 Sr-81 Br-82 Rb-82m Rb-82 Sr-82 Br-83 Kr-83m Rb-83 Se-83 Sr-83 Br-84 Rb-84 Kr-85m Kr-85 Sr-85m Sr-85 Rb-86 Y-86m Y-86 Zr-86 Kr-87 Rb-87 Sr-87m Y-87 Kr-88 Nb-88 Rb-88 Y-88 Zr-88 Nb-89b Nb-89a Rb-89 Sr-89
6-25 Skin dose [Sv/(Bq s m−2)] 4.38·10−17 5.94·10−18 5.27·10−18 3.60·10−17 1.61·10−22 1.20·10−19 2.55·10−18 5.75·10−17 1.48·10−20 2.61·10−18 1.24·10−19 7.66·10−20 1.33·10−17 2.40·10−19 4.15·10−18 4.04·10−17 5.43·10−17 5.99·10−17 4.23·10−17 1.45·10−20 6.77·10−19 3.47·10−21 1.05·10−17 5.23·10−17 1.69·10−17 5.04·10−17 1.91·10−17 3.37·10−18 1.67·10−19 4.51·10−18 1.06·10−17 7.19·10−18 4.53·10−18 7.38·10−17 5.59·10−18 3.33·10−17 8.93·10−22 6.73·10−18 9.47·10−18 3.99·10−17 1.00·10−16 4.41·10−17 5.27·10−17 8.32·10−18 4.33·10−17 4.89·10−17 5.26·10−17 3.69·10−18
Effective dose [Sv/(Bq s m−2)] 2.17·10−17 4.77·10−18 4.33·10−18 2.35·10−17 7.61·10−23 4.75·10−20 1.39·10−18 2.23·10−17 5.33·10−22 2.12·10−18 9.48·10−20 5.28·10−20 1.06·10−17 1.91·10−19 2.29·10−19 2.40·10−17 4.55·10−17 5.00·10−17 1.93·10−17 5.24·10−22 1.43·10−19 1.35·10−22 8.61·10−18 4.15·10−17 1.36·10−17 2.99·10−17 1.58·10−17 2.56·10−18 4.40·10−20 3.68·10−18 8.73·10−18 1.71·10−18 3.69·10−18 6.12·10−17 4.52·10−18 1.36·10−17 5.55·10−22 5.51·10−18 7.77·10−18 3.25·10−17 7.11·10−17 1.13·10−17 4.53·10−17 6.80·10−18 2.39·10−17 3.35·10−17 3.52·10−17 6.67·10−20
6-26 Radionuclide Zr-89 Mo-90 Nb-90 Sr-90 Y-90m Y-90 Sr-91 Y-91m Y-91 Sr-92 Y-92 Mo-93m Mo-93 Nb-93m Tc-93m Tc-93 Y-93 Zr-93 Nb-94 Ru-94 Tc-94m Tc-94 Y-94 Nb-95m Nb-95 Tc-95m Tc-95 Y-95 Zr-95 Nb-96 Tc-96m Tc-96 Nb-97m Nb-97 Ru-97 Tc-97m Tc-97 Zr-97 Nb-98 Tc-98 Mo-99 Rh-99m Rh-99 Tc-99m Tc-99 Pd-100 Rh-100 Mo-101
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 2.41·10−17 1.71·10−17 8.42·10−17 1.41·10−20 1.31·10−17 9.85·10−18 1.98·10−17 1.12·10−17 4.07·10−18 2.69·10−17 2.53·10−17 4.52·10−17 2.99·10−20 5.26·10−21 1.39·10−17 2.87·10−17 1.64·10−17 0.00 3.26·10−17 1.09·10−17 4.63·10−17 5.51·10−17 4.70·10−17 1.28·10−18 1.59·10−17 1.38·10−17 1.63·10−17 3.90·10−17 1.54·10−17 5.11·10−17 9.38·10−19 5.16·10−17 1.52·10−17 1.57·10−17 4.73·10−18 3.51·10−20 3.44·10−20 9.34·10−18 5.85·10−17 2.95·10−17 4.45·10−18 1.40·10−17 1.22·10−17 2.38·10−18 7.94·10−22 1.71·10−18 5.49·10−17 3.16·10−17
Effective dose [Sv/(Bq s m−2)] 2.00·10−17 1.38·10−17 7.06·10−17 2.72·10−21 1.07·10−17 1.74·10−19 1.21·10−17 9.19·10−18 1.32·10−19 2.28·10−17 4.71·10−18 3.83·10−17 2.23·10−21 3.94·10−22 1.20·10−17 2.46·10−17 1.78·10−18 0.00 2.72·10−17 9.01·10−18 3.19·10−17 4.59·10−17 1.95·10−17 1.03·10−18 1.32·10−17 1.14·10−17 1.36·10−17 1.52·10−17 1.28·10−17 4.26·10−17 7.71·10−19 4.30·10−17 1.26·10−17 1.14·10−17 3.83·10−18 7.18·10−21 3.02·10−21 3.17·10−18 4.18·10−17 2.45·10−17 2.58·10−18 1.16·10−17 9.99·10−18 1.95·10−18 4.94·10−22 1.29·10−18 4.64·10−17 2.33·10−17
Radionuclide Pd-101 Rh-101m Rh-101 Tc-101 Ag-102 Rh-102m Rh-102 Ag-103 Pd-103 Rh-103m Ru-103 Ag-104m Ag-104 Cd-104 Tc-104 Ag-105 Rh-105 Ru-105 Ag-106m Ag-106 Rh-106m Rh-106 Ru-106 Cd-107 Rh-107 Pd-107 Ag-108m Ag-108 Ag-109m Cd-109 In-109 Pd-109 Ag-110m Ag-110 In-110b In-110a Sn-110 Ag-111 In-111 Sn-111 Ag-112 In-112 Cd-113m Cd-113 In-113m Sn-113 In-114m In-114
[Ref. p. 6-42 Skin dose [Sv/(Bq s m−2)] 6.60·10−18 6.22·10−18 5.09·10−18 9.09·10−18 7.72·10−17 1.08·10−17 4.42·10−17 1.72·10−17 5.50·10−20 6.22·10−21 9.87·10−18 2.98·10−17 5.52·10−17 4.80·10−18 6.22·10−17 1.07·10−17 1.65·10−18 1.76·10−17 5.77·10−17 1.93·10−17 6.01·10−17 2.36·10−17 0.00 2.78·10−19 8.30·10−18 0.00 3.38·10−17 4.49·10−18 8.32·10−20 1.46·10−19 1.35·10−17 9.38·10−19 5.64·10−17 1.52·10−17 6.27·10−17 3.85·10−17 6.03·10−18 1.36·10−18 7.98·10−18 1.17·10−17 3.22·10−17 6.56·10−18 1.72·10−20 7.14·10−22 5.34·10−18 1.81·10−19 1.81·10−18 5.58·10−20
Effective dose [Sv/(Bq s m−2)] 5.39·10−18 5.05·10−18 4.12·10−18 5.78·10−18 5.73·10−17 8.28·10−18 3.67·10−17 1.29·10−17 8.91·10−21 9.09·10−22 8.12·10−18 2.01·10−17 4.59·10−17 3.90·10−18 3.40·10−17 8.72·10−18 1.34·10−18 1.36·10−17 4.83·10−17 1.23·10−17 5.01·10−17 3.91·10−18 0.00 1.62·10−19 5.40·10−18 0.00 2.79·10−17 3.77·10−19 4.61·10−20 5.67·10−20 1.12·10−17 7.58·10−20 4.73·10−17 7.96·10−19 5.23·10−17 2.67·10−17 4.87·10−18 4.71·10−19 6.45·10−18 8.58·10−18 1.16·10−17 4.56·10−18 2.54·10−21 4.45·10−22 4.36·10−18 9.77·10−20 1.46·10−18 4.68·10−20
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Ag-115 Cd-115m Cd-115 In-115m In-115 Sb-115 In-116m Sb-116m Sb-116 Te-116 Cd-117m Cd-117 In-117m In-117 Sb-117 Sn-117m Sb-118m In-119m In-119 Sb-119 Sn-119m I-120m I-120 Sb-120b Sb-120a Xe-120 I-121 Sn-121m Sn-121 Te-121m Te-121 Xe-121 I-122 Sb-122 Xe-122 I-123 Sn-123m Sn-123 Te-123m Te-123 Xe-123 I-124 Sb-124n Sb-124m Sb-124 Cs-125 I-125 Sb-125 Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 2.62·10−17 4.44·10−18 5.53·10−18 3.28·10−18 3.09·10−21 2.04·10−17 4.99·10−17 6.38·10−17 4.72·10−17 9.18·10−19 4.07·10−17 2.48·10−17 4.30·10−18 1.45·10−17 3.33·10−18 2.83·10−18 5.17·10−17 1.21·10−17 2.00·10−17 1.08·10−19 5.18·10−20 1.24·10−16 7.52·10−17 4.94·10−17 1.16·10−17 8.29·10−18 8.62·10−18 2.61·10−20 1.16·10−21 4.22·10−18 1.18·10−17 4.22·10−17 3.36·10−17 1.29·10−17 1.07·10−18 3.04·10−18 4.54·10−18 2.99·10−18 2.70·10−18 9.89·10−20 1.35·10−17 2.38·10−17 3.50·10−22 7.65·10−18 3.89·10−17 1.70·10−17 2.24·10−19 8.82·10−18
Effective dose [Sv/(Bq s m−2)] 1.21·10−17 4.48·10−19 4.03·10−18 2.65·10−18 1.52·10−21 1.56·10−17 4.21·10−17 5.33·10−17 3.67·10−17 6.64·10−19 3.45·10−17 1.86·10−17 1.50·10−18 1.18·10−17 2.67·10−18 2.26·10−18 4.35·10−17 3.38·10−19 1.33·10−17 2.41·10−20 1.13·10−20 9.05·10−17 4.65·10−17 4.15·10−17 7.69·10−18 6.72·10−18 6.84·10−18 7.64·10−21 7.58·10−22 3.41·10−18 9.68·10−18 3.04·10−17 1.66·10−17 7.70·10−18 8.05·10−19 2.41·10−18 2.23·10−18 1.69·10−19 2.16·10−18 2.49·10−20 1.04·10−17 1.86·10−17 7.38·10−23 6.10·10−18 3.10·10−17 1.15·10−17 6.36·10−20 7.21·10−18
Radionuclide Sn-125 Te-125m Xe-125 Ba-126 Cs-126 I-126 Sb-126m Sb-126 Sn-126 Cs-127 Sb-127 Sn-127 Te-127m Te-127 Xe-127 Ba-128 Cs-128 I-128 Sb-128b Sb-128a Sn-128 Cs-129 I-129 Sb-129 Te-129m Te-129 Xe-129m Cs-130 I-130 Sb-130 Ba-131m Ba-131 Cs-131 I-131 La-131 Sb-131 Te-131m Te-131 Xe-131m Cs-132 I-132m I-132 La-132 Te-132 Ba-133m Ba-133 I-133 Te-133m
6-27 Skin dose [Sv/(Bq s m−2)] 1.47·10−17 1.94·10−19 5.09·10−18 3.06·10−18 4.43·10−17 9.69·10−18 3.68·10−17 6.02·10−17 7.67·10−19 8.44·10−18 1.50·10−17 4.28·10−17 6.64·10−20 1.81·10−19 5.34·10−18 1.23·10−18 2.92·10−17 8.51·10−18 6.64·10−17 5.08·10−17 1.30·10−17 5.42·10−18 1.48·10−19 3.15·10−17 1.96·10−18 3.95·10−18 4.32·10−19 1.44·10−17 4.52·10−17 7.32·10−17 1.16·10−18 9.05·10−18 1.31·10−19 8.03·10−18 1.49·10−17 4.15·10−17 2.95·10−17 1.42·10−17 1.58·10−19 1.44·10−17 6.88·10−18 4.99·10−17 4.69·10−17 4.43·10−18 1.13·10−18 7.71·10−18 1.42·10−17 5.26·10−17
Effective dose [Sv/(Bq s m−2)] 5.49·10−18 5.81·10−20 4.07·10−18 2.44·10−18 1.92·10−17 7.70·10−18 2.69·10−17 4.90·10−17 5.82·10−19 6.85·10−18 1.19·10−17 3.26·10−17 2.06·10−20 8.80·10−20 4.28·10−18 9.36·10−19 1.57·10−17 1.57·10−18 5.35·10−17 3.45·10−17 1.06·10−17 4.35·10−18 5.11·10−20 2.47·10−17 5.44·10−19 1.00·10−18 2.33·10−19 8.81·10−18 3.70·10−17 5.61·10−17 9.03·10−19 7.35·10−18 4.20·10−20 6.56·10−18 1.11·10−17 3.18·10−17 2.43·10−17 7.14·10−18 8.13·10−20 1.19·10−17 5.40·10−18 3.94·10−17 3.41·10−17 3.54·10−18 8.70·10−19 6.18·10−18 1.05·10−17 3.96·10−17
6-28 Radionuclide Te-133 Xe-133m Xe-133 Ce-134 Cs-134m Cs-134 I-134 La-134 Te-134 Ba-135m Ce-135 Cs-135m Cs-135 I-135 La-135 Xe-135m Xe-135 Cs-136 Nd-136 Pr-136 Ba-137m Ce-137m Ce-137 Cs-137 La-137 Pr-137 Cs-138 La-138 Nd-138 Pr-138m Pr-138 Xe-138 Ba-139 Ce-139 Nd-139m Nd-139 Pr-139 Ba-140 La-140 Ba-141 Ce-141 La-141 Nd-141m Nd-141 Pm-141 Sm-141m Sm-141 Ba-142
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 2.67·10−17 5.84·10−19 5.59·10−19 1.72·10−19 3.57·10−19 3.24·10−17 5.80·10−17 2.34·10−17 1.84·10−17 9.93·10−19 3.74·10−17 3.30·10−17 2.87·10−22 3.27·10−17 3.76·10−19 9.07·10−18 5.60·10−18 4.42·10−17 5.38·10−18 5.22·10−17 1.27·10−17 8.22·10−19 3.51·10−19 9.23·10−20 1.52·10−19 1.16·10−17 6.15·10−17 2.45·10−17 5.01·10−19 5.20·10−17 3.32·10−17 2.81·10−17 9.72·10−18 2.78·10−18 3.20·10−17 9.73·10−18 2.32·10−18 4.23·10−18 4.89·10−17 2.62·10−17 1.39·10−18 1.09·10−17 1.61·10−17 1.18·10−18 2.29·10−17 4.46·10−17 3.77·10−17 2.29·10−17
Effective dose [Sv/(Bq s m−2)] 1.61·10−17 4.26·10−19 4.00·10−19 6.73·10−20 2.64·10−19 2.69·10−17 4.52·10−17 1.21·10−17 1.50·10−17 7.60·10−19 3.07·10−17 2.74·10−17 1.55·10−22 2.68·10−17 2.37·10−19 7.37·10−18 4.23·10−18 3.70·10−17 4.25·10−18 3.60·10−17 1.03·10−17 6.18·10−19 2.14·10−19 3.62·10−21 5.59·10−20 8.37·10−18 4.02·10−17 2.09·10−17 3.31·10−19 4.23·10−17 1.43·10−17 1.90·10−17 8.29·10−19 2.19·10−18 2.64·10−17 6.72·10−18 1.74·10−18 3.07·10−18 3.93·10−17 1.45·10−17 1.12·10−18 9.09·10−19 1.31·10−17 8.98·10−19 1.28·10−17 3.37·10−17 2.41·10−17 1.78·10−17
Radionuclide La-142 Pm-142 Pr-142m Pr-142 Sm-142 Ce-143 La-143 Pm-143 Pr-143 Ce-144 Pm-144 Pr-144m Pr-144 Eu-145 Gd-145 Pm-145 Pr-145 Sm-145 Eu-146 Gd-146 Pm-146 Sm-146 Eu-147 Gd-147 Nd-147 Pm-147 Pr-147 Sm-147 Tb-147 Eu-148 Gd-148 Pm-148m Pm-148 Eu-149 Gd-149 Nd-149 Pm-149 Tb-149 Eu-150b Eu-150a Pm-150 Tb-150 Gd-151 Nd-151 Pm-151 Sm-151 Tb-151 Eu-152m Eu-152
[Ref. p. 6-42 Skin dose [Sv/(Bq s m−2)] 6.11·10−17 3.83·10−17 0.00 8.94·10−18 1.67·10−18 6.97·10−18 1.94·10−17 6.17·10−18 5.06·10−19 3.33·10−19 3.23·10−17 1.03·10−19 1.58·10−17 2.91·10−17 5.08·10−17 2.44·10−19 5.41·10−18 5.42·10−19 5.12·10−17 3.82·10−18 1.56·10−17 0.00 9.63·10−18 2.71·10−17 2.76·10−18 3.30·10−22 2.39·10−17 0.00 3.94·10−17 4.47·10−17 0.00 4.17·10−17 1.81·10−17 9.06·10−19 8.14·10−18 9.74·10−18 1.15·10−18 3.38·10−17 3.09·10−17 1.61·10−18 3.58·10−17 4.10·10−17 8.52·10−19 2.29·10−17 6.95·10−18 2.46·10−23 1.79·10−17 9.77·10−18 2.31·10−17
Effective dose [Sv/(Bq s m−2)] 4.59·10−17 1.53·10−17 0.00 1.12·10−18 1.23·10−18 4.50·10−18 1.92·10−18 5.04·10−18 1.28·10−20 2.61·10−19 2.67·10−17 5.18·10−20 8.17·10−19 2.44·10−17 3.80·10−17 1.21·10−19 3.12·10−19 2.85·10−19 4.27·10−17 2.93·10−18 1.28·10−17 0.00 7.90·10−18 2.25·10−17 2.08·10−18 1.96·10−22 1.44·10−17 0.00 2.71·10−17 3.71·10−17 0.00 3.45·10−17 9.94·10−18 6.59·10−19 6.59·10−18 6.34·10−18 2.02·10−19 2.72·10−17 2.54·10−17 7.76·10−19 2.45·10−17 2.87·10−17 6.08·10−19 1.55·10−17 5.30·10−18 3.62·10−24 1.46·10−17 4.95·10−18 1.93·10−17 Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Gd-152 Gd-153 Sm-153 Tb-153 Eu-154 Tb-154 Dy-155 Eu-155 Ho-155 Sm-155 Tb-155 Eu-156 Sm-156 Tb-156m Tb-156n Tb-156 Dy-157 Eu-157 Ho-157 Tb-157 Eu-158 Tb-158 Dy-159 Gd-159 Ho-159 Tb-160 Er-161 Ho-161 Tb-161 Ho-162m Ho-162 Tm-162 Yb-162 Ho-164m Ho-164 Dy-165 Er-165 Dy-166 Ho-166m Ho-166 Tm-166 Yb-166 Ho-167 Tm-167 Yb-167 Er-169 Lu-169 Yb-169 Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 0.00 1.33·10−18 9.29·10−19 4.03·10−18 2.58·10−17 4.51·10−17 1.14·10−17 9.06·10−19 9.53·10−18 4.93·10−18 2.16·10−18 2.93·10−17 2.28·10−18 2.51·10−19 3.96·10−20 3.64·10−17 7.05·10−18 6.10·10−18 9.49·10−18 2.23·10−20 3.11·10−17 1.59·10−17 4.09·10−19 1.38·10−18 6.51·10−18 2.29·10−17 1.83·10−17 5.95·10−19 3.53·10−19 1.11·10−17 2.98·10−18 3.86·10−17 2.18·10−18 4.32·10−19 5.05·10−19 2.26·10−18 3.62·10−19 4.94·10−19 3.59·10−17 5.66·10−18 3.69·10−17 9.43·10−19 7.69·10−18 2.44·10−18 4.11·10−18 8.53·10−22 2.03·10−17 5.02·10−18
Effective dose [Sv/(Bq s m−2)] 0.00 9.54·10−19 6.08·10−19 3.20·10−18 2.10·10−17 3.88·10−17 9.39·10−18 7.06·10−19 6.25·10−18 1.48·10−18 1.65·10−18 2.25·10−17 1.80·10−18 1.55·10−19 2.63·10−20 3.05·10−17 5.68·10−18 4.02·10−18 7.58·10−18 1.22·10−20 1.81·10−17 1.32·10−17 2.35·10−19 7.53·10−19 5.19·10−18 1.91·10−17 1.52·10−17 3.67·10−19 2.25·10−19 9.19·10−18 2.34·10−18 2.96·10−17 1.69·10−18 2.61·10−19 1.92·10−19 4.16·10−19 2.20·10−19 3.49·10−19 2.98·10−17 4.97·10−19 3.11·10−17 6.35·10−19 6.14·10−18 1.89·10−18 3.18·10−18 5.43·10−22 1.70·10−17 3.89·10−18
Radionuclide Hf-170 Lu-170 Tm-170 Er-171 Lu-171 Tm-171 Er-172 Hf-172 Lu-172 Ta-172 Tm-172 Hf-173 Lu-173 Ta-173 Tm-173 Lu-174m Lu-174 Ta-174 Hf-175 Ta-175 Tm-175 Yb-175 Lu-176m Lu-176 Ta-176 W-176 Hf-177m Lu-177m Lu-177 Re-177 Ta-177 W-177 Yb-177 Hf-178m Lu-178m Lu-178 Re-178 Ta-178b Ta-178a W-178 Yb-178 Hf-179m Lu-179 Ta-179 W-179 Hf-180m Os-180 Re-180
6-29 Skin dose [Sv/(Bq s m−2)] 1.06·10−17 4.79·10−17 6.24·10−19 8.55·10−18 1.36·10−17 7.02·10−21 1.07·10−17 1.40·10−18 3.78·10−17 3.56·10−17 1.29·10−17 7.59·10−18 1.94·10−18 1.45·10−17 8.59·10−18 7.57·10−19 2.06·10−18 1.54·10−17 7.25·10−18 1.80·10−17 2.39·10−17 8.07·10−19 1.90·10−18 9.98·10−18 4.20·10−17 2.47·10−18 4.56·10−17 1.99·10−17 6.68·10−19 1.47·10−17 8.91·10−19 1.76·10−17 5.60·10−18 4.85·10−17 2.42·10−17 9.38·10−18 3.01·10−17 2.03·10−17 1.70·10−18 1.52·10−19 7.64·10−19 1.81·10−17 2.64·10−18 3.56·10−19 6.13·10−19 2.05·10−17 8.23·10−19 2.44·10−17
Effective dose [Sv/(Bq s m−2)] 8.57·10−18 4.10·10−17 6.50·10−20 6.21·10−18 1.12·10−17 4.75·10−21 8.79·10−18 1.00·10−18 3.16·10−17 2.60·10−17 8.07·10−18 6.13·10−18 1.46·10−18 9.38·10−18 6.67·10−18 5.42·10−19 1.65·10−18 1.02·10−17 5.85·10−18 1.51·10−17 1.81·10−17 6.57·10−19 1.78·10−19 8.11·10−18 3.56·10−17 1.89·10−18 3.71·10−17 1.61·10−17 5.39·10−19 9.90·10−18 6.60·10−19 1.44·10−17 3.14·10−18 3.97·10−17 1.84·10−17 2.42·10−18 2.01·10−17 1.64·10−17 1.35·10−18 1.06·10−19 6.00·10−19 1.47·10−17 5.48·10−19 2.45·10−19 4.06·10−19 1.67·10−17 6.04·10−19 1.98·10−17
6-30 Radionuclide Ta-180m Ta-180 Hf-181 Os-181 Re-181 W-181 Hf-182m Hf-182 Ir-182 Os-182 Re-182b Re-182a Ta-182m Ta-182 Hf-183 Ta-183 Hf-184 Ir-184 Re-184m Re-184 Ta-184 Ir-185 Os-185 Ta-185 W-185 Ir-186a Ir-186b Pt-186 Re-186m Re-186 Ta-186 Ir-187 Re-187 W-187 Ir-188 Pt-188 Re-188m Re-188 W-188 Ir-189 Os-189m Pt-189 Re-189 Ir-190n Ir-190m Ir-190 Os-190m Ir-191m
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 5.75·10−19 1.11·10−17 1.13·10−17 2.43·10−17 1.53·10−17 4.61·10−19 1.89·10−17 4.91·10−18 3.97·10−17 8.53·10−18 3.70·10−17 2.30·10−17 4.49·10−18 2.57·10−17 1.70·10−17 5.44·10−18 5.34·10−18 4.03·10−17 7.48·10−18 1.78·10−17 3.44·10−17 1.12·10−17 1.46·10−17 8.03·10−18 2.34·10−21 3.28·10−17 2.08·10−17 1.50·10−17 1.69·10−19 1.12·10−18 4.13·10−17 6.96·10−18 0.00 1.03·10−17 3.07·10−17 3.57·10−18 1.04·10−18 7.88·10−18 3.88·10−20 1.19·10−18 5.34·10−22 6.00·10−18 1.89·10−18 3.20·10−17 5.86·10−22 2.95·10−17 3.32·10−17 1.11·10−18
Effective dose [Sv/(Bq s m−2)] 4.00·10−19 8.97·10−18 9.29·10−18 2.02·10−17 1.26·10−17 3.24·10−19 1.55·10−17 4.00·10−18 2.29·10−17 6.93·10−18 3.09·10−17 1.93·10−17 3.59·10−18 2.16·10−17 1.27·10−17 4.35·10−18 3.90·10−18 3.20·10−17 6.13·10−18 1.48·10−17 2.75·10−17 9.44·10−18 1.20·10−17 2.89·10−18 1.66·10−21 2.74·10−17 1.61·10−17 1.24·10−17 1.15·10−19 2.92·10−19 2.67·10−17 5.69·10−18 0.00 8.05·10−18 2.62·10−17 2.85·10−18 7.83·10−19 1.05·10−18 3.13·10−20 9.09·10−19 9.45·10−24 4.87·10−18 1.10·10−18 2.63·10−17 1.10·10−23 2.42·10−17 2.73·10−17 8.75·10−19
Radionuclide Os-191m Os-191 Pt-191 Ir-192m Ir-192 Au-193 Hg-193m Hg-193 Os-193 Pt-193m Pt-193 Au-194 Hg-194 Ir-194m Ir-194 Os-194 Tl-194m Tl-194 Au-195m Au-195 Hg-195m Hg-195 Ir-195m Ir-195 Pb-195m Pt-195m Tl-195 Hg-197m Hg-197 Pt-197m Pt-197 Tl-197 Au-198m Au-198 Pb-198 Tl-198m Tl-198 Au-199 Hg-199m Pb-199 Pt-199 Tl-199 Au-200m Au-200 Bi-200 Pb-200 Pt-200 Tl-200
[Ref. p. 6-42 Skin dose [Sv/(Bq s m−2)] 9.26·10−20 1.18·10−18 5.46·10−18 3.16·10−18 1.72·10−17 2.67·10−18 2.09·10−17 3.47·10−18 2.30·10−18 1.42·10−19 1.66·10−21 2.11·10−17 2.32·10−21 4.90·10−17 9.47·10−18 9.62·10−21 5.06·10−17 1.59·10−17 4.03·10−18 1.11·10−18 4.05·10−18 3.69·10−18 8.42·10−18 1.49·10−18 3.37·10−17 9.94·10−19 2.52·10−17 1.58·10−18 9.19·10−19 1.41·10−18 4.18·10−19 7.93·10−18 1.10·10−17 9.07·10−18 8.60·10−18 2.46·10−17 3.96·10−17 1.67·10−18 3.40·10−18 2.94·10−17 7.09·10−18 4.65·10−18 4.34·10−17 1.23·10−17 5.03·10−17 3.64·10−18 1.02·10−18 2.64·10−17
Effective dose [Sv/(Bq s m−2)] 6.74·10−20 9.21·10−19 4.38·10−18 2.58·10−18 1.41·10−17 2.11·10−18 1.73·10−17 2.72·10−18 1.17·10−18 1.06·10−19 3.38·10−23 1.77·10−17 5.71·10−23 4.03·10−17 1.68·10−18 4.92·10−21 3.96·10−17 1.30·10−17 3.27·10−18 8.45·10−19 3.29·10−18 3.00·10−18 6.67·10−18 6.48·10−19 2.70·10−17 7.59·10−19 2.11·10−17 1.27·10−18 7.00·10−19 1.11·10−18 3.01·10−19 6.49·10−18 8.91·10−18 7.01·10−18 7.00·10−18 2.02·10−17 3.36·10−17 1.35·10−18 2.74·10−18 2.47·10−17 3.50·10−18 3.76·10−18 3.58·10−17 4.77·10−18 4.07·10−17 2.93·10−18 7.90·10−19 2.21·10−17
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Au-201 Bi-201 Pb-201 Tl-201 Bi-202 Pb-202m Pb-202 Tl-202 Bi-203 Hg-203 Pb-203 Po-203 Tl-204 Bi-205 Pb-205 Po-205 Bi-206 Tl-206 At-207 Bi-207 Po-207 Tl-207 Tl-208 Pb-209 Tl-209 Bi-210m Bi-210 Pb-210 Po-210 At-211 Bi-211 Pb-211 Po-211 Bi-212 Pb-212 Po-212 Bi-213 Po-213 Bi-214 Pb-214 Po-214 At-215 Po-215 At-216 Po-216 At-217 At-218 Po-218 Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 2.62·10−18 2.76·10−17 1.53·10−17 1.37·10−18 5.56·10−17 4.23·10−17 2.07·10−21 9.40·10−18 4.74·10−17 4.92·10−18 6.04·10−18 3.37·10−17 1.92·10−19 3.34·10−17 2.24·10−21 3.18·10−17 6.65·10−17 3.03·10−18 2.65·10−17 3.21·10−17 2.70·10−17 2.48·10−18 6.74·10−17 3.80·10−20 4.51·10−17 5.32·10−18 1.20·10−18 2.22·10−20 1.77·10−22 5.73·10−19 9.71·10−19 3.08·10−18 1.63·10−19 7.85·10−18 2.88·10−18 0.00 4.53·10−18 0.00 3.53·10−17 5.23·10−18 1.73·10−21 4.07·10−21 3.71·10−21 2.25·10−20 3.51·10−22 6.40·10−21 4.28·10−20 1.89·10−22
Effective dose [Sv/(Bq s m−2)] 9.30·10−19 2.25·10−17 1.26·10−17 1.07·10−18 4.61·10−17 3.52·10−17 3.93·10−23 7.67·10−18 4.01·10−17 4.01·10−18 4.89·10−18 2.77·10−17 1.77·10−20 2.83·10−17 4.39·10−23 2.66·10−17 5.57·10−17 5.59·10−20 2.22·10−17 2.61·10−17 2.25·10−17 8.28·10−20 5.58·10−17 3.19·10−21 3.43·10−17 4.34·10−18 2.43·10−20 1.05·10−20 1.47·10−22 4.51·10−19 7.93·10−19 9.07·10−19 1.35·10−19 3.23·10−18 2.34·10−18 0.00 2.32·10−18 0.00 2.57·10−17 4.18·10−18 1.44·10−21 3.33·10−21 3.05·10−21 1.78·10−20 2.92·10−22 5.26·10−21 2.50·10−20 1.57·10−22
Radionuclide Rn-218 Fr-219 Rn-219 Fr-220 Rn-220 Fr-221 Fr-222 Ra-222 Rn-222 Ac-223 Fr-223 Ra-223 Ac-224 Ra-224 Ac-225 Ra-225 Ac-226 Ra-226 Th-226 Ac-227 Pa-227 Ra-227 Th-227 Ac-228 Pa-228 Ra-228 Th-228 Th-229 Pa-230 Th-230 U-230 Pa-231 Th-231 U-231 Np-232 Pa-232 Th-232 U-232 Np-233 Pa-233 U-233 Np-234 Pa-234m Pa-234 Pu-234 Th-234 U-234 Np-235
6-31 Skin dose [Sv/(Bq s m−2)] 1.59·10−20 7.25·10−20 1.17·10−18 2.00·10−19 8.12·10−21 6.22·10−19 6.16·10−18 1.93·10−19 8.40·10−21 8.91·10−20 1.69·10−18 2.50·10−18 3.64·10−18 2.03·10−19 2.86·10−19 9.80·10−20 3.00·10−18 1.32·10−19 1.46·10−19 2.45·10−21 3.16·10−19 4.34·10−18 2.09·10−18 2.15·10−17 2.28·10−17 0.00 3.81·10−20 1.49·10−18 1.30·10−17 8.32·10−21 2.38·10−20 7.69·10−19 2.15·10−19 1.13·10−18 2.44·10−17 1.92·10−17 5.04·10−21 8.66·10−21 1.49·10−18 4.02·10−18 8.26·10−21 2.85·10−17 8.02·10−18 3.93·10−17 1.09·10−18 1.24·10−19 5.61·10−21 3.02·10−20
Effective dose [Sv/(Bq s m−2)] 1.31·10−20 5.94·10−20 9.56·10−19 1.61·10−19 6.68·10−21 5.06·10−19 1.08·10−19 1.58·10−19 6.91·10−21 7.01·10−20 6.90·10−19 2.03·10−18 2.95·10−18 1.66·10−19 2.28·10−19 4.58·10−20 2.07·10−18 1.07·10−19 1.17·10−19 1.80·10−21 2.47·10−19 2.62·10−18 1.69·10−18 1.65·10−17 1.90·10−17 0.00 2.85·10−20 1.18·10−18 1.09·10−17 4.62·10−21 1.56·10−20 6.02·10−19 1.41·10−19 8.79·10−19 2.02·10−17 1.60·10−17 2.05·10−21 3.38·10−21 1.20·10−18 3.27·10−18 4.81·10−21 2.41·10−17 3.33·10−19 3.24·10−17 8.73·10−19 9.49·10−20 1.55·10−21 1.23·10−20
6-32
6 External dosimetry Skin dose [Sv/(Bq s m−2)] 3.27·10−21 8.76·10−19 4.22·10−18 1.55·10−18 8.92·10−18 2.28·10−17 1.50·10−17 1.92·10−17 4.40·10−21 4.38·10−21 2.46·10−18 1.88·10−18 6.61·10−18 4.11·10−21 3.34·10−21 4.80·10−23 6.95·10−18 7.31·10−19 1.86·10−17 3.91·10−21 0.00 7.78·10−18 2.29·10−18 1.64·10−18 4.03·10−21 3.93·10−21 5.66·10−22 9.07·10−21 1.41·10−18 1.27·10−23 1.01·10−17 1.01·10−19 4.63·10−21 6.14·10−20 1.89·10−18 2.08·10−18 2.83·10−20
Effective dose [Sv/(Bq s m−2)] 2.92·10−22 4.94·10−19 3.42·10−18 1.25·10−18 7.02·10−18 1.74·10−17 1.15·10−17 1.59·10−17 7.06·10−22 4.43·10−22 1.96·10−18 1.52·10−18 5.41·10−18 4.69·10−22 3.35·10−22 1.86·10−23 5.68·10−18 3.39·10−19 1.52·10−17 4.45·10−22 0.00 6.46·10−18 1.84·10−18 1.32·10−18 6.60·10−22 5.05·10−22 3.04·10−22 5.56·10−21 1.14·10−18 1.45·10−24 8.04·10−18 4.54·10−20 9.12·10−22 2.36·10−20 1.52·10−18 1.69·10−18 9.02·10−21
Table 6.11. Skin dose and effective dose coefficients for air submersion; [93Eck1]. Effective Dose Radionuclide Skin Dose Radionuclide Skin Dose [Sv/(Bq s m−2)] [Sv/(Bq s m−2)] [Sv/(Bq s m−2)] He-3 0.00 0.00 C-14 2.43·10−16 −15 −15 Be-7 2.74·10 2.19·10 O-15 1.04·10−13 −14 −16 Be-10 1.29·10 1.38·10 F-18 6.94·10−14 −14 −14 C-11 7.91·10 Ne-19 4.56·10 1.21·10−13 −14 −14 N-13 8.68·10 Na-22 4.57·10 1.33·10−13
Effective Dose [Sv/(Bq s m−2)] 2.60·10−18 4.59·10−14 4.56·10−14 4.62·10−14 1.02·10−13
Radionuclide Pu-235 U-235 Np-236a Np-236b Pu-236 U-236 Am-237 Np-237 Pu-237 U-237 Am-238 Cm-238 Np-238 Pu-238 U-238 Am-239 Np-239 Pu-239 U-239 Am-240 Cm-240 Np-240m Np-240 Pu-240 U-240 Am-241 Cm-241 Pu-241 Am-242m Am-242 Cm-242 Pu-242 Am-243 Cm-243 Pu-243 Am-244m Am-244 Cf-244 Cm-244
Skin dose [Sv/(Bq s m−2)] 1.52·10−18 3.01·10−18 2.12·10−18 8.38·10−19 6.04·10−21 4.46·10−21 7.21·10−18 4.03·10−19 7.74·10−19 2.37·10−18 1.79·10−17 1.26·10−18 1.19·10−17 5.03·10−21 3.54·10−21 4.20·10−18 3.17·10−18 2.87·10−21 2.10·10−18 2.07·10−17 5.87·10−21 1.16·10−17 2.69·10−17 4.83·10−21 2.64·10−20 2.89·10−19 9.94·10−18 2.95·10−23 2.05·10−20 2.69·10−19 5.42·10−21 4.03·10−21 7.59·10−19 2.44·10−18 3.91·10−19 2.61·10−18 1.65·10−17 6.02·10−21 4.89·10−21
Effective dose [Sv/(Bq s m−2)] 1.22·10−18 2.45·10−18 1.69·10−18 6.72·10−19 8.63·10−22 8.36·10−22 5.88·10−18 2.96·10−19 6.12·10−19 1.88·10−18 1.49·10−17 1.01·10−18 9.41·10−18 5.78·10−22 4.20·10−22 3.39·10−18 2.57·10−18 1.01·10−21 6.37·10−19 1.72·10−17 6.01·10−22 5.76·10−18 2.21·10−17 5.66·10−22 5.79·10−21 1.85·10−19 8.12·10−18 2.20·10−23 6.37·10−21 1.88·10−19 6.38·10−22 4.94·10−22 5.78·10−19 1.98·10−18 3.01·10−19 4.80·10−20 1.35·10−17 6.80·10−22 4.79·10−22
[Ref. p. 6-42
Radionuclide Pu-244 Am-245 Bk-245 Cm-245 Pu-245 Am-246m Am-246 Bk-246 Cf-246 Cm-246 Pu-246 Bk-247 Cm-247 Cf-248 Cm-248 Bk-249 Cf-249 Cm-249 Bk-250 Cf-250 Cm-250 Es-250 Cf-251 Es-251 Cf-252 Fm-252 Cf-253 Es-253 Fm-253 Cf-254 Es-254m Es-254 Fm-254 Fm-255 Fm-257 Md-257 Md-258
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Na-24 Al-26 Al-28 Mg-28 P-30 Si-31 P-32 Si-32 P-33 S-35 Cl-36 Ar-37 Cl-38 K-38 Ar-39 Cl-39 K-40 Ar-41 Ca-41 K-42 K-43 Sc-43 K-44 Sc-44m Sc-44 Ti-44 Ca-45 K-45 Ti-45 Sc-46 Ca-47 Sc-47 V-47 Cr-48 Sc-48 V-48 Ca-49 Cr-49 Sc-49 V-49 Cr-51 Mn-51 Fe-52 Mn-52m Mn-52 Mn-53 Mn-54 Co-55 Fe-55 Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 2.75·10−13 1.81·10−13 1.88·10−13 8.33·10−14 1.56·10−13 3.78·10−14 4.49·10−14 8.27·10−16 1.38·10−15 2.92·10−16 1.47·10−14 0.00 1.94·10−13 2.66·10−13 1.07·10−14 1.36·10−13 4.20·10−14 1.01·10−13 0.00 1.15·10−13 7.11·10−14 7.91·10−14 2.35·10−13 1.72·10−14 1.58·10−13 6.79·10−15 1.46·10−15 1.74·10−13 7.07·10−14 1.17·10−13 8.02·10−14 1.28·10−14 1.08·10−13 2.40·10−14 2.01·10−13 1.72·10−13 2.46·10−13 9.65·10−14 5.43·10−14 0.00 1.75·10−15 1.18·10−13 5.17·10−14 2.13·10−13 1.99·10−13 0.00 4.67·10−14 1.39·10−13 0.00
Effective Dose [Sv/(Bq s m−2)] 2.08·10−13 1.28·10−13 8.87·10−14 6.38·10−14 4.68·10−14 4.83·10−16 5.36·10−16 8.68·10−18 1.45·10−17 3.11·10−18 1.66·10−16 0.00 7.58·10−14 1.56·10−13 1.15·10−16 6.90·10−14 7.92·10−15 6.14·10−14 0.00 1.48·10−14 4.35·10−14 4.88·10−14 1.14·10−13 1.24·10−14 9.87·10−14 4.70·10−15 1.53·10−17 9.20·10−14 3.89·10−14 9.36·10−14 5.06·10−14 4.67·10−15 4.49·10−14 1.87·10−14 1.57·10−13 1.36·10−13 1.66·10−13 4.68·10−14 7.16·10−16 0.00 1.38·10−15 4.51·10−14 3.27·10−14 1.13·10−13 1.62·10−13 0.00 3.83·10−14 9.16·10−14 0.00
Radionuclide Co-56 Mn-56 Ni-56 Co-57 Ni-57 Co-58m Co-58 Fe-59 Ni-59 Co-60m Co-60 Cu-60 Fe-60 Co-61 Cu-61 Co-62m Cu-62 Zn-62 Ni-63 Zn-63 Cu-64 Ga-65 Ni-65 Zn-65 Cu-66 Ga-66 Ge-66 Ni-66 Cu-67 Ga-67 Ge-67 Ga-68 Ge-68 As-69 Ge-69 Zn-69m Zn-69 As-70 Ga-70 Se-70 As-71 Ge-71 Zn-71m As-72 Ga-72 Zn-72 As-73 Ga-73
6-33 Skin Dose [Sv/(Bq s m−2)] 2.13·10−13 1.51·10−13 9.61·10−14 6.63·10−15 1.17·10−13 3.05·10−19 5.58·10−14 7.13·10−14 0.00 3.46·10−16 1.45·10−13 2.82·10−13 1.64·10−16 3.24·10−14 6.50·10−14 2.25·10−13 1.44·10−13 2.52·10−14 0.00 1.23·10−13 1.64·10−14 1.19·10−13 7.18·10−14 3.29·10−14 7.69·10−14 2.11·10−13 4.26·10−14 1.01·10−15 1.18·10−14 8.50·10−15 1.68·10−13 1.01·10−13 6.62·10−18 1.43·10−13 5.96·10−14 2.44·10−14 1.81·10−14 2.89·10−13 4.17·10−14 8.36·10−14 3.78·10−14 6.71·10−18 1.21·10−13 1.70·10−13 1.86·10−13 1.00·10−14 2.78·10−16 4.37·10−14
Effective Dose [Sv/(Bq s m−2)] 1.73·10−13 8.16·10−14 7.82·10−14 4.97·10−15 9.12·10−14 6.06·10−20 4.44·10−14 5.62·10−14 0.00 2.00·10−16 1.19·10−13 1.87·10−13 1.79·10−18 3.74·10−15 3.72·10−14 1.30·10−13 4.60·10−14 1.92·10−14 0.00 5.00·10−14 8.50·10−15 5.28·10−14 2.67·10−14 2.72·10−14 4.89·10−15 1.23·10−13 3.00·10−14 1.06·10−17 4.90·10−15 6.49·10−15 6.45·10−14 4.29·10−14 1.01·10−19 4.61·10−14 3.99·10−14 1.84·10−14 1.99·10−16 1.92·10−13 8.40·10−16 4.40·10−14 2.53·10−14 1.02·10−19 6.99·10−14 8.26·10−14 1.31·10−13 6.17·10−15 1.55·10−16 1.39·10−14
6-34 Radionuclide Se-73m Se-73 As-74 Br-74m Br-74 Kr-74 Br-75 Ge-75 Se-75 As-76 Br-76 Kr-76 As-77 Br-77 Ge-77 Kr-77 Se-77m As-78 Ge-78 Kr-79 Rb-79 Se-79 Br-80m Br-80 Rb-80 Sr-80 Kr-81m Kr-81 Rb-81m Rb-81 Se-81m Se-81 Sr-81 Br-82 Rb-82m Rb-82 Sr-82 Br-83 Kr-83m Rb-83 Se-83 Sr-83 Br-84 Rb-84 Kr-85m Kr-85 Sr-85m Sr-85
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 2.39·10−14 8.31·10−14 5.80·10−14 3.31·10−13 3.40·10−13 1.16·10−13 1.01·10−13 2.71·10−14 2.16·10−14 9.61·10−14 1.97·10−13 2.37·10−14 1.20·10−14 1.77·10−14 1.02·10−13 9.74·10−14 6.99·10−15 1.65·10−13 2.75·10−14 1.50·10−14 1.28·10−13 3.71·10−16 7.13·10−16 2.02·10−14 2.11·10−13 1.44·10−16 9.42·10−15 4.04·10−16 4.01·10−16 4.46·10−14 1.40·10−15 3.94·10−14 1.44·10−13 1.54·10−13 1.68·10−13 1.58·10−13 1.42·10−16 1.85·10−14 3.56·10−17 2.77·10−14 1.69·10−13 5.20·10−14 1.88·10−13 5.71·10−14 2.24·10−14 1.32·10−14 1.23·10−14 2.83·10−14
Effective Dose [Sv/(Bq s m−2)] 1.09·10−14 4.78·10−14 3.40·10−14 1.96·10−13 2.26·10−13 5.20·10−14 5.43·10−14 1.78·10−15 1.68·10−14 2.06·10−14 1.26·10−13 1.86·10−14 5.09·10−16 1.40·10−14 4.98·10−14 4.51·10−14 3.63·10−15 6.03·10−14 1.23·10−14 1.12·10−14 6.08·10−14 3.94·10−18 2.37·10−16 3.73·10−15 5.77·10−14 5.00·10−18 5.56·10−15 2.44·10−16 1.63·10−16 2.73·10−14 5.48·10−16 8.69·10−16 6.24·10−14 1.21·10−13 1.34·10−13 5.01·10−14 4.92·10−18 5.34·10−16 1.20·10−18 2.21·10−14 1.14·10−13 3.60·10−14 9.02·10−14 4.18·10−14 6.87·10−15 2.40·10−16 9.48·10−15 2.24·10−14
Radionuclide Rb-86 Y-86m Y-86 Zr-86 Kr-87 Rb-87 Sr-87m Y-87 Kr-88 Nb-88 Rb-88 Y-88 Zr-88 Nb-89b Nb-89a Rb-89 Sr-89 Zr-89 Mo-90 Nb-90 Sr-90 Y-90m Y-90 Sr-91 Y-91m Y-91 Sr-92 Y-92 Mo-93m Mo-93 Nb-93m Tc-93m Tc-93 Y-93 Zr-93 Nb-94 Ru-94 Tc-94m Tc-94 Y-94 Nb-95m Nb-95 Tc-95m Tc-95 Y-95 Zr-95 Nb-96 Tc-96m
[Ref. p. 6-42 Skin Dose [Sv/(Bq s m−2)] 4.85·10−14 1.28·10−14 2.17·10−13 1.56·10−14 1.37·10−13 3.15·10−15 2.15·10−14 2.51·10−14 1.35·10−13 3.12·10−13 1.83·10−13 1.54·10−13 2.26·10−14 1.56·10−13 1.63·10−13 1.87·10−13 3.69·10−14 7.07·10−14 5.52·10−14 2.66·10−13 9.20·10−15 3.75·10−14 6.24·10−14 8.14·10−14 3.11·10−14 3.85·10−14 8.56·10−14 1.14·10−13 1.32·10−13 2.43·10−16 4.28·10−17 4.62·10−14 8.30·10−14 8.50·10−14 0.00 9.52·10−14 2.95·10−14 1.55·10−13 1.51·10−13 1.80·10−13 1.12·10−14 4.30·10−14 3.76·10−14 4.42·10−14 1.59·10−13 4.50·10−14 1.52·10−13 2.68·10−15
Effective Dose [Sv/(Bq s m−2)] 4.94·10−15 9.59·10−15 1.69·10−13 1.17·10−14 3.97·10−14 3.30·10−17 1.41·10−14 1.99·10−14 9.71·10−14 1.89·10−13 3.33·10−14 1.30·10−13 1.73·10−14 6.62·10−14 8.65·10−14 1.01·10−13 4.37·10−16 5.31·10−14 3.64·10−14 2.05·10−13 9.83·10−17 2.77·10−14 7.92·10−16 3.27·10−14 2.37·10−14 6.22·10−16 6.41·10−14 1.32·10−14 1.06·10−13 1.73·10−17 3.05·10−18 3.53·10−14 6.96·10−14 5.28·10−15 0.00 7.20·10−14 2.36·10−14 8.64·10−14 1.22·10−13 5.39·10−14 2.74·10−15 3.49·10−14 2.99·10−14 3.58·10−14 4.66·10−14 3.36·10−14 1.14·10−13 2.09·10−15
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Tc-96 Nb-97m Nb-97 Ru-97 Tc-97m Tc-97 Zr-97 Nb-98 Tc-98 Mo-99 Rh-99m Rh-99 Tc-99m Tc-99 Pd-100 Rh-100 Mo-101 Pd-101 Rh-101m Rh-101 Tc-101 Ag-102 Rh-102m Rh-102 Ag-103 Pd-103 Rh-103m Ru-103 Ag-104m Ag-104 Cd-104 Tc-104 Ag-105 Rh-105 Ru-105 Ag-106m Ag-106 Rh-106m Rh-106 Ru-106 Cd-107 Pd-107 Rh-107 Ag-108m Ag-108 Ag-109m Cd-109 In-109 Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 1.40·10−13 4.16·10−14 6.51·10−14 1.32·10−14 5.55·10−16 2.71·10−16 5.55·10−14 1.96·10−13 8.53·10−14 3.14·10−14 3.94·10−14 3.42·10−14 7.14·10−15 2.74·10−15 6.11·10−15 1.63·10−13 1.14·10−13 1.94·10−14 1.71·10−14 1.49·10−14 4.77·10−14 2.45·10−13 3.68·10−14 1.19·10−13 5.84·10−14 3.90·10−16 4.49·10−17 2.77·10−14 1.00·10−13 1.56·10−13 1.38·10−14 2.25·10−13 2.90·10−14 1.07·10−14 6.73·10−14 1.58·10−13 7.27·10−14 1.81·10−13 1.09·10−13 0.00 1.50·10−15 0.00 4.42·10−14 9.05·10−14 4.00·10−14 5.59·10−16 9.95·10−16 3.91·10−14
Effective Dose [Sv/(Bq s m−2)] 1.14·10−13 3.31·10−14 2.99·10−14 9.91·10−15 3.72·10−17 2.26·10−17 8.90·10−15 1.14·10−13 6.41·10−14 6.99·10−15 3.06·10−14 2.63·10−14 5.25·10−15 2.87·10−17 3.98·10−15 1.33·10−13 6.48·10−14 1.42·10−14 1.29·10−14 1.09·10−14 1.50·10−14 1.57·10−13 2.15·10−14 9.68·10−14 3.43·10−14 5.32·10−17 6.02·10−18 2.08·10−14 5.48·10−14 1.23·10−13 1.04·10−14 9.61·10−14 2.26·10−14 3.47·10−15 3.56·10−14 1.29·10−13 3.18·10−14 1.35·10−13 1.06·10−14 0.00 5.11·10−16 0.00 1.41·10−14 7.24·10−14 1.25·10−15 1.59·10−16 2.28·10−16 2.98·10−14
Radionuclide Pd-109 Ag-110m Ag-110 In-110b In-110a Sn-110 Ag-111 In-111 Sn-111 Ag-112 In-112 Cd-113m Cd-113 In-113m Sn-113 In-114m In-114 Ag-115 Cd-115m Cd-115 In-115m In-115 Sb-115 In-116m Sb-116m Sb-116 Te-116 Cd-117m Cd-117 In-117m In-117 Sb-117 Sn-117m Sb-118m In-119m In-119 Sb-119 Sn-119m I-120m I-120 Sb-120b Sb-120a Xe-120 I-121 Sn-121m Sn-121 Te-121m Te-121
6-35 Skin Dose [Sv/(Bq s m−2)] 2.15·10−14 1.57·10−13 8.22·10−14 1.71·10−13 1.29·10−13 1.66·10−14 2.19·10−14 2.29·10−14 4.22·10−14 1.33·10−13 2.88·10−14 8.48·10−15 2.41·10−15 2.18·10−14 8.20·10−16 1.05·10−14 2.95·10−15 1.11·10−13 3.99·10−14 2.97·10−14 1.81·10−14 6.18·10−15 6.52·10−14 1.58·10−13 1.82·10−13 1.50·10−13 3.37·10−15 1.29·10−13 8.79·10−14 3.17·10−14 5.16·10−14 1.03·10−14 1.25·10−14 1.46·10−13 7.11·10−14 8.20·10−14 7.09·10−16 3.42·10−16 3.86·10−13 2.55·10−13 1.39·10−13 4.46·10−14 2.40·10−14 2.72·10−14 1.07·10−15 3.71·10−15 1.23·10−14 3.18·10−14
Effective Dose [Sv/(Bq s m−2)] 4.20·10−16 1.27·10−13 2.46·10−15 1.39·10−13 7.15·10−14 1.25·10−14 1.38·10−15 1.68·10−14 2.30·10−14 3.23·10−14 1.19·10−14 9.06·10−17 2.53·10−17 1.12·10−14 3.15·10−16 3.89·10−15 1.59·10−16 3.46·10−14 1.48·10−15 1.05·10−14 6.86·10−15 6.55·10−17 4.02·10−14 1.18·10−13 1.45·10−13 1.02·10−13 1.98·10−15 9.89·10−14 5.14·10−14 4.07·10−15 3.06·10−14 7.15·10−15 6.11·10−15 1.19·10−13 1.26·10−15 3.53·10−14 1.50·10−16 7.04·10−17 2.49·10−13 1.31·10−13 1.14·10−13 2.00·10−14 1.79·10−14 1.78·10−14 5.24·10−17 3.90·10−17 8.99·10−15 2.50·10−14
6-36 Radionuclide Xe-121 I-122 Sb-122 Xe-122 I-123 Sn-123m Sn-123 Te-123m Te-123 Xe-123 I-124 Sb-124n Sb-124m Sb-124 Cs-125 I-125 Sb-125 Sn-125 Te-125m Xe-125 Ba-126 Cs-126 I-126 Sb-126m Sb-126 Sn-126 Cs-127 Sb-127 Sn-127 Te-127m Te-127 Xe-127 Ba-128 Cs-128 I-128 Sb-128b Sb-128a Sn-128 Cs-129 I-129 Sb-129 Te-129m Te-129 Xe-129m Cs-130 I-130 Sb-130 Ba-131m
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 1.40·10−13 1.25·10−13 6.03·10−14 3.36·10−15 9.40·10−15 3.58·10−14 3.28·10−14 8.48·10−15 6.32·10−16 4.52·10−14 7.39·10−14 2.33·10−18 2.46·10−14 1.26·10−13 5.97·10−14 1.39·10−15 2.65·10−14 7.13·10−14 1.94·10−15 1.50·10−14 9.26·10−15 1.62·10−13 3.37·10−14 1.24·10−13 1.73·10−13 6.65·10−15 2.38·10−14 5.58·10−14 1.41·10−13 8.49·10−16 1.14·10−14 1.57·10−14 3.85·10−15 1.07·10−13 5.38·10−14 1.99·10−13 1.73·10−13 4.50·10−14 1.52·10−14 1.10·10−15 1.05·10−13 1.49·10−14 3.57·10−14 8.29·10−15 5.48·10−14 1.36·10−13 2.29·10−13 3.94·10−15
Effective Dose [Sv/(Bq s m−2)] 8.62·10−14 4.31·10−14 2.02·10−14 2.19·10−15 6.49·10−15 6.14·10−15 6.98·10−16 5.81·10−15 1.51·10−16 2.82·10−14 5.04·10−14 4.67·10−19 1.58·10−14 8.62·10−14 3.01·10−14 3.73·10−16 1.87·10−14 1.54·10−14 3.35·10−16 1.08·10−14 6.41·10−15 4.96·10−14 2.01·10−14 7.01·10−14 1.28·10−13 1.84·10−15 1.78·10−14 3.12·10−14 9.03·10−14 1.12·10−16 3.34·10−16 1.12·10−14 2.54·10−15 4.06·10−14 4.33·10−15 1.41·10−13 9.08·10−14 2.77·10−14 1.13·10−14 2.81·10−16 6.71·10−14 1.56·10−15 2.86·10−15 9.14·10−16 2.30·10−14 9.67·10−14 1.50·10−13 2.64·10−15
Radionuclide Ba-131 Cs-131 I-131 La-131 Sb-131 Te-131m Te-131 Xe-131m Cs-132 I-132m I-132 La-132 Te-132 Ba-133m Ba-133 I-133 Te-133m Te-133 Xe-133m Xe-133 Ce-134 Cs-134m Cs-134 I-134 La-134 Te-134 Ba-135m Ce-135 Cs-135m Cs-135 I-135 La-135 Xe-135m Xe-135 Cs-136 Nd-136 Pr-136 Ba-137m Ce-137m Ce-137 Cs-137 La-137 Pr-137 Cs-138 La-138 Nd-138 Pr-138m Pr-138
[Ref. p. 6-42 Skin Dose [Sv/(Bq s m−2)] 2.55·10−14 7.84·10−16 2.98·10−14 4.87·10−14 1.40·10−13 8.85·10−14 6.89·10−14 4.82·10−15 3.92·10−14 2.22·10−14 1.58·10−13 1.49·10−13 1.39·10−14 1.36·10−14 2.19·10−14 5.83·10−14 1.74·10−13 1.06·10−13 1.04·10−14 4.97·10−15 9.60·10−16 2.88·10−15 9.45·10−14 1.87·10−13 8.88·10−14 6.35·10−14 1.30·10−14 1.10·10−13 9.10·10−14 9.06·10−16 1.11·10−13 1.49·10−15 2.97·10−14 3.12·10−14 1.25·10−13 1.71·10−14 1.69·10−13 3.73·10−14 1.20·10−14 1.45·10−15 8.63·10−15 8.68·10−16 4.01·10−14 2.17·10−13 7.09·10−14 1.92·10−15 1.52·10−13 1.25·10−13
Effective Dose [Sv/(Bq s m−2)] 1.92·10−14 2.38·10−16 1.69·10−14 2.91·10−14 8.84·10−14 6.55·10−14 1.92·10−14 3.49·10−16 3.11·10−14 1.42·10−14 1.05·10−13 9.41·10−14 9.32·10−15 2.44·10−15 1.62·10−14 2.76·10−14 1.07·10−13 4.34·10−14 1.28·10−15 1.33·10−15 3.52·10−16 7.95·10−16 7.06·10−14 1.22·10−13 3.15·10−14 3.94·10−14 2.16·10−15 7.93·10−14 7.25·10−14 9.50·10−18 7.54·10−14 7.75·10−16 1.90·10−14 1.10·10−14 9.94·10−14 1.15·10−14 9.72·10−14 2.69·10−14 1.83·10−15 7.30·10−16 9.28·10−17 3.00·10−16 2.20·10−14 1.15·10−13 5.84·10−14 1.07·10−15 1.13·10−13 3.72·10−14
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Xe-138 Ba-139 Ce-139 Nd-139m Nd-139 Pr-139 Ba-140 La-140 Ba-141 Ce-141 La-141 Nd-141m Nd-141 Pm-141 Sm-141m Sm-141 Ba-142 La-142 Pm-142 Pr-142m Pr-142 Sm-142 Ce-143 La-143 Pm-143 Pr-143 Ce-144 Pm-144 Pr-144m Pr-144 Eu-145 Gd-145 Pm-145 Pr-145 Sm-145 Eu-146 Gd-146 Pm-146 Sm-146 Eu-147 Gd-147 Nd-147 Pm-147 Pr-147 Sm-147 Tb-147 Eu-148 Gd-148 Pm-148m Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 1.07·10−13 6.16·10−14 8.94·10−15 9.17·10−14 3.50·10−14 8.75·10−15 2.52·10−14 1.66·10−13 1.07·10−13 1.02·10−14 6.58·10−14 4.67·10−14 4.24·10−15 8.42·10−14 1.39·10−13 1.27·10−13 8.37·10−14 2.16·10−13 1.44·10−13 0.00 5.67·10−14 6.44·10−15 3.96·10−14 9.64·10−14 1.72·10−14 1.76·10−14 2.93·10−15 8.71·10−14 5.08·10−16 8.43·10−14 8.33·10−14 1.66·10−13 1.22·10−15 4.44·10−14 2.64·10−15 1.43·10−13 1.33·10−14 4.64·10−14 0.00 2.77·10−14 7.67·10−14 1.95·10−14 8.11·10−16 9.75·10−14 0.00 1.27·10−13 1.22·10−13 0.00 1.18·10−13
Effective Dose [Sv/(Bq s m−2)] 5.48·10−14 2.54·10−15 5.97·10−15 7.12·10−14 1.77·10−14 4.75·10−15 8.07·10−15 1.11·10−13 3.92·10−14 3.10·10−15 2.88·10−15 3.45·10−14 2.59·10−15 3.39·10−14 9.07·10−14 6.44·10−14 4.84·10−14 1.37·10−13 4.01·10−14 0.00 3.50·10−15 3.43·10−15 1.21·10−14 5.78·10−15 1.35·10−14 1.94·10−16 7.63·10−16 6.95·10−14 2.20·10−16 2.65·10−15 6.78·10−14 1.09·10−13 5.49·10−16 1.12·10−15 1.26·10−15 1.15·10−13 8.61·10−15 3.34·10−14 0.00 2.14·10−14 5.98·10−14 5.72·10−15 8.67·10−18 3.90·10−14 0.00 7.29·10−14 9.83·10−14 0.00 9.01·10−14
Radionuclide Pm-148 Eu-149 Gd-149 Nd-149 Pm-149 Tb-149 Eu-150b Eu-150a Pm-150 Tb-150 Gd-151 Nd-151 Pm-151 Sm-151 Tb-151 Eu-152m Eu-152 Gd-152 Gd-153 Sm-153 Tb-153 Eu-154 Tb-154 Dy-155 Eu-155 Ho-155 Sm-155 Tb-155 Eu-156 Sm-156 Tb-156m Tb-156n Tb-156 Dy-157 Eu-157 Ho-157 Tb-157 Eu-158 Tb-158 Dy-159 Gd-159 Ho-159 Tb-160 Er-161 Ho-161 Tb-161 Ho-162m Ho-162
6-37 Skin Dose [Sv/(Bq s m−2)] 7.97·10−14 3.09·10−15 2.42·10−14 4.99·10−14 2.19·10−14 1.02·10−13 8.50·10−14 2.05·10−14 1.34·10−13 1.31·10−13 3.25·10−15 9.12·10−14 3.32·10−14 1.90·10−19 5.07·10−14 4.85·10−14 6.90·10−14 0.00 5.00·10−15 1.45·10−14 1.23·10−14 8.29·10−14 1.38·10−13 3.27·10−14 3.39·10−15 3.46·10−14 4.01·10−14 7.29·10−15 9.98·10−14 1.46·10−14 1.11·10−15 3.56·10−16 1.04·10−13 1.94·10−14 3.57·10−14 2.90·10−14 1.06·10−16 1.21·10−13 4.70·10−14 1.89·10−15 1.91·10−14 1.98·10−14 7.34·10−14 5.23·10−14 2.59·10−15 7.69·10−15 3.22·10−14 1.01·10−14
Effective Dose [Sv/(Bq s m−2)] 2.76·10−14 1.95·10−15 1.75·10−14 1.68·10−14 7.08·10−16 7.51·10−14 6.64·10−14 2.22·10−15 6.77·10−14 7.75·10−14 1.88·10−15 4.21·10−14 1.40·10−14 2.46·10−20 3.87·10−14 1.36·10−14 5.28·10−14 0.00 3.11·10−15 2.04·10−15 8.86·10−15 5.75·10−14 1.14·10−13 2.56·10−14 2.14·10−15 1.65·10−14 4.43·10−15 4.84·10−15 6.38·10−14 4.93·10−15 6.24·10−16 9.73·10−17 8.34·10−14 1.48·10−14 1.09·10−14 2.04·10−14 5.34·10−17 5.00·10−14 3.58·10−14 9.93·10−16 2.16·10−15 1.43·10−14 5.19·10−14 4.11·10−14 1.40·10−15 8.93·10−16 2.54·10−14 6.70·10−15
6-38 Radionuclide Tm-162 Yb-162 Ho-164m Ho-164 Dy-165 Er-165 Dy-166 Ho-166m Ho-166 Tm-166 Yb-166 Ho-167 Tm-167 Yb-167 Er-169 Lu-169 Yb-169 Hf-170 Lu-170 Tm-170 Er-171 Lu-171 Tm-171 Er-172 Hf-172 Lu-172 Ta-172 Tm-172 Hf-173 Lu-173 Ta-173 Tm-173 Lu-174m Lu-174 Ta-174 Hf-175 Ta-175 Tm-175 Yb-175 Lu-176m Lu-176 Ta-176 W-176 Hf-177m Lu-177m Lu-177 Re-177 Ta-177
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 1.24·10−13 6.99·10−15 1.93·10−15 8.33·10−15 2.82·10−14 1.61·10−15 5.79·10−15 9.90·10−14 4.46·10−14 1.08·10−13 3.88·10−15 2.95·10−14 1.17·10−14 1.38·10−14 2.83·10−15 5.90·10−14 1.73·10−14 3.00·10−14 1.46·10−13 1.81·10−14 4.22·10−14 3.80·10−14 3.17·10−17 3.21·10−14 5.46·10−15 1.07·10−13 1.16·10−13 5.76·10−14 2.23·10−14 6.45·10−15 5.08·10−14 3.89·10−14 2.89·10−15 6.53·10−15 5.36·10−14 2.15·10−14 5.32·10−14 9.10·10−14 6.93·10−15 2.72·10−14 3.74·10−14 1.25·10−13 8.74·10−15 1.39·10−13 5.89·10−14 7.13·10−15 5.17·10−14 3.36·10−15
Effective Dose [Sv/(Bq s m−2)] 8.50·10−14 4.92·10−15 1.06·10−15 8.03·10−16 1.35·10−15 8.96·10−16 1.21·10−15 7.84·10−14 1.72·10−15 8.78·10−14 2.35·10−15 1.59·10−14 5.39·10−15 9.48·10−15 2.97·10−17 4.75·10−14 1.13·10−14 2.29·10−14 1.21·10−13 3.67·10−16 1.64·10−14 3.00·10−14 1.77·10−17 2.29·10−14 3.40·10−15 8.64·10−14 7.10·10−14 2.30·10−14 1.66·10−14 4.42·10−15 2.55·10−14 1.72·10−14 1.84·10−15 4.94·10−15 2.75·10−14 1.54·10−14 4.24·10−14 4.81·10−14 1.75·10−15 7.65·10−16 2.11·10−14 1.03·10−13 5.98·10−15 9.67·10−14 4.24·10−14 1.50·10−15 2.76·10−14 2.15·10−15
Radionuclide W-177 Yb-177 Hf-178m Lu-178m Lu-178 Re-178 Ta-178b Ta-178a W-178 Yb-178 Hf-179m Lu-179 Ta-179 W-179 Hf-180m Os-180 Re-180 Ta-180m Ta-180 Hf-181 Os-181 Re-181 W-181 Hf-182m Hf-182 Ir-182 Os-182 Re-182b Re-182a Ta-182m Ta-182 Hf-183 Ta-183 Hf-184 Ir-184 Re-184m Re-184 Ta-184 Ir-185 Os-185 Ta-185 W-185 Ir-186a Ir-186b Pt-186 Re-186m Re-186 Ta-186
[Ref. p. 6-42 Skin Dose [Sv/(Bq s m−2)] 5.11·10−14 3.60·10−14 1.36·10−13 9.06·10−14 5.68·10−14 1.04·10−13 5.87·10−14 5.65·10−15 6.09·10−16 1.07·10−14 5.26·10−14 2.99·10−14 1.45·10−15 2.58·10−15 5.82·10−14 3.19·10−15 7.11·10−14 3.67·10−15 3.26·10−14 3.62·10−14 7.03·10−14 4.76·10−14 1.84·10−15 5.82·10−14 1.46·10−14 1.35·10−13 2.46·10−14 1.08·10−13 6.71·10−14 1.93·10−14 7.85·10−14 6.83·10−14 2.62·10−14 3.12·10−14 1.21·10−13 2.19·10−14 5.00·10−14 1.16·10−13 3.52·10−14 4.01·10−14 5.20·10−14 4.52·10−15 9.55·10−14 6.41·10−14 4.10·10−14 7.24·10−16 2.03·10−14 1.49·10−13
Effective Dose [Sv/(Bq s m−2)] 3.91·10−14 8.82·10−15 1.03·10−13 4.80·10−14 7.12·10−15 5.73·10−14 4.32·10−14 4.12·10−15 3.83·10−16 1.62·10−15 3.84·10−14 1.66·10−15 9.00·10−16 1.50·10−15 4.33·10−14 1.96·10−15 5.33·10−14 1.43·10−15 2.35·10−14 2.42·10−14 5.52·10−14 3.37·10−14 1.16·10−15 4.08·10−14 1.03·10−14 6.07·10−14 1.83·10−14 8.49·10−14 5.39·10−14 9.94·10−15 5.99·10−14 3.39·10−14 1.19·10−14 1.04·10−14 8.75·10−14 1.67·10−14 3.99·10−14 7.25·10−14 2.74·10−14 3.18·10−14 8.23·10−15 4.97·10−17 7.51·10−14 4.33·10−14 3.27·10−14 4.14·10−16 9.97·10−16 7.02·10−14
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Ir-187 Re-187 W-187 Ir-188 Pt-188 Re-188m Re-188 W-188 Ir-189 Os-189m Pt-189 Re-189 Ir-190n Ir-190m Ir-190 Os-190m Ir-191m Os-191m Os-191 Pt-191 Ir-192m Ir-192 Au-193 Hg-193m Hg-193 Os-193 Pt-193m Pt-193 Au-194 Hg-194 Ir-194m Ir-194 Os-194 Tl-194m Tl-194 Au-195m Au-195 Hg-195m Hg-195 Ir-195m Ir-195 Pb-195m Pt-195m Tl-195 Hg-197m Hg-197 Pt-197m Pt-197 Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 2.03·10−14 0.00 4.09·10−14 9.18·10−14 1.18·10−14 3.91·10−15 5.35·10−14 2.91·10−15 4.14·10−15 7.16·10−18 1.82·10−14 2.15·10−14 8.89·10−14 7.52·10−18 8.24·10−14 9.12·10−14 4.07·10−15 3.67·10−16 4.35·10−15 1.71·10−14 8.81·10−15 5.53·10−14 9.16·10−15 6.21·10−14 1.26·10−14 2.44·10−14 3.07·10−15 2.07·10−17 6.19·10−14 2.65·10−17 1.34·10−13 5.85·10−14 5.22·10−17 1.47·10−13 4.41·10−14 1.35·10−14 4.12·10−15 1.38·10−14 1.11·10−14 3.53·10−14 2.19·10−14 9.97·10−14 5.92·10−15 7.52·10−14 1.02·10−14 3.35·10−15 1.86·10−14 1.06·10−14
Effective Dose [Sv/(Bq s m−2)] 1.54·10−14 0.00 2.13·10−14 7.52·10−14 7.90·10−15 2.56·10−15 3.13·10−15 1.10·10−16 2.77·10−15 1.24·10−19 1.34·10−14 3.08·10−15 6.81·10−14 1.38·10−19 6.32·10−14 7.03·10−14 2.62·10−15 2.31·10−16 2.78·10−15 1.21·10−14 6.84·10−15 3.61·10−14 6.03·10−15 4.69·10−14 7.70·10−15 3.29·10−15 3.76·10−16 4.07·10−19 4.94·10−14 6.23·10−19 1.04·10−13 4.73·10−15 2.17·10−17 1.03·10−13 3.41·10−14 8.52·10−15 2.73·10−15 8.78·10−15 8.38·10−15 1.78·10−14 2.17·10−15 7.12·10−14 2.44·10−15 5.94·10−14 3.62·10−15 2.26·10−15 3.25·10−15 9.73·10−16
Radionuclide Tl-197 Au-198m Au-198 Pb-198 Tl-198m Tl-198 Au-199 Hg-199m Pb-199 Pt-199 Tl-199 Au-200m Au-200 Bi-200 Pb-200 Pt-200 Tl-200 Au-201 Bi-201 Pb-201 Tl-201 Bi-202 Pb-202m Pb-202 Tl-202 Bi-203 Hg-203 Pb-203 Po-203 Tl-204 Bi-205 Pb-205 Po-205 Bi-206 Tl-206 At-207 Bi-207 Po-207 Tl-207 Tl-208 Pb-209 Tl-209 Bi-210m Bi-210 Pb-210 Po-210 At-211 Bi-211
6-39 Skin Dose [Sv/(Bq s m−2)] 2.43·10−14 3.75·10−14 4.08·10−14 2.66·10−14 7.40·10−14 1.16·10−13 8.23·10−15 2.71·10−14 8.55·10−14 4.38·10−14 1.49·10−14 1.27·10−13 6.36·10−14 1.43·10−13 1.31·10−14 1.13·10−14 7.50·10−14 2.78·10−14 8.99·10−14 4.43·10−14 4.89·10−15 1.57·10−13 1.17·10−13 2.72·10−17 2.63·10−14 1.39·10−13 1.56·10−14 1.87·10−14 1.00·10−13 1.24·10−14 9.70·10−14 2.92·10−17 9.10·10−14 1.90·10−13 3.36·10−14 7.76·10−14 9.31·10−14 7.67·10−14 3.06·10−14 2.34·10−13 9.35·10−15 1.59·10−13 1.63·10−14 2.30·10−14 1.28·10−16 4.81·10−19 1.96·10−15 3.07·10−15
Effective Dose [Sv/(Bq s m−2)] 1.78·10−14 2.39·10−14 1.81·10−14 1.86·10−14 5.26·10−14 9.47·10−14 3.67·10−15 7.63·10−15 6.83·10−14 9.32·10−15 1.02·10−14 9.32·10−14 1.32·10−14 1.08·10−13 8.17·10−15 2.33·10−15 5.98·10−14 2.62·10−15 6.08·10−14 3.35·10−14 3.25·10−15 1.24·10−13 9.29·10−14 4.96·10−19 2.00·10−14 1.13·10−13 1.04·10−14 1.30·10−14 7.59·10−14 1.71·10−16 7.98·10−14 5.45·10−19 7.29·10−14 1.51·10−13 3.95·10−16 6.09·10−14 7.04·10−14 6.08·10−14 4.53·10−16 1.69·10−13 1.00·10−16 9.65·10−14 1.12·10−14 2.58·10−16 4.48·10−17 3.89·10−19 1.37·10−15 2.04·10−15
6-40 Radionuclide Pb-211 Po-211 Bi-212 Pb-212 Po-212 Bi-213 Po-213 Bi-214 Pb-214 Po-214 At-215 Po-215 At-216 Po-216 At-217 At-218 Po-218 Rn-218 Fr-219 Rn-219 Fr-220 Rn-220 Fr-221 Fr-222 Ra-222 Rn-222 Ac-223 Fr-223 Ra-223 Ac-224 Ra-224 Ac-225 Ra-225 Ac-226 Ra-226 Th-226 Ac-227 Pa-227 Ra-227 Th-227 Ac-228 Pa-228 Ra-228 Th-228 Th-229 Pa-230 Th-230 U-230 Pa-231
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 3.06·10−14 4.47·10−16 4.05·10−14 1.35·10−14 0.00 3.39·10−14 0.00 1.28·10−13 2.77·10−14 4.71·10−18 1.12·10−17 1.01·10−17 8.03·10−17 9.57·10−19 1.86·10−17 2.12·10−16 7.56·10−19 4.30·10−17 2.04·10−16 3.38·10−15 8.53·10−16 2.20·10−17 2.02·10−15 4.76·10−14 5.51·10−16 2.28·10−17 3.05·10−16 2.30·10−14 8.87·10−15 1.08·10−14 6.35·10−16 9.40·10−16 3.01·10−15 2.15·10−14 4.79·10−16 6.37·10−16 1.10·10−17 1.08·10−15 3.19·10−14 6.50·10−15 7.88·10−14 6.56·10−14 0.00 1.50·10−16 5.41·10−15 3.73·10−14 4.51·10−17 1.07·10−16 2.44·10−15
Effective Dose [Sv/(Bq s m−2)] 2.59·10−15 3.56·10−16 8.95·10−15 6.24·10−15 0.00 6.16·10−15 0.00 7.25·10−14 1.09·10−14 3.81·10−18 8.51·10−18 7.79·10−18 5.38·10−17 7.75·10−19 1.37·10−17 9.71·10−17 4.21·10−19 3.40·10−17 1.53·10−16 2.46·10−15 4.40·10−16 1.72·10−17 1.32·10−15 5.79·10−16 4.03·10−16 1.77·10−17 1.87·10−16 2.20·10−15 5.47·10−15 8.01·10−15 4.29·10−16 6.37·10−16 2.40·10−16 5.57·10−15 2.84·10−16 3.21·10−16 5.12·10−18 7.38·10−16 7.01·10−15 4.43·10−15 4.49·10−14 5.16·10−14 0.00 8.10·10−17 3.36·10−15 2.91·10−14 1.48·10−17 4.56·10−17 1.57·10−15
Radionuclide Th-231 U-231 Np-232 Pa-232 Th-232 U-232 Np-233 Pa-233 U-233 Np-234 Pa-234m Pa-234 Pu-234 Th-234 U-234 Np-235 Pu-235 U-235 Np-236a Np-236b Pu-236 U-236 Am-237 Np-237 Pu-237 U-237 Am-238 Cm-238 Np-238 Pu-238 U-238 Am-239 Np-239 Pu-239 U-239 Am-240 Cm-240 Np-240m Np-240 Pu-240 U-240 Am-241 Cm-241 Pu-241 Am-242m Am-242 Cm-242 Pu-242
[Ref. p. 6-42 Skin Dose [Sv/(Bq s m−2)] 2.52·10−15 3.82·10−15 6.94·10−14 5.57·10−14 3.44·10−17 5.92·10−17 4.78·10−15 1.66·10−14 4.57·10−17 8.41·10−14 5.48·10−14 1.24·10−13 3.46·10−15 7.50·10−16 4.25·10−17 1.82·10−16 4.78·10−15 8.64·10−15 9.17·10−15 5.76·10−15 4.83·10−17 3.57·10−17 2.14·10−14 1.54·10−15 2.54·10−15 9.97·10−15 5.09·10−14 3.94·10−15 4.31·10−14 4.09·10−17 2.91·10−17 1.56·10−14 1.60·10−14 1.86·10−17 2.61·10−14 5.79·10−14 4.68·10−17 5.93·10−14 9.15·10−14 3.92·10−17 3.12·10−15 1.28·10−15 3.14·10−14 1.17·10−19 1.36·10−16 8.20·10−15 4.29·10−17 3.27·10−17
Effective Dose [Sv/(Bq s m−2)] 4.58·10−16 2.56·10−15 5.38·10−14 4.26·10−14 7.24·10−18 1.17·10−17 3.39·10−15 8.55·10−15 1.42·10−17 6.83·10−14 1.21·10−15 8.72·10−14 2.49·10−15 2.94·10−16 6.11·10−18 4.19·10−17 3.45·10−15 6.46·10−15 4.74·10−15 1.92·10−15 4.68·10−18 3.86·10−18 1.55·10−14 8.87·10−16 1.76·10−15 5.29·10−15 4.04·10−14 2.85·10−15 2.56·10−14 3.50·10−18 2.50·10−18 9.26·10−15 6.95·10−15 3.48·10−18 2.13·10−15 4.67·10−14 4.17·10−18 1.55·10−14 5.88·10−14 3.42·10−18 5.87·10−17 6.74·10−16 2.11·10−14 6.33·10−20 2.49·10−17 6.09·10−16 4.02·10−18 2.90·10−18
Landolt-Börnstein New Series VIII/4
Ref. p. 6-42] Radionuclide Am-243 Cm-243 Pu-243 Am-244m Am-244 Cf-244 Cm-244 Pu-244 Am-245 Bk-245 Cm-245 Pu-245 Am-246m Am-246 Bk-246 Cf-246 Cm-246 Pu-246 Bk-247 Cm-247 Cf-248 Cm-248
Landolt-Börnstein New Series VIII/4
6 External dosimetry Skin Dose [Sv/(Bq s m−2)] 2.75·10−15 9.79·10−15 8.15·10−15 3.11·10−14 5.25·10−14 4.65·10−17 3.91·10−17 2.69·10−17 1.62·10−14 1.58·10−14 5.36·10−15 4.00·10−14 8.56·10−14 6.42·10−14 5.31·10−14 3.35·10−17 3.49·10−17 8.82·10−15 7.43·10−15 1.79·10−14 3.17·10−17 2.67·10−17
Effective Dose [Sv/(Bq s m−2)] 1.85·10−15 5.30·10−15 9.61·10−16 3.63·10−16 3.59·10−14 4.74·10−18 3.40·10−18 2.08·10−18 1.45·10−15 9.26·10−15 3.49·10−15 1.86·10−14 4.74·10−14 3.06·10−14 4.27·10−14 3.92·10−18 3.10·10−18 5.35·10−15 4.20·10−15 1.38·10−14 3.25·10−18 2.35·10−18
Radionuclide Bk-249 Cf-249 Cm-249 Bk-250 Cf-250 Cm-250 Es-250 Cf-251 Es-251 Cf-252 Fm-252 Cf-253 Es-253 Fm-253 Cf-254 Es-254m Es-254 Fm-254 Fm-255 Fm-257 Md-257 Md-258
6-41 Skin Dose [Sv/(Bq s m−2)] 4.07·10−17 1.91·10−14 1.59·10−14 6.43·10−14 3.02·10−17 0.00 2.21·10−14 1.12·10−14 5.35·10−15 3.08·10−17 2.95·10−17 1.66·10−15 4.55·10−17 4.55·10−15 9.83·10−20 3.76·10−14 5.65·10−16 3.43·10−17 3.95·10−16 7.18·10−15 6.20·10−15 1.82·10−16
Effective Dose [Sv/(Bq s m−2)] 4.68·10−19 1.45·10−14 1.02·10−15 4.12·10−14 3.09·10−18 0.00 1.76·10−14 5.01·10−15 3.65·10−15 3.63·10−18 3.45·10−18 1.75·10−17 1.60·10−17 3.12·10−15 1.01·10−20 2.11·10−14 1.57·10−16 4.76·10−18 8.82·10−17 4.15·10−15 4.52·10−15 3.89·10−17
6-42
6 External dosimetry
6.7 References 68Dre 69Sny 69Spi 70Sto 75Emm 75ICR 77ICR 78Sny 79Lic 80Cri 82Kra 83ICR 83Rou 85ICR 85Nel 85Sai 86Gro 88DOE 88ICR 88Zan 89Vei 90Hol 90Sai 91Bri
Drexler, G.: Proceedings of the Symposium on Microdosimetry, Ispra (Italy), 13-15 November 1967. European Communities, Brussels. Report No. EUR 3747 d-f-e, 1968, p. 433. Snyder, W.S., Ford, M.R., Warner, G.G., Fisher jr., H.L.: Medical Internal Radiation Dose Committee (MIRD) Pamphlet No. 5, Supplement No. 3: J. Nucl. Med. 10 (1969). Spiers, F.W.: Delayed effects of bone-seeking radionuclides, Mays, C.W., Jee, W.S.S., Lloyd, R.D., Stover, B.J., Dougherty, J.H., Taylor, G. (eds.), Salt Lake City: University of Utah Press, 1969, p. 95. Storm, E., Israel, H.I.: Nucl. Data Tables A 7 (1970) 565. Emmett, M.B.: Oak Ridge National Laboratory Report No ORNL4972. Oak Ridge, TN, 1975. International Commission on Radiological Protection: ICRP Publication 23. Oxford, UK: Pergamon Press, 1975. International Commission on Radiological Protection: ICRP Publication 26. Oxford, UK: Pergamon Press, 1977. Snyder, W.S., Ford, M.R., Warner, G.G., Fisher jr., H.L.: Medical Internal Radiation Dose Committee (MIRD) Pamphlet No. 5, New York: The Society of Nuclear Medicine, 1978. Lichtenstein, H., Cohen, M.O., Steinber, H.A., Trubetzkoys, E.S., Beer, M.: Computer code manual of the electric power research institutes (MAGI) EPRI-CCM-8. Palo Alto, California: Mathematical application group Inc., 1979. Cristy, M.: Oak Ridge National Laboratory Report No. ORNL/NUREG/TM-367. Oak Ridge, TN, 1980. Kramer, R., Zankl, M., Williams, G., Drexler, G.: GSF-Report S-885. Neuherberg, Germany: GSF-National Research Center for Environment and Health, 1982. International Commission on Radiological Protection: ICRP Publication 38. Oxford, UK: Pergamon Press, 1983. Roussin, R.W., Knight, J.R., Hubbell, J.H., Howerton, R.J.: Report No. ORNL-RSIC-46 (ENDF-335), Oak Ridge, TN: Radiation Shielding Information Center, Oak Ridge National Laboratory, 1983. International Commission on Radiation Units and Measurements: ICRU Report 39. Bethesda, MD: ICRU Publications, 1985. Nelson, W.R., Hirayama, H., Rogers, D.W.O.: SLAC-265-UC-32. Stanford, CA: Stanford Linear Accelerator Center, 1985. Saito, K., Moriuchi, S.: Radiat. Prot. Dosim. 12 (1985) 21. Grosswendt, B, Roos, M.: Medizin Physik (1986) 265. Department of Energy DOE/EH-0070 DOE, Washington DC: Department of Energy, 1988. International Commission on Radiation Units and Measurements: ICRU Report 43. Bethesda, MD: ICRU Publications, 1988. Zankl, M., Veit, R., Williams, G., Schneider, K., Fendel, H., Petoussi, N., Drexler, G.: Radiat. Environ. Biophys. 27 (1988) 153. Veit, R., Zankl, M., Petoussi, N., Mannweiler, E., Williams, G., Drexler, G.: GSF-Report 3/89. Neuherberg, Germany: GSF - National Research Center for Environment and Health, 1989. Hollnagel, R.A.: Radiat Prot. Dosim. 30 (1990) 149. Saito, K., Petoussi, N., Zankl, M., Veit, R., Jacob, P., Drexler, G.: GSF-Report 2/90. Neuherberg, Germany: GSF-National Research Center for Environment and Health, 1990. Briesmeister, J.F.: Los Alamos National Laboratory Report LA-12625-M. Los Alamos, New Mexico, 1977.
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6 External dosimetry 91Gro 91ICR 91Pet 92Hal 92ICR1 92ICR2 92Zan 93Eck1 93Eck2 93Gro 93ICR 93Yam 94Gro 94Gua1 94Gua2 94Hir 94ICR 94Yam 94Zub 95ISO 95Sai 95Til 96Cha 96Dim 96ICR 97Zan 98Cla 98ICR 98Sai 99Sta 99Zan 00Pel
6-43
Grosswendt, B.: Radiat Prot. Dosim. 35 (1991) 221. International Commission on Radiological Protection: ICRP Publication 60. Oxford, UK: Pergamon Press, 1991. Petoussi, N., Jacob, P., Zankl, M., Saito, K.: Radiat. Prot. Dosim. 37 (1991) 31. Halbleib, J.A., Kensek R.P., Mehlhorn, T.A., Valdez, G.D., Seltzer, S.M., Berger, M.J.: Report No SAND-1634 / UC-405. Albuquerque, New Mexico and Livermore, California: SANDIA National Laboratories, 1992. International Commission on Radiation Units and Measurements: ICRU Report 47. Bethesda, MD: ICRU Publications, 1992. International Commission on Radiation Units and Measurements: ICRU Report 48. Bethesda, MD: ICRU Publications, 1992. Zankl, M., Petoussi, N., Drexler, G.: Health Phys. 62 (1992) 395. Eckerman, K.F., Ryman, J.C.: Federal Guidance Report No. 12. Oak Ridge, TN: Oak Ridge National Laboratory, 1993. Eckerman, K.F., Westfall, R.J., Ryman, J.C., Cristy, M.: Oak Ridge National Laboratory Report No ORNL/TM-12350. Oak Ridge, TN: Oak Ridge National Laboratory, 1993. Grosswendt, B.: PTB-Report Dos-22, Braunschweig: Physikalisch-Technische Bundesanstalt, 1993. International Commission on Radiation Units and Measurements: ICRU Report 51. Bethesda, MD: ICRU Publications, 1993. Yamaguchi, Y.: Radioisotopes 42 (1993) 35. Grosswendt, B.: Radiat Prot. Dosim. 54 (1994) 85. Guaraldi , R., Padoani, F.: ENEA Report RT/ERG/94/17, 1994. Guaraldi , R., Padoani, F.: ENEA Report RT/ERG/94/21, 1994. Hirayama, H.: Radiat. Prot. Dosim. 51 (1994) 107. International Commission on Radiation Units and Measurements: ICRU Report 53. Bethesda, MD: ICRU Publications, 1994. Yamaguchi, Y.: Radiat. Prot. Dosim. 55 (1994) 123. Zubal, I.G., Harrell, C.R., Smith, E.O., Rattner, Z., Gindi, G., Hoffer, P.B.: Med. Phys. 21 (1994) 299. International Organization for Standardization: ISO FDIS 4037-3, 1995. Saito, K., Jacob, P.: Radiat. Prot. Dosim. 58 (1) (1995) 29. Till, E., Zankl, M., Drexler G.: GSF-Report 27/95. Neuherberg, Germany: GSF-National Research Center for Environment and Health, 1995. Chartier, J.-L., Grosswendt, B., Gualdrini, G.F., Hirayama, H., Ma, C.-M., Padoani, F., Petoussi, N., Seltzer, S.M., Terrisol, M.: Radiat. Prot. Dosim. 63 (1996) 7. Dimbylow, P.J.: Proc. Voxel phantom development 6-7 July 1996, Dimbylow, P.J. (ed.), Chilton, UK: National Radiological Protection Board, 1996, p. 1. International Commission on Radiological Protection: ICRP Publication 74. Oxford, UK: Pergamon Press, 1996. Zankl, M., Drexler, G., Petoussi-Henss, N., Saito, K.: GSF-Report 8/97. Neuherberg, Germany: GSF - National Research Center for Environment and Health, 1997. Clark, M.J., Chartier, J.-L., Siebert, B.R.L., Zankl, M.: Radiat. Prot. Dosim. 78 (1998) 91. International Commission on Radiation Units and Measurements: ICRU Report 57. Bethesda, MD: ICRU Publications, 1998. Saito, K., Petoussi-Henss, N., Zankl, M.: Health Phys. 74 (6) (1998) 698. Stabin, M.G., Tagesson, M., Thomas, S.R., Ljungberg, M., Strand, S.E.: Appl. Radiat. Isot. 50 (1999) 73. Zankl, M.: Health Phys. 76 (1999) 162. Pellicioni, M.: Radiat. Prot. Dosim. 88 (2002) 279.
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6-44 00Xu 01Zan 02Eck 02Pet 02Roe
6 External dosimetry Xu, X.G, Chao, T. C., Bozkurt A.: Health Phys. 78 (2000) 476. Zankl, M., Wittmann, A.: Radiat. Environ. Biophys. 40 (2001) 153. Eckerman, K.F. Private communication, 2002. Petoussi-Henss, N., Zankl, M., Fill, U., Regulla, D.: Phys. Med. Biol. 47 (2002) 89. Roesler, S., Heinrich, W., Schraube, H.: Radiat. Prot. Dosim. 98 (2002) 367.
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7 Internal dosimetry of radionuclides
7-1
7 Internal dosimetry of radionuclides
This Chapter reviews the behaviour of radionuclides in the body. It summarises the biokinetic and dosimetric models that have been developed by the International Commission on Radiological Protection (ICRP) for assessing radiation doses, and hence risks, from intakes by different routes, including inhalation and ingestion. These models have been widely accepted around the world for use in radiological protection. They have been incorporated in the European and International Basic Safety Standards as well as in many national regulations and guidance notes around the world. Future developments in this area are also examined. Finally, methods that can be used to assess intakes of radionuclides by direct and indirect monitoring procedures and requirements for dose assessment are summarised.
7.1 Introduction People may be exposed to radionuclides in a number of ways. They may be taken into the body as a result of occupational exposure or uptake from the environment. They are used extensively in medical diagnosis and treatment as well as in biomedical research. People may also be exposed externally by submersion in a radioactive cloud. For occupational exposure, the main route of intake is by inhalation, although a fraction of any material deposited in the respiratory system will be transferred to the throat and swallowed, giving the opportunity for absorption in the gastrointestinal tract. Intakes by direct ingestion may occur and some radionuclides may be absorbed through the intact skin. Damage to the intact skin by cuts or other wounds can also result in the entry of radionuclides into the body. For members of the public, the main route of intake of radionuclides will be by ingestion in food and drinking water although intakes by inhalation may also occur, in particular in the case of accidental releases into the environment. For medical applications the method of administration will depend upon the specific nature of the diagnostic investigation or treatment. Knowledge of the behaviour of radionuclides in the body is important for assessing radiation doses resulting from intakes or superficial contamination. For occupational and public exposure the calculation of radiation doses provides the basis for controlling exposures to within accepted limits, for assessing the consequences of the presence of radionuclides in the working or natural environment or determining the need for treatment in the case of accidental intakes. In medical situations radiation doses are needed for optimising diagnostic and treatment schedules. In the case of administration of radionuclides for clinical research, for example on the behaviour of radiopharmaceuticals, assessment of radiation doses is needed for estimating risks for ethical considerations. The radiation dose received by a tissue as a result of the intake of a radionuclide will depend upon a number of factors. These include: the route of intake, the physico-chemical form, its biokinetic behaviour and pathways in the body, organ(s) of accumulation, rate of removal (by physical decay, biological turnover and excretion) and the quality of the emitted radiation (α, β, γ). Biological variation (age, sex, dietary habits etc) will also influence behaviour in the body. Thus the determination of the radiation dose to tissues and the assessment of the possible biological effects resulting from the intake of a particular radionuclide requires a knowledge of all the pertinent physical, physiological, biokinetic and chemical data. Landolt-Börnstein New Series VIII/4
7-2
7 Internal dosimetry of radionuclides
[Ref. p. 7-68
This Chapter reviews the biokinetic behaviour of radionuclides in the body and illustrates the key features through examples. It summarises the models that have been developed by the International Commission on Radiological Protection (ICRP) for assessing doses and hence risks from intakes of radionuclides and examines some future developments in this area. These models have been widely accepted around the world for use in radiological protection. They provide the basis for dose coefficients (doses per unit intakes, Sv Bq−1) for assessing radiation doses from intakes by inhalation and ingestion and have been incorporated in the European and International Basic Safety Standards [96E1, 96I2] as well as in many national regulations and guidance notes around the world. The use of these dose coefficients for assessing the risks from intakes of radionuclides is illustrated. A recent development by ICRP has been the issue of a report giving dose coefficients for the embryo and foetus following intakes of radionuclides by the mother before or during pregnancy [01I1]. A further report is presently being prepared that will give doses to the newborn child from radionuclides consumed in mothers’ milk. The development of the document is summarised. A new dosimetric model for the human alimentary tract is also being prepared. The conceptual basis for this model is reviewed. Finally, methods that can be used to assess intakes of radionuclides by monitoring procedures and the requirements for dose assessment are described in principal; more detailed information is given in Sections 10.3.2 and 10.3.3 of Chapter 10.
7.2 Biokinetics of radionuclides in the body The principal routes by which radionuclides may enter and move around the body and which must be considered in internal dosimetry are summarised in Figure 7.1. Radionuclides passing through the gastrointestinal tract, or deposited in the air passages of the lungs, in a wound or on the outer layer of skin will irradiate these tissues. Soluble forms of radionuclide(s) that are transportable can readily enter the bloodstream and their subsequent fate depends upon their chemical characteristics. If poorly transportable they will only slowly enter the bloodstream or the lymphatic system. Any insoluble particles entering the systemic circulation will be taken up by the reticuloendothelial cells of the liver, spleen and red bone marrow. Here they may remain for up to the life-span of the individual. To facilitate calculation of doses to tissues following intakes of radionuclides, the ICRP has developed a number of generalised biokinetic models to describe their movement and behaviour in the body. Specific models were given by ICRP for adult workers in Publication 30 [79I1, 80I1, 80I2, 88I2] to describe the behaviour of radionuclides in the main organs of intake − the lungs and gastrointestinal tract as well as the skin. For radionuclides that have entered the blood and systemic circulation, activity subsequently deposited in tissues was generally assumed to be uniformly distributed throughout them and therefore the radiation dose depends solely on the organ mass and both the physical half-life and the biological half-time of the radionuclide (see Chapters 3 and 4). A specific model was needed for the skeleton, however, because of the morphology of skeletal bone and the heterogeneous distribution of deposited activity. Biokinetic models were also given in the various parts and supplements of Publication 30 to describe the behaviour of radionuclides in the body after their entry into the blood. More recently, ICRP has provided age-dependent biokinetic models for selected radionuclides in Publications 56, 67, 69, 71 and 72 [89I1, 93I1, 95I1, 95I2, 96I1] and has given dose coefficients (Sv Bq−1) for six ages: 3-month-old infants, 1-, 5-, 10- and 15-year old children and adults. The requirement for age-dependent models and dose coefficients became apparent in the aftermath of the Chernobyl accident when it was realised that, whilst some countries had developed such models there were no models that were generally accepted around the world. Such models are essential for assessing doses to the public from intakes of radionuclides in foods and drinking water, for making comparisons with dose limits and for informing decisions on the acceptability for consumption of foods that may be marketed in many countries. For the development of “age-dependent” models there was a need to include anatomical and physiological information, such as age dependent mass and turnover rate of the skeleton. These, socalled physiologically based models provide a framework in which both human and animal data on the behaviour of radionuclides in the body can be integrated and allow a more realistic approach to the calculation of doses to individuals of different ages. They also have the important advantage that they can take into account excretion and are therefore more appropriate for the interpretation of bioassay data. Landolt-Börnstein New Series VIII/4
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7 Internal dosimetry of radionuclides
7-3
Extrinsic removal Inhalation Skin Lymph nodes
Exhalation
Ingestion
Respiratory tract
Direct absorption
Liver
Gastrointestinal tract
Transfer compartment
Sweat
Wound
Subcutaneous tissue
Other organs
Kidney
Urinary bladder Skin Urine
Faeces
Fig. 7.1. Summary of the main routes in intake, transfers and excretion of radionuclides in the body; [97I2].
In addition to the development of age-dependent biokinetic models a new human respiratory tract model (HRTM) was issued in Publication 66 [94I2]. This model has been applied in all the recent calculations of dose coefficients for workers and the public issued by ICRP (see compilations in 94I1, 96I1, 99I1). Table 7.1 summarises the recent ICRP publications giving revised biokinetic models and dose coefficients. A further development has been the issue of a report giving dose coefficients for the embryo and foetus following intakes of radionuclides by the mother before or during pregnancy. ICRP Publication 88 gives biokinetic models for 31 elements and also dose coefficients for selected radionuclides [01I1]. Presently being prepared by ICRP is a further report that will give doses to the newborn child from radionuclides consumed in mothers’ milk. The development of the document is summarised in Section 7.2.8. A new dosimetric model for the human alimentary tract is also being developed that will also be age-dependent. The conceptual basis for this new model is summarised in Section 7.2.2.
7.2.1 Inhalation A model for describing the deposition and clearance of inhaled radionuclides in adults who are occupationally exposed was first given in Publication 30 of ICRP [79I1]. This lung model separated the respiratory system into three distinct regions, the naso-pharynx (NP), the tracheo-bronchial region (TB) and the pulmonary region (P). It gave information on the deposition of inhaled radionuclides in these regions as a function of the activity median aerodynamic diameter (AMAD) of inhaled particulates and on the rate of clearance of the material from the respiratory system in terms of three default clearance Classes. These had clearance times from the pulmonary part of the respiratory system of days (Class D), weeks (Class W) and years (Class Y). The default particle size was taken to have an AMAD of 1 µm. Information was also given on the transfer of the three Classes of material to lymphatic tissue associated with the respiratory system. A main feature of this lung model is that it calculated only the average dose to the lungs (TB and P regions). Although the model was developed for adults it has been used for younger ages but without any changes to parameter values, other than organ mass.
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7 Internal dosimetry of radionuclides
[Ref. p. 7-68
Table 7.1 Summary of ICRP reports on dose coefficients for workers and members of the public from intakes of radionuclides. ICRP Application Intake Contents Publication No. (year) 56 (1989) Publica Inhalation and Age-dependent systemic models, and tissue dose ingestion coefficients for selected radioisotopes, for H, C, Sr, Zr, Nb, Ru, I, Cs, Ce, Pu, Am and Np. Issued before ICRP Publication 60 [91I1], and hence giving dose equivalents using the tissue weighting factors from ICRP Publication 26 [77I1]. It was also issued before ICRP Publication 66 [94I2] and hence used the ICRP Publication 30 lung model [79I1]. The dose coefficients given in ICRP Publication 56 were superseded by those in ICRP Publications 67 and 71, which used the tissue weighting factors from Publication 60. 67 (1993) Publica Ingestion Age-dependent systemic models, and tissue dose coefficients for selected radioisotopes, for S, Co, Ni, Zn, Mo, Tc, Ag, Te, Ba, Pb, Po and Ra. Updated systemic models are given for Sr, Pu, Am and Np. 68 (1994) Workers Inhalation and Effective dose coefficients for workers, for about 800 ingestion radionuclides: selected radioisotopes of the 91 elements covered in ICRP Publication 30, Parts 1-4 [79I1, 80I1 and I2, 88I2]. The inhalation dose coefficients for workers exposed to 226Ra given in ICRP Publication 68 were revised in Annexe B of ICRP Publication 72. Applies the Human Respiratory Tract Model, HRTM [94I2]. 69 (1995) Publica Ingestion Age-dependent systemic models, and tissue dose coefficients for selected radioisotopes, for Fe, Sb, Se, Th and U. 71 (1995) Publica Inhalation Tissue dose coefficients for selected radioisotopes of elements covered in ICRP Publications 56, 67 and 69, plus Ca and Cm for which age-dependent systemic models are given. Applies the HRTM [94I2]. 72 (1996) Publica Inhalation and Effective dose coefficients for members of the public for ingestion radioisotopes of the 31 elements covered in ICRP Publications 56, 67, 69, and 71, plus radioisotopes of the further 60 elements covered in ICRP Publications 30 and 68. Applies the HRTM [94I2]. CD-ROM Publica Inhalation and A database of equivalent doses to individual tissues (1998) and ingestion corresponding to the effective dose coefficients in ICRP workers Publications 68 and 72. Inhalation dose coefficient for 10 particle sizes. 88 (2001) Embryo Inhalation and Dose coefficients for the offspring for intakes by the mother and foetus ingestion by (worker or public) before or during pregnancy of the mother radionuclides of the 31 elements covered in Publications 68 and 72. CD-ROM2 Embryo Inhalation and Database of dose coefficients extending information on (2002) and foetus ingestion by radionuclides in Publication 88. the mother a Age-dependent dose coefficients (3 months, 1-, 5-, 10-, and 15-years and adult)
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7 Internal dosimetry of radionuclides
7-5
The Human Respiratory Tract Model (HRTM) described in ICRP Publication 66 [94I2] was developed to replace the lung model given in ICRP Publication 30. It takes into account extensive data on the behaviour of inhaled materials that had become available since the ICRP Publication 30 model was developed. As in the earlier model, deposition and clearance are treated separately. The scope of the model was extended to apply explicitly to all members of the population, giving reference values for 3-month-old infants, 1-, 5-, 10- and 15-y-old children and male and female adults. The main features of the model are summarised below. In the new model, the respiratory tract is represented by five regions (Fig. 7.2). The extrathoracic (ET) airways are divided into ET1, the anterior nasal passage and ET2, which consists of the posterior nasal and oral passages, the pharynx and larynx. The thoracic regions are bronchial (BB: trachea, generation 0 and bronchi, airway generations 1-8), bronchiolar (bb: airway generations 9-15), and alveolar-interstitial (AI: the gas exchange region). Lymphatic tissue is associated with the extrathoracic and thoracic airways (LNET and LNTH, respectively). Reference values of dimensions and scaling factors for subjects of different ages are specified in the model. A main feature of the HRTM, compared with the Publication 30 model, is the calculation of doses to these specific tissues in the five regions and allowance for their differences in radiosensitivity.
Posterior Nasal Passage ü ý þ
Pharynx
ET 1 Extrathoracic
Nasal Part Oral Part
ET 2
Larynx
BB Trachea Thoracic Bronchial Main Bronchi Bronchi
Bronchioles
bb
Bronchiolar Alveolar Interstitial
Al
bb Bronchioles Terminal Bronchioles
Al
Respiratory Bronchioles Alveolar Duct + Alveoli
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Fig. 7.2. Respiratory tract regions defined in the Human Respiratory Tract Model; [94I2].
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7 Internal dosimetry of radionuclides
[Ref. p. 7-68
7.2.1.1 Deposition The amount of an aerosol inhaled depends upon breathing parameters and these are influenced by age, body size and level of physical exertion. Deposition of inhaled particles in the HRTM is calculated for each region of the respiratory system, with account taken of both inhalation and exhalation. This is done as a function of particle size, breathing parameters and/or work load, and is assumed to be independent of chemical form. Four standard levels of activity are defined in the HRTM ranging from sleep through to heavy exercise, and the different proportions of time spent at these reference levels are specified for representative individuals at the six standard ages [94I2]. The fraction of inhaled aerosol deposited in the various regions of the lung depends in turn upon the particle size, typically taken as a log-normal distribution. Age dependent default deposition parameters are given for a range of particle sizes from 0.6 nm activity median thermodynamic diameter (AMTD) to 20 µm activity median aerodynamic diameter (AMAD). Previously, a 1 µm AMAD was taken as the default particle size for occupational exposure, but ICRP Publication 66 now recommends 5 µm AMAD as being more typical of the workplace. For members of the public the default is taken as 1 µm AMAD. Table 7.2 compares regional deposition in the respiratory system for the models of ICRP Publications 30 and 66. For the old and new defaults for workers, total deposition is about 30 % higher in the new model (82 % c.f. 63 %), with the extrathoracic region dominating, although a large fraction of this is in the ET1 region and thus unavailable for systemic uptake. Conversely, deposition in the deep lung (bronchiolar and alveolar-interstitial regions) is a factor of four higher in the ICRP 30 model (25 % c.f. 6.4 %). Table 7.2 Comparison of regional deposition for ICRP 30 Lung and ICRP 66 Respiratory Tract Models. Publication 30 model Publication 66 model Adult Adult mem- Worker ber of public 1 µma 1 µma 1 µma 5 µma Region [%] Region [%] [%] [%] Nasal passage (NP) 30 Extrathoracic (ET1) 15 17 34 Extrathoracic (ET2) 19 21 40 Total (34) (38) (74) Trachea and bronchial tree (TB)
8
Bronchial (BB) Bronchiolar (bb) Total
1.3 2.0 (3.3)
1.2 1.7 (2.9)
1.8 1.1 (2.9)
Pulmonary region (P)
25
Alveolar-interstitial (AI)
11
11
5.3
Total
63
Total
48
52
82
a Activity Median Aerodynamic Diameter (AMAD) The variation in deposition parameters between individuals, depending upon age, gender and habits, is an important difference from the Publication 30 model [79I1] for which particle size was the only factor that influenced deposition (see 1 µm entries of Table 7.2). Previously, this distinction was not needed, since the Publication 30 model was intended only for reference adults who were occupationally exposed. In contrast, the ICRP Publication 66 model [94I2] has been designed for application to all members of the population. In the new model, deposition, but not clearance, is strongly influenced by age.
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7.2.1.2 Clearance Subsequent to deposition, material is cleared from the respiratory tract. For material deposited in the anterior nose ET1, clearance is affected extrinsically by such means as nose-blowing or sneezing. The ET1 deposit is cleared directly from the body and makes no subsequent contribution to gut or systemic tissue doses. Removal from all other regions is treated as two competing processes: particle transport (by mucociliary clearance to the throat or translocation to lymph nodes) and absorption to blood. It is assumed that these clearance processes compete independently with each other and have no age or gender dependence. Transport processes include mechanical transport to the gut by mucociliary action and removal by macrophages to the lymph nodes. Particle transport rates are taken to be fixed for all materials and a single compartment model describes clearance by this mechanism (Fig. 7.3). Absorption rates, however, are determined by solubility of inhaled materials and default parameters are recommended for Fast (Type F), Moderate (Type M) and Slow (Type S) absorption. This corresponds roughly to the ICRP Publication 30 classification scheme and chemical forms previously assigned to Class D, W or Y are now provisionally treated as Type F, M or S, respectively. The correspondence between the two schemes is not exact, e.g. the D, W or Y classification refers to whether total pulmonary lung clearance (by absorption to blood or clearance to the throat and then through the gut) is of the order of days, weeks or years, whereas Type F, M or S refers only to the absorption component. The mechanical clearance of the deposited activity is not dependent on the chemical form. The main clearance components for the two models, in the form of approximate biological half-times, are summarised in Table 7.3. Qualitatively, residence times in the lung are reduced, quite drastically, for Type F compared to Class D and elevated for Types M and S relative to Classes W and Y. Anterior nasal Nasooropharynx/ larynx
Extrathoracic LNET
LNTH
ET2′
ETseq
0.01
Bronchi
Bronchioles
ET1
0.001
0.01
BB2
bb2
GI tract
BB1 2
0.03 bbseq
100
Environment
10
0.03 BBseq
1
bb1 0.0001 0.001 0.02
Alveolar interstitial
0.00002
AI3 AI2 AI1
Thoracic
Fig. 7.3 Compartment model representing time-dependent particle transport from each respiratory tract region. Rates shown alongside arrows are reference values in units of d−1. It is assumed that (i) the AI deposit is divided between AI1, AI2 and AI3 in the ratio 0.3:0.6:0.1; (ii) the fraction of the deposit in BB and bb that is cleared slowly (BB2 and bb2) is 50 % for particles of physical size <2.5 µm and decreases with diameter >2.5 µm, and the fraction retained in the airway wall (BBseq and bbseq) is 0.7 % at all sizes; (iii) 0.05 % of material deposited in region ET2 is retained in its wall (ETseq) and the rest in compartment ET2′ which clears rapidly to the GI tract. The model as shown above would describe the retention and clearance of a completely insoluble material. However, there is in general simultaneous absorption to body fluids of material from all the compartments except ET1; [94I1].
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Table 7.3. Comparison of approximate clearance half-times for Publication 30 Lung and Publication 66 Respiratory Tract Model ICRP 30 ICRP 66
Class D W Y
Pulmonary clearance T½ 0.5 d (100 %) 50 d (60 %) 500 d (60 %)
Type F M S
Dominant absorption T½ 10 min (10 %) 140 d (90 %) 7000 d (99.9 %)
By considering the relative rates of the two independent clearance processes (mechanical and absorption) in the HRTM and the amount absorbed in the gut after clearance to the throat it is possible to calculate the fraction of material ultimately transferred to the blood and systemic circulation, both directly from the lungs and indirectly via the gut. Because particle transport rates are fixed for all lung Types, the proportion of material escalating to the gut increases as the classification changes from Type F to Type S. It is interesting to compare the amounts transferred to the circulation for inhalation of the different Types. This is illustrated in Fig. 7.4, ignoring the effect of radioactive decay. For Type F material, such as soluble (transportable) forms of radioisotopes of caesium and iodine (e.g. 137Cs, 131I) there is rapid translocation to the blood with about 25 % of the intakes being taken up by within a day. In contrast for Type S materials, such as 239PuO2, transfer to the blood is much slower with about 0.15 % transferred after 1000 days. Examples of the lung clearance Types adopted for various chemical forms of a selection of radionuclides are given in Table 7.4. Although the HRTM model provides these default clearance Types there is also provision for including material specific absorption parameters when information is available. ICRP has issued a guidance document on this application of the HRTM [02I1].
Fraction of inhaled activity transferred to blood
1 Type F 10-1
Type M
10-2 Type S 10-3
10-4 10-2
10-1
1 10 10 2 Time after intake [d ]
10 3
10 4
Fig. 7.4. Cumulative fraction of inhaled activity absorbed into blood directly from the respiratory tract as a function of time after intake for each default absorption Type (in the absence of radioactive decay), for a reference worker; [02I1].
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Table 7.4 Examples of lung clearance Types adopted by ICRP for workers; [94I1]. Radionuclide Chemical form Inhalation Type Tritium Tritiated water SR2a Cobalt Unspecified compounds, Mb Oxides, hydroxide, halides and nitrate S Strontium Unspecified compounds Fb Strontium titanate S Zirconium Unspecified compounds F Oxides, hydroxide, halides and nitrate M Zirconium carbide S Niobium Unspecified compounds M Oxides and hydroxide S Ruthenium Unspecified compounds F Halides Mb Oxides and hydroxides S Iodine All compounds Fb Vapour Vb Caesium All compounds Fb Cerium Unspecified compounds Mb Oxides, hydroxides and fluorides S Polonium Unspecified compounds F Oxides, hydroxides and nitrate M Radium All compounds M Thorium Unspecified compounds M Oxide and hydroxides S Uranium Most hexavalent compounds, e.g. UFO6, F UO2F2 and UO2(NO3)2 Less soluble compounds, e.g. UO3, UF4, Mb UCI4 and most other hexavalent compounds Highly soluble compounds, e.g. UO2, U3O8 S Plutonium Unspecified compounds Mb Insoluble oxides Sb Americium All compounds Mb Sc Trace contaminant Curium All compounds Mb a Excretion and retention functions for inhalation of 3H2O given in Figure 7.15. b Excretion and retention functions for inhalation of 5 µm AMAD aerosols given in Figures 7.16-7.25. c Trace contaminant formed from 241Pu in matrices of nuclear fuel in insoluble (Type S) forms (Fig. 7.25). 7.2.1.3 Gases and vapours For radionuclides inhaled in particulate form, it is assumed that entry into and deposition in the respiratory tract is governed by the size distribution of the aerosol particles [94I2]. The situation is different for gases and vapours, for which the radionuclide has a specific behaviour at its site of entry into the respiratory tract, depending on the chemistry of the compound. Almost all inhaled molecules contact airway surfaces, but usually return to the air unless they dissolve in, or react with, the surface lining. The fraction of an inhaled gas or vapour that is deposited in each region thus depends on its solubility and reactivity. Generally, however, the regional deposition of a gas or vapour cannot be predicted on a mechanistic basis, from knowledge of its physical and chemical properties, but has to be obtained from an in vivo experimental study. Landolt-Börnstein New Series VIII/4
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The HRTM assigns gases and vapours to three classes: • Class SR-1 (soluble or reactive). Deposition may occur throughout the respiratory tract. Retention in respiratory tract tissues and uptake to the systemic circulation may be less than 100 % of the inhaled activity, although this is the default assumption; e.g. tritium gas and tritiated methane, carbon monoxide. • Class SR-2 (highly soluble or reactive). Total deposition occurs in the extrathoracic airways (ET2). Subsequent retention in the respiratory tract and absorption to body fluids are determined by the chemical properties of the specific gas or vapour; e.g. tritiated water, organically bound tritium and carbon dioxide. • Class SR-0 (insoluble and non-reactive). Negligible deposition in the respiratory tract. External irradiation from submersion in the cloud of gas, and internal irradiation from gas within the respiratory tract. e.g. from all radioisotopes of argon, krypton and xenon (except 37Ar, 94I1). Subsequent retention in the respiratory tract and absorption to body fluids are determined by the chemical properties of the specific gas or vapour. ICRP Publications 68 and 71 as well as the guidance document [94I1, 95I2, 02I1] give information on the assignment of gases and vapours to these three classes, and for selected Class SR-1 compounds information on fractional deposition and subsequent clearance. As an alternative to any of the three default Types defined in ICRP Publication 66, very fast uptake to body fluids (Type V) may be recommended. Although consideration has to be given to the total respiratory tract deposition, regional deposition does not need to be assessed for such materials, since, for the purposes of dose calculation, they can be treated as if they were injected directly into body fluids. Examples are tritiated water and tritiated methane, methyl iodide and methane. 7.2.1.4 Dosimetry In the ICRP Publication 30 lung model, doses to the respiratory system were averaged over 1 kg of lung tissue and the energy of charged particle emissions in the TB, P and respiratory lymph node regions was assumed to be completely absorbed within the lung, i.e. the absorbed fractions were unity for charged particles. In the HRTM model, doses are calculated to several specific regions of the lung and account is taken of variations in radiosensitivity. Absorbed fractions are energy dependent and prescribed functions are given for all source and target combinations and particle types. The target cells identified for the assessment of doses are: basal cells of the epithelium in both extrathoracic regions; basal and secretory cells in the bronchial epithelium; Clara cells (a type of secretory cell) in the bronchiolar epithelium; and endothelial cells, such as those of capillary walls and type II epithelial cells, in the alveolar-interstitial (AI) region. The overall dose to the lung is then taken to be a weighted sum of the doses to the following regions: bronchial, bronchiolar, pulmonary and lymphatic with weighting factors of 0.333, 0.333, 0.333 and 0.001, (the sum is 1), respectively. These weights are known as regional apportionment factors to distinguish them from the tissue weighting factors used in the calculation of effective dose (Chapter 4). They represent the contribution from each region towards the total radiation detriment associated with irradiation of the lung.
7.2.2 Ingestion The model of the gastrointestinal tract (GI) presently used by ICRP to describe the behaviour of ingested radionuclides is that given in ICRP Publication 30 [79I1]. Radionuclides contaminating food or drink, or cleared from the lung by mucociliary action are swallowed, pass down the oesophagus and enter the gastrointestinal (GI) tract, which is treated as four compartments (Fig. 7.5). Absorption is usually described by f1 values which give fractional absorption into the systemic circulation (e.g. f1 = 1.0, absorption = 100 %, f1 = 0.01, absorption = 1 %). The transport of material through the GI tract is described in terms of movement through the four regions. Landolt-Börnstein New Series VIII/4
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Stomach (ST): Its contents are acidic, and little absorption takes place other than for very soluble radionuclides such as caesium or iodine for which absorption from the stomach is assumed to be complete (f1 = 1.0). All other radionuclides are assumed to be absorbed in the small intestine. The residence time for food in the stomach varies from minutes to hours depending on many factors − the amount and composition of food, exercise and emotions. In the dosimetric model the mean residence time is taken to be 1 h. Small Intestine (SI): The principle site of absorption. The contents are alkaline, so that elements which hydrolyse such as rare earths and actinides are not normally readily absorbed. The mean residence time is assumed to be 4 hours. Recommended f1 values are given in ICRP publications for specific radionuclides (e.g. 226Ra: f1 = 0.2, 144Ce: f1 = 0.0005, 239PuO2: f1 = 0.00001). Upper Large Intestine (ULI): Water is absorbed here from the semi-liquid contents. The mean residence time is taken to be 13 hours. Lower Large Intestine (LLI): This region acts as a store for food residues and often forms the critical organ for long-lived non-transportable ingested radionuclides. The mean residence time is taken to be 24 hours.
6WRPDFK 67
6PDOOLQWHVWLQH 6,
%RG\IOXLGV
8SSHU/DUJH ,QWHVWLQH
/RZHU/DUJH ,QWHVWLQH
Fig. 7.5. Biokinetic model for the gastrointestinal tract (based upon 79I1).
Table 7.5 gives the transit time and mass of the contents for the different regions of the GI tract that are assumed for dose calculations. Typical f1 values are given in Table 7.6 for some important elements.
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7 Internal dosimetry of radionuclides
Table 7.5 Regional masses and residence times in the gastrointestinal tract; [79I1]. Portion of GI tract that is the Mass of contents [g] critical tissue Stomach (ST) 250 Small Intestine (SI) 400 Upper Large Intestine (ULI) 220 Lower Large Intestine (LLI) 135 a Transfer rate
[Ref. p. 7-68
ICRP Publication 30 dosimetric model for the Mean residence time [days]
Ka days−1
1/24 2/24 13/24 24/24
24 6 1.8 1
Table 7.6 Examples of f1 values adopted by ICRP for workers; [94I1]. Radionuclide Chemical form f1 Tritium Tritiated water 1.0a Organically bound tritium 1.0a Cobalt Unspecified compounds 0.1a Oxides, hydroxides and inorganic 0.05 compounds Strontium Titanate 0.01 Unspecified compounds 0.3a Zirconium All compounds 0.002 Niobium All compounds 0.010 Ruthenium All compounds 0.05a Iodine All compounds 1.0a Caesium All compounds 1.0a Cerium All compounds 5 × 10−4 a Polonium All compounds 0.10 Radium All compounds 0.20 Thorium Oxide and hydroxides 2.0 × 10−4 Unspecified compounds 5.0 × 10−4 Uranium Unspecified compounds 0.02a Most tetravalent compounds, e.g. UO2, 0.002 U3O8, UF4 Plutonium Nitrate 1 × 10−4 a Insoluble oxides 1 × 10−5 a Unspecified compounds 5 × 10−4 a Americium All compounds 5 × 10−4 a Curium All compounds 5 × 10−4 a Chemical forms of radionuclides for which retention and excretion functions are given in Figures 7.15-7.25. The f1 values recommended for workers are not necessarily appropriate for food and drinking water. Moreover the absorption of radionuclides tends to be greater in the newborn although the results of animal studies suggest that gut absorption decreases as age increases, reaching adult f1 values by about the time of weaning in most cases. An expert group set up by the Nuclear Energy Agency (NEA) within the Organisation for Economic Cooperation and Development (OECD) [88N1] suggested f1 values to be used as average values for the first year of life. The expert group recommended that for fractional absorption values between 0.01 and 0.5 in adults, an increase by a factor of 2 be assumed for the first year of life; but for elements with a fractional absorption in adults of 0.001 or less, a value 10 times that of the adult should be assumed. This general approach has been adopted in the current ICRP documents when more specific data are not available.
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In the dosimetric model for the GI tract doses are calculated to the radiosensitive mucosal cell layer. For low LET radiation (β particles and γ-rays) this is nominally taken to be one half of the average energy absorbed per gram of the contents and for high LET (α) radiation one two-hundredth. The Publication 30 model for the GI tract has a number of distinct limitations in use and although it has been used for calculating doses to infants and children it does not have specific age-dependent parameter values. For this reason a Task Group of ICRP is developing a new model for the Human Alimentary Tract (HAT). The proposed new model is illustrated in Figure 7.6.
Salivary glands
Oral cavity Pharynx
Teeth
Oeso Oeso phagus 1 phagus 2
Oral mucosa
Slow
Fast
B L O O D
Secretory organs
Stomach
Stomach wall
Hepatic artery
Small intest
SI wall
Liver
Left colon
LC wall
Right colon
RC wall
Sigmoid Rectum
SR wall
B L O O D
Portal vein
Fig. 7.6. Proposed structure for the new human alimentary tract model; [03M1].
The revision of the Publication 30 model was motivated by a number of developments: • The 1990 recommendations of ICRP introduced specific risk estimates and tissue weighting factors, wT for radiation-induced cancer of the oesophagus, stomach and colon, requiring dose estimates for each of these regions [91I1]. The Publication 30 model did not include the oral cavity, or the oesophagus and treated the colon as two regions – upper and lower large intestine (Figure 7.5). The new model for the alimentary tract will comprise the oral cavity, including the mouth, teeth, salivary glands and pharynx, the oesophagus, the stomach, the small intestine, including duodenum, jejunum and ileum, the large intestine, including ascending, transverse and descending colon, rectum and anal canal. • Since the development of the ICRP Publication 30 model, a considerable body of data has become available on the transit of materials through the different regions of the alimentary tract. These data have been obtained using non-invasive, mainly scintigraphic techniques and include studies of differences between solid and liquid phases, age and gender related differences and the effect of disease conditions. These data are being used to set default transit rates for the defined regions of the alimentary tract for the six age groups given in ICRP Publication 56 (Table 7.1) [89I1]. • Information has become available for morphometrical and physiological parameters and on the location of sensitive cells in different regions of the alimentary tract. • More information has become available concerning absorption, retention and transfer from different regions of the alimentary tract. • Extensive age-, gender- and health-dependent information is available.
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Development of the new model has been described [03M1]. The new HAT model will be applicable to children and adults under all circumstances of exposure. It considers the movement of radionuclides throughout the alimentary tract from ingestion to elimination. It takes account of sites of radionuclide absorption and retention in the alimentary tract and routes of secretion of absorbed radionuclides into the alimentary tract. Doses will be calculated for sensitive cells in each region: mouth, oesophagus, stomach, small intestine and colon. The new model is more detailed and morphological than the previous gastro-intestinal tract model [79I1]. The new model is physiologically based. It includes consideration of absorption in regions other than the small intestine when such information is available. The model can also be used for radiopharmaceuticals. The model provides the flexibility needed to calculate dose to the alimentary tract for a wide range of exposure conditions and for specific individuals. A gut transfer factor, equivalent to the f1 value, will take account of absorption from the small intestine and from other regions of the alimentary tract where information is available. An important development in the new model is the calculation of doses to sensitive cells in the different regions of the alimentary tract. The location of sensitive epithelial stem cells in the various regions is considered separately; that is for the mouth, oesophagus, stomach, small intestine and colon. Doses from radionuclides in the gut lumen, retained radionuclides and radionuclides in transit to blood are considered. It is expected that the report will be used as the basis for future dosimetric calculations for both ingested radionuclides and radionuclides passed through the throat and swallowed after inhalation.
7.2.3 Cuts and wounds The presence of cuts, abrasions, burns or other pathological damage to the skin may greatly increase the ability of radioactive materials to reach subcutaneous tissues and thence the blood and systemic circulation. Although much of the material deposited at a wound site may be retained at the site, and can be surgically excised, soluble (transportable) material can be transferred to the blood and hence to other parts of the body. These events occur only as a result of accidents, each event will, therefore, be unique and need to be assessed by occupational health physicists and medical staff. To date, ICRP has not given advice on the interpretation of wound monitoring data following accidents involving radionuclides. The biokinetic models that have been developed for various radionuclides are, however, applicable to the soluble component of any deposit in cuts or wounds that enters the blood circulation. Insoluble material will be slowly translocated to regional lymphatic tissue, where it will gradually dissolve and eventually enter the blood. A variable fraction of insoluble material can be retained at the wound site or in lymphatic tissue for the remainder of the individual's life. If particulate material enters the blood it deposits principally in phagocytic cells in the liver, spleen and bone marrow. The United States National Committee on Radiological Protection and Measurements (NCRP) has established a Committee to review the problem of wound dosimetry. The report it is preparing will contain an extensive compilation of human and experimental data on the behaviour of radionuclides at wound sites. Four default categories for wound retention of deposited material have been proposed as summarised below: • • • •
weakly retained (≤10 % retained at one day, <1 % retained at 16 days); moderately retained (11-55 % retained at one day; ≤5 % at 64 days); strongly retained (32-85 % retained at one day; 8-40 % at 64 days); and avidly retained (>80 % retained in one day; >50 % at 64 days).
In addition default categories are also being considered for colloids, particles and fragments. In reviewing the experimental data available, various chemical forms of elements/radionuclides are being allocated to these categories on the basis of either the results of studies in experimental animals or their chemical characteristics [03G1]. Once the report is complete it will need to be reviewed by the NCRP and it is possible these categories may change. The ICRP is awaiting publication of the report before deciding on its future work in this area. Landolt-Börnstein New Series VIII/4
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7.2.4 Absorption through intact skin The intact skin provides an effective barrier against the entry of most radioactive materials into the body, exceptions of practical importance being tritium oxide as liquid or vapour, organic carbon compounds and iodine as vapour or in solution. No generalised model has been adopted by ICRP for estimating absorption of radionuclides through the skin although it would be possible to develop specific models. For example, the behaviour of tritiated organic compounds following direct absorption through the skin would be expected to be significantly different from that after inhalation or ingestion. For skin contamination, both the radiation dose to the area of skin contaminated and the dose to the whole body as a result of absorption need to be considered. ICRP [77I1, 91I1] has recommended that for skin contamination doses should be calculated to sensitive cells, assumed to be at a depth of 70 µm (as a reasonable average value). ICRP [79I1] addressed the uptake of tritiated water vapour by assuming the uptake is instantaneously distributed within body water in the same manner as the inhaled water vapour. That is, for airborne HTO vapour, the dose per unit uptake through the intact skin is the same as the dose per unit activity inhaled. For deposited activity doses are to be calculated as an average to each cm2 of skin tissue. This applies to activity distributed over the skin surface or aggregated in particles. No specific models are recommended by ICRP for calculating doses from β particles deposited on the skin (also see Section 8.2).
7.2.5 Systemic behaviour of radionuclides The fraction of an intake of a radionuclide entering the systemic circulation is referred to as the uptake. In Publication 30 [79I1, 80I1, 80I2, 86I1] ICRP reviewed biokinetic data for each element for use in the calculation of limits on internal exposure to radionuclides by workers for intakes by inhalation and ingestion. Element-specific biokinetic models were given for the distribution and retention of radionuclides following their entry into the blood. The ICRP 30 models applied specifically to workers and not to members of the public. More recently, Publications 56, 67, 69 and 71 have revised the biokinetic models for selected elements and these have been applied in the calculation of dose coefficients for both workers and members of the public [89I1, 93I1, 94I1, 95I1, 95I2] (Table 7.1). If a radionuclide that enters the blood is an isotope of an element that is required by the body then it will follow the normal metabolic pathways for that element (e.g. Na, P, K, Ca, Fe). If it has similar chemical properties to an element that is normally present then it will tend to follow the biokinetic pathways of that element, although its rate of movement between the various compartments in the body may be different (e.g. 90Sr and 226Ra behave similarly to Ca, 137Cs and 86Rb similarly to K). For other radionuclides their behaviour in the body will depend upon their affinity for biological ligands and other transport systems in the body and, as a result, the extent of uptake is unpredictable and must be assessed from the available human or animal data (e.g. 95Nb, 106Ru, 239Pu, 241Am). Radionuclides entering the blood may distribute throughout the body (e.g. 3H, 42K, 137Cs); they may selectively deposit in a particular tissue (e.g. 131I in the thyroid; 90Sr in bone) or they may deposit in significant quantities in a number of tissues (e.g. 239Pu, 241Am, 144Ce). Some examples of the behaviour of selected radionuclides are given below. Limited information is also given on methods of treatment for accidental intakes. More information on decorporation of radionuclides is given in Chapter 9. 7.2.5.1 Elements that distribute widely in body tissues Hydrogen Tritium labelled water (HTO), given either orally or by intravenous injection, is rapidly absorbed from the lungs and absorption from the gut is also essentially complete (f1= 1.0). HTO distributes throughout the body water and is subsequently lost from the body with a biological half-time of about 10 days as a result of excretion in the urine, sweat, faeces and via the lungs (i.e. about 7 % of the total body water is lost per day). The addition of HTO to the body water has been a standard method of determining total body water by isotope dilution. For example, following intravenous injection of 1000 kBq of HTO the activity in a urine sample 6 hours later was 20 Bq ml−1. The total body water is then: Landolt-Börnstein New Series VIII/4
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1,000,000 Bq = 50,000 ml (50 litres) 20 Bq/ml The total body water in Reference Man is 42 litres [75I1]. The rate of loss of tritiated water can be increased by increasing the fluid intake (see Chapter 9). Doubling the fluid intake from 2 to about 4 litres per day can reduce the half-time of tritiated water to about 5-6 days. This rate of loss of tritiated water is also found in countries with a hot climate, such as India. In practice, a small fraction of tritium in body tissues becomes incorporated into organic compounds − amino acids, carbohydrates, etc. − and is retained with a longer half-time. For adults, ICRP [89I1] assumes that this fraction is 0.03 (3 %) and is lost with a half-time of 40 days, while the half-time of 10 days applies to the remaining fraction of 0.97 (97 %). For members of the public, ingesting foods containing tritium, absorption of organically-bound forms and their incorporation into body tissues will lead to longer retention of a larger component. ICRP [89I1] assumes that in the adult, the half-times of 10 days and 40 days apply to equal fractions (0.5) of activity entering blood; values the same as those recommended for organically bound forms of 14C. Total body water =
Caesium 137
Cs, together with 90Sr, is a major component of nuclear fission. As a result of nuclear weapons testing Cs has been injected into the atmosphere and weapons fallout has resulted in the contamination of the food chain and man [77U1]. The first observation of the presence of weapons fallout 137Cs in man was reported in 1956 [56M1]. Since then it has been shown to be present in everyone as a result of contamination of the environment by nuclear test explosions, routine releases from nuclear sites and by the accident at Chernobyl in 1986. 137Cs may enter the body either by inhalation or through the foodchain. Absorption from the gut is almost 100 % (f1 = 1.0, see Section 7.2.2). Once inside the body caesium ions (Cs+) behave very similarly to potassium ions (K+) and are rapidly taken up by cells. There is a considerable concentration factor between the cells and plasma. Generally tissue/plasma concen-tration ratios are the same for K+ and Cs+ but there are a number of exceptions and some variations between different species. In particular, muscle accumulates Cs+ more effectively than K+ and is the main site of long term deposits in the body. At equilibrium the muscle accounts for more than 50 % of the total body 137 Cs in man and bone about 8 %. Accumulation of Cs+ by cells is both by diffusion and by the ion pump that normally accumulates K+. There is a continual turnover of Cs+ in body tissues. For the purposes of dosimetry the retention of 137Cs in man is taken to have two components [89I1]. The first accounts for about 10 % of the administered activity and is excreted mainly in the urine with a half-time(T½) of about 2 days. The second component (90 %) has a half-time of about 110 days in males (range about 50 to 150 days). The long half-time mainly reflects the slow turnover rate in muscle tissue. The retention half-time for the long-term component in females is less than in males, with a mean value for adults of about 60 - 65 days. The use of the ICRP value of 110 days is therefore likely to be conservative for adult females. In children the half time is less than in adults. Thus for a 5 year-old the half- time is taken to be 55 days [89I1]. In cases of accidental intakes, Prussian Blue (ferric ferrocyanide) can increase the rate of excretion of 137 Cs from the body (see Chapter 9). If Prussian Blue is ingested it will accumulate any caesium secreted into the gut preventing it being re-absorbed. The half-time of retention can be reduced to about 40 days; the rate of loss is dependent upon the turnover rate of 137Cs in tissues and its loss into the gastrointestinal tract. 137
Ruthenium 106
Ru is also produced in nuclear fission. Its absorption from the gastrointestinal tract is quite low; the value for gut absorption (f1) is taken to be 0.05. The distribution of 106Ru in mice, rats, monkeys and dogs is fairly uniform throughout all tissues after an initial period during which the kidneys contain the highest concentration. The animal data have been used by ICRP [89I1] to define a retention function for Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-17
ruthenium in man. Any ruthenium entering the blood is assumed to be distributed throughout the body with retention components of 35 % (T½ = 8 days); 30 % (T½=35 days) and 20 % (T½=1000 days). The remaining activity entering the blood (15 %) is taken to be promptly excreted. Because of the relatively short retention time in the body of most of the intake and a physical half-life of 368 days, the effective dose from ingestion of 106Ru is dominated by the radiation dose to the large intestine from beta-particles emitted while it passes through the GI tract. 7.2.5.2 Elements that deposit mainly in particular organs or tissues Iodine Radioactive isotopes of iodine are important components of nuclear fission, particularly in the first few days and weeks after a release into the atmosphere. If taken into the body they are accumulated by the thyroid gland, as was demonstrated particularly after the accident at Chernobyl. The most important isotope is 131I which has a physical half-life of 8.04 days. Radioactive isotopes of iodine are also widely used in medicine. They are used to demonstrate changes in thyroid function, to treat hyperthyroidism or to kill tumour cells in the treatment of thyroid cancer. The thyroid gland consists of a bi-lobed body in the neck region. It produces the hormones thyroxine and tri-iodothyronine which are important for regulating the body's metabolic rate. Disorders of the gland can result in either an under- or over-active gland (hypo- or hyperthyroidism). The gland weighs about 20 g in the healthy adult (2 g in the newborn child) and is made up of 20 to 40 million spherical vesicles (follicles) per lobe. Each follicle is surrounded by a single layer of cuboidal epithelial cells (acinar cells) lying upon a basement membrane and in proximity to numerous blood capillaries. The vesicles are filled with a structureless semi-fluid material − the so-called “colloid” which contains the active component of the gland (a protein-storage form of hormone). When the thyroid is quiescent, colloid is abundant and the follicles large (about 300 µm in diameter). When the thyroid is active colloid is scanty and follicles small (about 50 µm in diameter). The gland contains about 10,000 µg of iodide in the average normal adult. For adults in Europe about 225 µg of stable iodide enters the extracellular (iodine) space from the diet each day, absorption occurring across the small intestine within 1-2 hours. About 70 µg of iodide per day is trapped by the thyroid and converted to thyroid hormones while most of the rest is excreted in the urine. The amount of ingested iodide that is taken up by the gland is thus about 30 % [83S1, 87S1]. The fractional absorption varies between different individuals and there are significant differences between countries because of varying levels of stable iodine in the diet. For the purpose of dosimetric modelling ICRP has recommended that the uptake of radioiodine by the gland should be taken to be 30 %. The iodide synthesised into hormone leaves the gland with a half-time of about 80 days (adult) and enters other tissues. From this source most of the iodine (about 80 %) is metabolised back to free iodide with a half-time of about 8 days and re-enters the iodide space, the rest is excreted in the faeces. In adults the total amount of stable iodine excreted is approximately equal to the amount absorbed. Iodide or elemental radioactive iodine (e.g. 131I) may be ingested (f1 = 1.0) or the volatile compounds inhaled. As the gland is small, and it takes up about 30 % of radioiodine entering the blood, the concentration of radioiodine in the gland, and hence the radiation dose, is more than a thousand times that to other tissues. The turnover of stable iodine, and hence radioiodine, is low (T½ = 80 days in adults) and thus short-lived isotopes (e.g. 131I T½ = 8.0 days) will decay mainly in the gland rather than being returned to the blood. In children, although the turnover rate is faster the mass of the gland is smaller and hence for a similar intake of radioiodine the dose can be higher. Milk consumption is the most important pathway for the uptake of radioiodine from the human food chain after a release of radioiodine into the environment and children have a high consumption of milk. As a consequence children are the most sensitive (critical) group following such a release. Various drugs have been used to reduce the uptake of radioiodine into the thyroid gland after an intake. The safest and most effective procedure is to administer a large single oral dose (20-200 mg) of potassium iodide or iodide (see Chapter 9). It is effective within an hour and reduces the subsequent uptake of radioiodine into the gland. The daily intake is suddenly increased from about 225 µg to Landolt-Börnstein New Series VIII/4
7-18
7 Internal dosimetry of radionuclides
[Ref. p. 7-68
20-200 mg. However, the uptake of iodide into the thyroid remains at about 70 µg d−1 so that only a small fraction of the stable iodine and hence radioiodine present in the iodide space is then transferred to the gland. Thus, if 20 mg of stable iodide is given promptly the fractional uptake of radioactive iodine by the gland can be reduced to less than 0.001 (0.1 %). Since the half-time of iodide in the iodide space is about 10 hours, with rapid uptake by the gland or excretion from the body, the effectiveness of administering large amounts of stable iodide diminishes the greater the delay; after 48 hours, it is of little value at all. It has no effect on any radioiodine that has been taken up by the gland. Administration of stable iodide is the treatment of choice for accidental intakes of radioiodine and is recommended by the World Health Organisation [99W1]. Alkaline earth elements The bone is a highly specialised form of connective tissue. It consists of a soft organic matrix of collagen and ground substance in which is deposited calcium hydroxyapatite (Ca10 (PO4)6 (OH)2). The cells of bone consist of a proliferating population of stem cells which differentiate into: osteocytes − the cells responsible for bone maintenance; osteoblasts − the cells responsible for bone formation; and osteoclasts − the cells responsible for bone removal. There are two types of bone: • Hard cortical bone, which makes up about 80 % of the bone mass and which is penetrated by blood vessels and Haversian systems; and • Trabecular bone or spongy bone which makes up the remaining 20 % of the bone mass. The trabeculae are generally 100-200 µm in diameter and usually do not contain blood vessels. The spaces between the trabeculae are large (up to 1000-2000 µm) and contain the vascular (red) marrow. The surface area of trabecular bone is estimated to be about 4 times that of cortical bone [79I1]. The surfaces of bone are covered with non-mineralised layers of connective tissue. This is the periosteum on the external bone surface and the endosteum on the internal surface. The processes of bone turnover, resulting from the laying down of new bone and removal of old bone, continue throughout life although slowing down with increasing age. Calcium is thus an important component of the skeleton and other alkaline earth elements can be substituted for it in the bone matrix. A number of substitutions are possible in the lattice structure without disturbing the symmetry of the crystal lattice. ICRP has developed age-dependent biokinetic models for the alkaline earth elements (Ca, Ba, Sr and Ra) [93I1]. These models allow for the recycling of radionuclides between the skeleton, blood and soft tissues. They are physiologically based and take account of bone growth and remodelling as a function of age. They can also be applied to the interpretation of bioassay data. These models are essentially dynamic in nature. Material initially deposited on bone surfaces may be buried in bone volume by the formation of new bone or resorbed to red bone marrow. Activity in marrow is either locally or systemically recycled back to bone surfaces, and is also replenished by resorption from bone volume. A fraction will also be excreted from the body. The rate at which some of the processes operate can depend on the type of bone, as well as age, and allowance is made for this by division into cortical and trabecular compartments. This division also accommodates age-dependent dosimetry − in adults most of the active marrow is associated with trabecular bone, whereas for children this is more evenly distributed between both types of bone. 7.2.5.3 Elements that deposit in a number of tissues Plutonium, americium and curium Plutonium and other higher actinides are produced in thermal reactors, although the relative amounts generated depend upon the irradiation time. At low irradiation times almost pure 239Pu\240Pu are produced (subsequently given as 239Pu). As the irradiation time increases other isotopes of plutonium (e.g. 241Pu) Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-19
and higher actinides (e.g. 241Am, 242Cm, 244Cm) are produced. 239Pu is the isotope that has been processed in greatest quantity and for which most biological data are available. 238Pu is also an important isotope and is used as a power source in satellites and in cardiac pacemakers. Plutonium metal is highly reactive, oxidising in moist air if present in a finely divided form. The oxide is chemically inert and insoluble, particularly if produced at a high temperature (1000 °C). Am and Cm metals readily react with oxygen to produce oxides which are much more soluble than PuO2. In the processing of nuclear fuel soluble nitrates and other soluble complexes are formed. The behaviour of plutonium in the body depends upon its chemical form. Plutonium dioxide is chemically inert and largely insoluble in biological fluids, particularly if produced at a high temperature. Soluble plutonium compounds (e.g. plutonium nitrate, plutonium citrate) readily hydrolyse at the pH of biological fluids. Following hydrolysis, there is a strong tendency to polymerise, forming a colloidal insoluble compound. Alternatively, soluble compounds can react with naturally occurring complexing agents in body fluids which can readily move around the body. Which of these reactions predominates depends upon the site of entry of plutonium. If into blood, then complex formation is likely; if into a wound or the lung, then colloidal polymers are likely to be formed. The main route of entry of plutonium into the body is by inhalation, although it can also enter through cuts and wounds. In adults very little is absorbed from the gut (ICRP has adopted gut absorption, f1, values of 5 × 10−4 for soluble plutonium compounds and plutonium in foodstuffs, 1 × 10−4 for plutonium nitrate, and 1 × 10−5 for plutonium oxide, 239PuO2) or across the intact skin. For americium and curium ICRP has adopted an f1 of 5 × 10−4 for all compounds [94I1, 96I1]. After inhalation, deposition of plutonium in the lungs is determined by particle size as detailed in Section 7.2.1.1. Subsequent clearance from the lungs depends upon its chemical form. Whatever the chemical form inhaled, a fraction, consisting of any soluble material will be rapidly transported to blood and this is excreted through the kidneys or deposited in tissues (mainly the liver and skeleton). The remaining material, consisting of colloidal polymers or material with a low solubility (e.g. PuO2), is initially retained in the lungs. Material retained in the lungs is largely taken up by scavenger cells (macrophages) in the lungs. These cells may migrate to lymph nodes or reach the muco-ciliary escalator, be swallowed and excreted in the faeces. Alternatively materials in macrophages may gradually dissolve and translocate to blood. In studies with experimental animals retention half-times of the long-term component of retention have varied from 100 to 1000 days or more although soluble compounds are cleared more rapidly. The relative proportions of the long and short retention components depend upon the material initially deposited. For example, in the case of a polydisperse aerosol of high temperature calcined plutonium dioxide deposited in the lungs the amount rapidly moving to blood is normally less than 0.4 % and would be treated as default absorption Type S in the HRTM (Section 7.2.1.2). For a plutonium nitrate aerosol it may be greater than 20 % and would be treated as Type M. Compounds of americium and curium, particularly the oxides, are more soluble than plutonium compounds in the respiratory system and more readily absorbed. They would generally be classified as Type M but any americium or curium trapped in insoluble particles of PuO2 would behave as Type S. The principal sites of deposition of Pu in the body after translocation to the blood are the liver and skeleton. This is also the case for Am and Cm. All three actinides are classified as bone-surface seeking elements. That is, they are assumed to be uniformly distributed on endosteal bone surfaces of cortical and trabecular bone after their deposition in the skeleton. In practice, surface deposits of Pu, Am and Cm have been shown to be progressively buried by the formation of new bone. In addition, activity is lost from bone surfaces during resorption and some transfer to bone marrow takes place, particularly for Pu. In ICRP Publication 67, a dynamic model has been adopted to describe their behaviour in the body and to take account of their movement in bone as well as between tissues and excretion (Fig. 7.7 [93I1]).
Landolt-Börnstein New Series VIII/4
7-20
7 Internal dosimetry of radionuclides
Other soft tissue
Intermediate turnover (ST 1)
Slow turnover (ST 2)
Rapid turnover (ST 0)
Skeleton
Liver Cortical surface
Cortical marrow
Trabecular surface
Trabecular marrow
Cortical volume
Liver 2 Liver 1
Blood Trabecular volume
[Ref. p. 7-68
GI tract contents Kidneys Urine
Urinary bladder contents
Other kidney tissue
Gonads
Faeces
Urinary path
Fig. 7.7. Diagram of the biokinetic model for plutonium and americium; [93I1].
Comparisons of relative skeletal retention in the biokinetic model are complicated by the compound structure of the skeleton and recycling between compartments which gives rise to several retention components. However, calculation shows that after short-term losses are complete it is possible to discern an effective retention half-time in the skeleton of nearly 100 years (Fig. 7.8). The figure also gives the retention of plutonium in the skeleton for 3-month-old infants and 10-year-old children. Because of the faster turnover of bone at younger ages, the initial uptake of plutonium by the skeleton is greater although the rate of loss is also faster. Fig. 7.9 gives comparable data for the liver. Although there are differences in the initial uptake, reflecting the effect of age and deposition in the skeleton, overall liver retention is similar for all age groups. The peak in uptake at 5-10 years reflects the uptake of activity lost to the blood from the skeleton. 60
60
Infant Age 10y Adult
50
Infant Age 10y Adult
Injected Pu in liver [%]
Injected Pu in skeleton [%]
80
40
40 30 20
20
10
0
10
30 20 Years after injection
40
50
Fig. 7.8. Model predictions of the plutonium content of the skeleton as a function of age and time after injection.
0
10
30 20 Years after injection
40
50
Fig. 7.9. Model predictions of the plutonium content of the liver as a function of age and time after injection.
Intravenous injection of the complexing agent diethylenetriaminapentaacetic acid (DTPA) as the calcium zinc salt is the only accepted therapeutic method for removing soluble forms of actinides from the body. It forms chelate complexes with these actinides which can be excreted in the urine and hence Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-21
can effectively clear them from the systemic circulation and some that has recently deposited in bone and other tissues. It is unable to remove intracellular deposits or activity that has been buried in bone and must, therefore, be administered soon after an intake. It can remove some soluble complexes from the lungs. Local injection of DPTA into contaminated wounds can remove more soluble forms of plutonium from the body than the same amount given intravenously, provided the DPTA completely infiltrates the wound site. Further information on DTPA treatment is given in Chapter 9.
7.2.6 Excretion In the biokinetic model described in ICRP Publication 30, no specific information was given on excretion in urine and faeces, although the models were used in Publication 54 [88I1] for interpreting excretion data. In the 1990 recommendations of the ICRP [91I1], however, the urinary bladder and the colon are given explicit tissue weighting factors wT (see Section 7.3.1) and the revised biokinetic models given by ICRP now give specific information on excretion pathways in urine and faeces [93I1]. For assessing doses from systemic activity lost into the faeces, the model for the gastrointestinal tract is used (Section 7.2.2) assuming secretion of radionuclides from the blood into the upper large intestine. A model for the urinary bladder has been adapted for calculating doses to the bladder wall [93I1]. The bladder is taken to be of fixed size containing 15, 25, 65, 75, 85 and 115 ml of urine in the 3-month-old, 1-, 5-, 10-, 15-yearold children and adults, respectively. These volumes represent the average content of the bladder during the time period between voids. The rate at which radionuclides enter the bladder is based on their elimination rates from body tissues and the urine to faecal excretion ratio adopted for the biokinetic data for each element. For some elements, the biokinetic data directly address excretion. The number of voids per day for the 3-month-old and 1-year-old are taken as 20 and 16, respectively. For all other ages, 6 voids per day are assumed.
7.2.7 Embryo and foetus During pregnancy, radionuclides that have entered the mother's body, either before or after conception, can irradiate the developing embryo and foetus. The radiation dose to the offspring will depend upon a number of factors. These include: • the transfer of radionuclide(s) to the developing offspring from maternal blood and from deposits in the tissues of the mother; • the distribution and retention in foetal tissues; • the physical half-life and formation of decay products; • growth of the offspring; and • photon irradiation from radionuclides in the placenta and maternal tissues. Radiation doses will also be received by the newborn child from radionuclides retained at birth. During the foetal period of development, radionuclides can cross the placenta to reach the tissues of the embryo and foetus from the maternal circulation. The processes involved in this transfer may include simple diffusion, facilitated transport, active transport, movement through pores and channels, and pinocytosis [83S1, 87S1]. Most human data on the placental transfer of radionuclides are available from studies with labelled metabolites, radiopharmaceuticals and other radionuclides used in nuclear medicine, although some data are also available for radionuclides in fallout from weapons testing or for radionuclide releases into the environment as a result of nuclear accidents (e.g. 90Sr, 131I, 137Cs). Analysis of autopsy samples has also given information on both naturally occurring and artificially produced radionuclides. Information is additionally available on levels of stable elements in the placenta and foetal tissues that can be compared with those in the adult. The rather limited amount of human data available has made it Landolt-Börnstein New Series VIII/4
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7 Internal dosimetry of radionuclides
[Ref. p. 7-68
essential to use the results of animal studies in the development of dosimetric models for man, although even here information can be very limited for many elements. Chemical analogy is also of value in model development. ICRP has issued Publication 88 [01I1] giving dose coefficients for the embryo and foetus following intakes of radionuclides by the mother. It covers selected radionuclides of the 31 elements covered in Publications 56, 67, 69 and 71 (Table 7.1) and applies to the offspring of both members of the general public and workers. In the development of biokinetic and dosimetric models, two approaches have been used. Where sufficient information is available, element-specific models have been given. This applies, for example, to tritiated water, caesium, iodine and the alkaline earths. When appropriate human data are not available, animal studies have provided the main basis for model development using a generic modelling approach. It has been assumed, in the absence of more specific information, that the dose to all tissues of the embryo, taken to be up to the end of 56 days after conception, (i.e. the end of the second month of gestation), can be approximated by the dose to the maternal uterus. All organs and tissues of the developing embryo thus receive the same dose. The general approach that has been adopted for calculating equivalent doses to the organs and tissues of the developing foetus from experimental studies in animals is to use average concentrations of a radionuclide in the foetus (CF) and mother (CM) obtained shortly after injection. Where possible the value of the CF:CM concentration ratio adopted has been based on results obtained in a number of different animal species. The total activity transferred to and retained in the foetus from 57 days of gestation to birth at 38 weeks (266 days) is calculated for each radionuclide from the CF:CM ratio. This ratio at the time of the intake is assumed to stay constant for the remaining period of the pregnancy. This is taken to be a conservative assumption. Some examples of (CF:CM) concentration ratios adopted are given in Table 7.7. The concentration ratio may depend upon the time of the intake in relation to the start of the pregnancy. Thus for an acute intake of plutonium at any time before pregnancy the CF:CM ratio is taken to be 0.03; this ratio is then maintained at this value through the period of gestation. For an acute intake during the first trimester of pregnancy, however, a ratio of 0.1 is adopted, subsequently increasing to 0.3 for an intake at the end of the second trimester (180 days) and 1.0 for an intake at term (266 days). Again this ratio is assumed to be kept constant over the remaining period of the pregnancy. Table 7.7 Concentration ratios for elements in the foetus and mother (CF:CM) following intakes by the mother before or during pregnancy and corresponding ratios for the placenta (CPl:CM)
Element H in HTO Organic carbon Sulphur Zinc Zirconium Ruthenium Caesiuma Cerium Plutonium Americium
CF:CM Intakes prior to pregnancy 1.6 1.5 1 2 0.2 0.01 1 0.01 0.03 0.01
Intakes during pregnancy 1.6 1.5 2 2 0.2 0.2 1 0.05 0.1/0.3/1b 0.1
CPl:CM 1 1.5 2 1 1 0.1/2c 1 0.1/1c 0.1/5c 0.1/2c
a Half-time of long-term component in mother during pregnancy taken to be 50 days b Intakes in 1st and at the end of the 2nd/3rd trimester (see text) c Intakes before/during pregnancy Dosimetric models were developed by ICRP that allowed for the calculation of doses to the embryo and to foetal tissues from radionuclides deposited either in the tissues of the embryo/foetus, in the placenta or in the mother. To provide data that could be used for assessing a range of possible intake scenarios dose coefficients have been given for acute and chronic intakes by the mother at various times Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-23
both before conception or during pregnancy. Dose coefficients for the offspring following ingestion of radionuclides by the mother are given for a range of f1 values in Publication 88, while dose coefficients for the offspring after inhalation of radionuclides by the mother are given for both 1 and 5 µm AMAD aerosols (the default particle sizes for members of the public and workers, respectively) and appropriate lung absorption Types. For acute exposures, intakes of radionuclides are taken to occur at the time of conception and after 5, 10, 15, 25 and 35 weeks of the pregnancy and at 6 months and 2½ years before conception. For chronic exposures, intakes are taken to occur during the year of pregnancy, starting from conception and for 1 year or for 5 years prior to conception. This range of intake scenarios was selected to allow doses to the offspring to be calculated for any pattern of intake by the mother. Equivalent doses to the date of birth have been given in Publication 88 for the brain, for the most sensitive 8-15 weeks of gestation, and for the tissue receiving the highest dose. The effective dose to birth has also been given using the wT values recommended by ICRP in Publication 60 [91I1]. Whilst these values are not strictly appropriate for exposures in utero, they have been used as no alternative weighting factors are available and the calculation of effective dose provides a useful quantity for comparison with doses to the reference adult. Effective doses (to age 70 years) received after birth are also given, together with the total committed effective dose (before and after birth) received by the offspring. The total committed effective dose to the offspring, eoffspring due to maternal intake of the radionuclide is the sum of the effective dose received during the in utero period, ein utero and the committed effective dose during the subsequent 70 years of post-natal life, epostnatal. That is: eoffspring = ein utero + epostnatal
(1)
where ein utero =
∫
8w 0
h&uterus (t )dt +
∑w ∫ T
T
38 w 8w
h&T (t )dt
(2)
and e postnatal =
∑w ∫ T
T
70 y birth
h&T (t )dt
(3)
where the limits of integration in the first two integrals are in weeks and in the last term is in years. The value h&T is the equivalent dose rate to individual tissues of the offspring during foetal life and after birth. In the case of the embryo the dose to the tissues of the uterus is taken as a surrogate for the dose to the embryo. In conjunction with the preparation of ICRP Publication 88 a CD-ROM has been issued which gives much more detailed information than in the publication [02I2]. In addition to the doses given in Publication 88 it provides equivalent doses to 15 organs and tissues in the foetus as well as an average equivalent dose to the remainder tissues. It also gives doses to the offspring to a number of times after birth (10, 20, 40, 70 years). Additionally, doses have been given for a range of ten inhaled particle sizes. Publication 88 gives only dose coefficients to the offspring and no information is provided on comparative doses to the adult. Such a comparison has, however, been published [02S1]. The main findings are that, in general, doses to the offspring are similar to or less than those to the reference adult. For a few radionuclides the dose to the offspring can exceed that to the adult. This applies to 3H, 14C, 35S and 59Fe, to radioisotopes of I and to radioisotopes of the alkaline earth elements including 45Ca, 89Sr, 90 Sr, 224Ra and 226Ra. For radioisotopes of iodine and the alkaline earth elements, the greatest doses result from intakes during the last trimester of the pregnancy when there is the greatest foetal demand for iodine and calcium. Whilst in most cases the doses to the offspring for the radionuclides covered in Publication 88 exceed those to the reference adult by a factor of about 2 to 3, in the case of some bone-seeking radionuclides the difference can be around a factor of 10 for intakes of short lived isotopes towards the end of the period of pregnancy. Some illustrative dose coefficients for the offspring following inhalation of 137Cs by the mother (as a member of the public) are given in Table 7.8. In this case the doses to the Landolt-Börnstein New Series VIII/4
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7 Internal dosimetry of radionuclides
[Ref. p. 7-68
offspring are highest for intakes early in the pregnancy as they reflect doses received by the mother over the period of the pregnancy. For intakes later in the pregnancy doses are lower as, although some 137Cs will be retained in the newborn child, the half-time of retention is less than in the mother. Table 7.8 Dose coefficients [Sv Bq−1] for the offspring from acute intakes by inhalation of 137Cs (T ½ = 30 y) by the mother, as a member of the public (AMAD = 1 µm, absorption Type F, f1 = 1.0). Reference adult = 4.6 × 10−9 Sv Bq-1 Scenario [weeks]a −130b −26 0c 5 10 15 25 35 a b c d na
Highest organ dose (in utero)d 7.0 × 10−13 6.2 × 10−10 2.5 × 10−9 2.4 × 10−9 2.3 × 10−9 2.2 × 10−9 1.7 × 10−9 6.1 × 10−10
Brain (8-15 weeks) 1.9 × 10−13 1.7 × 10−10 6.7 × 10−10 1.1 × 10−9 9.4 × 10−10 na na na
ein utero 7.0 × 10−13 6.2 × 10−10 2.5 × 10−9 2.4 × 10−9 2.3 × 10−9 2.2 × 10−9 1.7 × 10−9 6.1 × 10−10
epost natal 1.1 × 10−14 9.6 × 10−12 1.5 × 10−11 2.5 × 10−11 4.1 × 10−11 6.7 × 10−11 1.8 × 10−10 4.7 × 10−10
eoffspring 7.1 × 10−13 6.3 × 10−10 2.5 × 10−9 2.4 × 10−9 2.3 × 10−9 2.3 × 10−9 1.9 × 10−9 1.1 × 10−9
Intake at indicated time; negative times are prior to pregnancy −130 weeks = acute intake 2.5 years before conception 0 = acute intake at time of conception For 137Cs all tissues receive the same dose Not applicable
7.2.8 Transfer in maternal milk Models are presently being developed by ICRP for the transfer of radionuclides to mothers’ milk that will allow the calculation of dose coefficients for intakes by the offspring [03H1]. The publication will cover a review of biokinetic data relevant to an assessment of the transfer of radionuclides to breast milk following intake by the mother, the development of models, and the calculation of doses to the newborn child resulting from the transfer of radionuclides to milk after inhalation or ingestion by the mother. It is assumed that lactation lasts for a period up to 6 months after birth and that milk consumption increases to 800 ml d−1 over the first week and then remains constant to the end of lactation. Doses to the infant will be given for a range of intake scenarios. It is proposed that in the publication dose coefficients will be given for acute intakes by the mother at 26 weeks before conception, for intakes during pregnancy at 5, 15, and 35 weeks after conception and for intakes after birth at 1, 10 and 20 weeks of age. In addition doses from protracted exposures throughout pregnancy and lactation will also be included. These dose coefficients should give a sufficient amount of information to understand the implications for doses to the offspring for intakes at various times either before or after birth. Data for additional acute intake times and for chronic exposures as well as for inhalation of a range of particle sizes will be included on a CD-ROM. The dose coefficients for intakes by the 3-month-old infant given in previous publications [96I1] will be used to calculate the doses from the intakes by the suckling infant in milk. Some preliminary information is given in Table 7.9 (from 03H1) which gives the ratios of infant (offspring) dose to adult dose for chronic intake of various radionuclides throughout pregnancy and lactation. The values for lactation include contributions from activity retained in the mother from intakes during pregnancy as well as transfer to milk from intakes by her during lactation. The results of preliminary model calculations showed that intakes during pregnancy contribute an estimated 15 % of activity in milk for 137Cs, 210Po and 241Am, about 10 % for 45Ca and 90Sr, 4 % for 239Pu and negligible amounts for 131I. Doses to the infant from milk consumption are estimated to exceed adult doses in the cases of 45Ca and 131I. Very similar ratios of infant to adult dose are obtained when considering acute maternal intake by ingestion, during early lactation; that is, for maximum transfer to milk. Landolt-Börnstein New Series VIII/4
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Also shown in Table 7.9 are ratios of dose to the offspring and adult from activity transferred to the embryo and foetus during pregnancy. These ratios are based on the dose coefficients for the offspring given in ICRP Publication 88 [01I1]. These offspring doses from in utero exposure include contributions from activity retained in the tissues of the child at birth, ranging from about 90 % of the total “foetus” dose for 239Pu to less than 10 % for 137Cs. Only in the cases of 131I and 210Po do the doses to the offspring from transfer in milk exceed that resulting from in utero transfer. Table 7.9 Comparison of doses following chronic maternal ingestion of radionuclides throughout pregnancy and lactation Ratio of offspring : adult dosea Radionuclide Foetusb Infant in milkc 45 Ca 12 2.7 90 Sr 1.5 0.8 131 I 1.0 2.4 137 Cs 0.4 0.4 210 Po 0.1 0.2 239 Pu 0.04 <10−3 241 Am 0.01 <10−3 a Committed effective dose (environmental exposures). b Includes doses received in utero and from activity retained by the child at birth (based on 01I1) c Includes doses from activity transferred to milk as a result of maternal intakes during pregnancy and lactation (preliminary calculations).
7.3 Dosimetric models 7.3.1 Introduction The dose to organs of the body (this set of organs is referred to as target organs or target regions) depends in part on the distribution of the activity within the body (this set of anatomical regions is referred to as source regions) and the transport of the radiations emitted in nuclear transformations (decays) of the radionuclide residing in the source regions. In general, the “target regions” as well as “source regions” will be identified as organs of the body, but this need not be the case when knowledge suggests otherwise. For example, the short-lived decay products of radon, when inhaled, deposit on the surfaces of the respiratory airways from which they irradiate the basal cells of the bronchial epithelium as a target region of interest. Various procedures have been employed in computing the dose to target regions, given the information on the distribution of activity within the body. In the late 1960s - early 1970s, however, efforts were devoted to establishing a unified formulation that would be applicable to all types of radiations. The Medical Internal Radiation Dose (MIRD) Committee of the Society of Nuclear Medicine led efforts to formulate a computational scheme based on physical framework in support of the dosimetry of diagnostic radiopharmaceuticals. The formalism set forth by the MIRD Committee [76L1] deals with the various radiations emitted in nuclear transformation of radionuclides in a consistent manner. This formalism only addressed the absorbed dose quantity and was limited to short-lived radionuclides that emit electron (beta and conversion electrons) and gamma radiations. However, the rigorous physical basis of the system enabled it to be extended to the needs of radiological protection and the calculation of equivalent dose by the ICRP in its Publication 30 issued in 1979.
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In Chapter 4 the various radiological quantities are defined. The absorbed dose quantity D is a point dosimetric quantity that can be defined in any material. The absorbed dose within the target region T of the body is generally computed as the mean absorbed dose, denoted by DT , and thus is the average energy per unit mass absorbed by the target region. The mean absorbed dose can, of course, be written as the integral of point quantity D over the volume of the target. In radiological protection it is necessary to address all the emitted radiations and to recognize that some radiations, and thus their contribution to dose, are more biologically efficient than other radiations. Thus the basic dosimetric quantity of radiation protection is the equivalent dose in target organ T and is denoted by HT. The equivalent dose is defined as HT =
∑w D R
T ,R
(4)
R
where the summation extends over all radiations R contributing to the mean absorbed dose DT ,R in the target T. The radiation weighting factors wR represent judgments regarding the potential relative biological damage of radiation R without regard to the specific tissue or health consequence. The values in current use in dosimetry are given in Chapter 4, Section 4.5.2.2. The weighting factors reflect, in part, the density of the ionization within the target which is indicated by the linear energy transfer (LET) of the radiation. See Chapter 4 for further discussion. The tissues of the body, of course, vary in their sensitivity to ionising radiation particularly with respect to stochastic effects (cancer induction and hereditary disease). The effective dose quantity was introduced into the radiological protection system in ICRP Publication 26 and was further amplified and extended in Publication 60. The effective dose reflects the underlying information regarding the risk of stochastic effects among the irradiated tissues and aggregates these contributions into a single dosimetric quantity (see Chapter 4). The effective dose E is defined as E=
∑w H T
T
(5)
T
where the summation extends over the organs/tissues assigned tissue weighting factors wT as described in Chapter 4, Section 4.5.2.4. Table 4.3 gives tissue weighting factors recommended in Publication 26 [77I1] and in Publication 60 [91I1]. The effective dose resulting from radionuclides within the body can be added to that from external radiation fields to obtain a single quantity that encompasses both exposures as described in Section 7.7.2.
7.3.2 Absorbed fraction and specific absorbed fraction The formulation of a computational scheme that enabled the estimation of absorbed dose from all radiations emitted in nuclear transformations of radionuclides was largely achieved by the introduction of the absorbed fraction quantity. Consider a source region r within which radiation of type i is being emitted. If target region v absorbs energy from the radiation emitted in the source region r, then the absorbed fraction in v from r φi (v ← r) is defined as the quotient of the energy imparted to the target region v and the energy, exclusive of rest energy, emitted in the source region r. That is, the absorbed fraction can be expressed as
φi (v ← r) =
energy absorbed by target region v energy emitted in source region r
(6)
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The absorbed fraction is defined only for target regions which are volumes; however, no such constraint is placed on the source regions, i.e., it could be a point, line, surface, or volume. The absorbed fraction quantity embodies not only the geometric variables of size, shape, and spatial relationship of the regions, but also the extent to which the radiation is transported through the medium containing these regions. The specific absorbed fraction Φ i (v ← r) is defined as the absorbed fraction per unit mass of the target region; i.e,
Φ i (v ← r) =
φi (v ← r) mv
(7)
where mv is the mass of the target volume. The specific absorbed fraction has the property that it can be defined for all target regions, including the important case of a point target. Recall from Section 7.3.1 that absorbed dose is a point dosimetric quantity. There can, however, be no points in common between the source and target region unless one of the regions is a volume. If the source and target regions are in a homogeneous absorbing medium that is sufficiently large (relative to the range of the radiation) that edge effects are negligible, and if the activity is uniformly distributed in the source region, then the uniform isotropic model is said to apply. Under this model, the distribution of absorbed energy about the source region is a function only of distance from the source. The fraction of emitted energy absorbed per unit mass at a distance x can then be represented by the point-isotropic specific absorbed fraction Φ i ( x) . Since the emitted energy must be absorbed somewhere, the point-isotropic specific absorbed fraction must satisfy the constraint that ∞
∫
4πρ x 2Φ i ( x) dx = 1
(8)
0
where ρ is the density of the homogeneous medium. The point-isotropic specific absorbed fraction for the various radiations of interest provides the basic means of estimating specific absorbed fractions. Non-point source and target regions can be developed simply as the superposition of the point function. Thus the specific absorbed fraction between any nonpoint target region r and a point source P is the mean of the values of Φ i (x ) in the target region
Φ i (r ↔ P) = Φ i ( x)
(9)
Furthermore, the specific absorbed fraction in any region r1 from a source in another region r2 is the mean of the values of the point-isotropic specific absorbed fraction for all pairs of points in the regions; i.e.,
Φ i (r1 ↔ r2 ) = Φ i ( x)
(10)
where x is the distance between points randomly selected in r1 and r2. The doubled-headed arrow in equations 9 and 10 indicates that either region may be the source or target region. Equations 9 and 10 can be expressed in their integral representation, but for regions whose geometry is complex, i.e., other than spherical, recourse is often made to numerical evaluation using Monte Carlo techniques to randomly select the points. As noted above, the point-isotropic specific absorbed fraction is a function only of distance. Thus if the source and target regions are interchanged in Equations 6 and 7, the numerical value of the specific absorbed fraction does not change. This property of the uniform isotropic model is referred to as the Reciprocity Theorem. The conclusion of the theorem is that the specific absorbed fraction Φi is independent of which region is designated as the source and which the target is. In symbols Landolt-Börnstein New Series VIII/4
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Φ i (r1 ← r2 ) = Φ i (r2 ← r1 ) ≡ Φ i (r1 ↔ r2 )
[Ref. p. 7-68 (11)
and therefore
φi (v 2 ← v1 ) m2
=
φi (v1 ← v 2 ) m1
(12)
Note that these relationships apply to all radiations in the uniform isotropic model.
7.3.3 Computational models of the human anatomy 7.3.3.1 Mathematical phantoms The application and extension of the above in radiological protection necessitated the formulation of a computational model of the human body – a so called mathematical phantom. Such a mathematical description of the adult human was used to provide estimates of the absorbed fractions for photon emitters distributed in the model [69S1]. Much of the stimulus for this development came from the needs of the MIRD Committee and thus the mathematical model is often referred to as the MIRD phantom although over the years many modifications have been made to the model including the extension to children (Fig. 7.10). The MIRD phantom consists of three principle Sections: (1) an elliptical cylinder representing the arms, torso, and hips,; (2) two truncated elliptical cones representing the legs and feet; and (3) an elliptical cylinder representing the neck region and lower portion of the head, which is topped by a half ellipsoid. The organs of the body were represented by simple conic sections in some cases cut by planes and rotated. The defining equations were readily evaluated and thus the model was well suited for use with Monte Carlo techniques in the computation of photon absorbed fractions. Limitations in the computational hardware of the early 1970s, not the available anatomical information, restricted the realism of the modelling. 7.3.3.2 Voxel models Voxel models are human models based on computed tomographic or magnetic resonance images obtained from high resolution continuous scans of a single individual. The greyscale data of the medical images are interpreted into tissues (i.e. organs), a process known as segmentation. Each volume element, called voxel, has an identification number that identifies the discrete organ of that particular voxel. The models, consisting of millions of voxels, provide a three-dimensional representation of the human body and the spatial form of its constituent organs and structures. They were initially developed for radiological protection purposes to estimate the risk to a person or population due to an irradiation. For this purpose, a detailed model of the human body is required, together with computer codes simulating the radiation transport and energy deposition in the human body. They are at present the most precise representation of the human anatomy to be used for computational radiation transport simulation.
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Fig. 7.10. Schematic diagram of the MIRD phantom; [78S1]. Various modifications have been made by other investigators; [87C1].
Among other laboratories, the GSF- National Research Centre in Germany started the development of voxel models covering various ages and anatomical statures [02P1]: an 8 week old baby girl (Baby) and a 7 year old girl (Child) [88Z1], four male models (Golem (38 y), Visible Human (38 y), Frank (48 y) and Otoko (40 y)) and three adult female voxel models named Donna (40 y), Helga (26 y) and Irene (32 y) [02P1]. Fig. 7.11 shows three dimensional reconstructions of some organs of the above female models and demonstrates their anatomic realism. The male adult phantom Golem [01Z1] has height and weight similar to the ICRP Reference Man [75I1]. Otoko (01S1) stems from whole body CT data of a patient whose external dimensions are in good agreement with the Japanese Reference Man [89T1]. The Visible Human was constructed from CT data and photographic data from the Visible Human Project of the American National Library of Medicine, obtained from the CT pictures of a donated body of an executed 38-year-old man from Texas, USA. Both Visible Human and Frank are rather broadly built, the former being also very tall, and are probably more suitable to simulate bigger individuals. Similarly, Donna and Helga are taller and heavier than the reference adult female, as characterised in ICRP Reference Man [75I1] and ICRP Report 89 [02I3], whereas Irene has a weight below the reference female. Due to their anatomical realism, such models have been the subject of increasing interest and acceptance, and others have been developed elsewhere: Zubal et al [94Z1] at Yale developed a head and torso phantom as well as a head phantom with fine resolution from the CT data of an adult male with dimensions similar to the MIRD-5 mathematical phantom [78S1] who was imaged from head to midthigh. Dimbylow and his colleagues [96D1] at NRPB, UK, have developed the Norman model based on whole-body MRI scan data of a healthy volunteer. The exact dimensions of the voxels were scaled so that height and mass of the segmented model agreed with the values of the Reference Man [75I1]. A body representation that gained recent popularity is the Visible Human Male [98S1]; this data set was assembled using the original colour photographic anatomical slice images of the body mentioned above Landolt-Börnstein New Series VIII/4
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which was acquired by the Visible Human Project of the American National Library of Medicine. Xu and co-workers [00X1] at the Rensselear Polytechnic Institute have developed the VIP-Man based on these data, which is the most complete body description so far with respect to number of structures defined and voxel resolution. More recently, Kramer et al [03K1] created MAX based on the Yale model. Regarding paediatric models, Caon et al [99C1] constructed a trunk model of a 14-year-old female and Nipper et al [02N1] developed a newborn model which is based upon a high resolution CT scan of a 6-day female baby.
Fig. 7.11. View of the skin and some organs of the female models Helga, Donna and Irene (left) and of the skeleton and some organs of the male model Golem (right); [02Pet]. On the extreme right is an image constructed from the colour photographic anatomical slice images of the Visible Human Project of the American National Library of Medicine; [00X1].
Several image processing functions are used for segmentation of voxel models. However, none of these methods works automatically; but interactive methods controlled by an operator must be applied and thus the segmentation of a whole body model with several organs is a time-consuming procedure. Moreover, it is difficult to obtain a suitable fine, contiguous data set. The most important advantage of the voxel phantoms over mathematical phantoms, is their realism concerning anatomy: the organ shape as well as the organ location is realistic, since computed tomographic images from an actual subject were employed for their construction. Thus, they offer a clear improvement compared to the older mathematical models whose organs are described by relatively simple geometrical bodies. Furthermore, the distance between the organs, an important parameter, particularly for internal dosimetry where several organs are the so-called source organs and all the others are the targets, are realistically simulated by the voxel models, which is not always the case for the MIRD-type models. One of the most interesting characteristics of the voxel models is the possibility of varying their size and hence simulating smaller or bigger individuals. Their external dimensions can be adapted for each of the three dimensions independently. All internal dimensions of the resulting scaled-down or scaled-up version of the original model are consequently modified with the same scaling factors. However, the scaling factors should range within reasonable magnitudes, to avoid anatomical errors in the organ proportions. A limitation of the voxel models is that very small tissues, as for example the eye lenses and the skin, cannot be represented with their correct thickness because it is not possible to segment structures below the voxel resolution. An additional concern is that the supine position of the body during acquisition of the image data results in an upward shift of the abdominal organs and the compression of the lungs. However, for radionuclides with residence times in the body of a few days much of the dose is received under various postures including sitting or lying down. The position of the stomach is well known to be quite varied depending on its content. Strictly speaking voxel models represent themselves and not a “reference” or an “average” individual. For many situations in radiation protection, this is fully adequate, particularly if their external dimensions comply more or less to the reference or the exposure conditions cannot be accurately described. Moreover, since the models available range from very slim persons to large and heavy persons, they can be used to estimate the doses to an individual by selecting the model fitting best to the person under consideration. However, for certain situations involving regulations, guidelines or dose limitations, for example exposures of a population or radiation workers, dose values are required for a “reference” Landolt-Börnstein New Series VIII/4
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individual. As a consequence, there is a need to construct body models that combine a realistic anatomy with organ masses, shapes and locations that are representative. Probably the best way to achieve such an aim is to modify an appropriate voxel model (i.e., one that already resembles Reference Man data in its external body dimensions) to one having reference organ masses as well, retaining its realistic anatomy. This approach was used by Dimbylow [96D1] to construct NORMAN, the model of an adult male, and is now underway at the GSF-National Research Centre in Germany under the supervision of the ICRP. These reference voxel models are then appropriate for calculating organ doses for reference persons and, hence, establishing reference dose conversion coefficients for international recommendations such as those from ICRP.
7.3.4 Dose rate per unit activity, S-factor Consider an organ T of the body which at time t is absorbing energy from activity in a source region S of the body. Let the activity, i.e., the average rate of nuclear transformations (nt), in the source region be AS(t) and denote the mean energy emitted as radiation of type R per nuclear transformation by ∆R; that is ∆R = YR ER, where YR and ER are the yield (number per nt) and energy of radiation R, respectively. The rate at which energy is absorbed in T per unit mass at time t, which is by definition the mean absorbed dose rate, D& (T ← S,t ) due to the activity in S is R
D& R (T ← S,t ) = AS (t ) ∆R Φ R (T ← S)
(13)
In general, more than one type of radiation will be associated with the nuclear transformation process of a particular radionuclide, and thus the mean absorbed dose rate is D& (T ← S,t ) =
∑ D&
R
(T ← S, t ) = AS (t )
R
∑∆
R
Φ R (T ← S)
(14)
R
If the activity is present in a number of source organs, then an additional summation must be considered to derive the mean absorbed dose rate in T D& (T ,t ) =
∑ A (t )∑ ∆ S
R
S
Φ R (T ← S)
(15)
R
where the first summation is over all source regions S. Examination of the above equations reveals that the factors within the inner summation, i.e., ∆R and ΦR(T←S), reflect the physical data on the nuclear transformation process and the transport of the emitted radiations between S and T which depend in part on nature of the radiations and the anatomical relationship of these two regions. Given an agreed-upon analog of the human body for estimation of ΦR, then considerable effort can be saved through consideration of the additional quantity S. If we define S (T ← S) as S (T ← S) =
∑∆ Φ R
R
R
(T ← S) =
∑Y
R
ER Φ R (T ← S)
(16)
R
where YR is the yield of radiation R of energy E R . Then the expression for the mean absorbed dose rate reduces to
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7 Internal dosimetry of radionuclides
D& (T ,t ) =
∑ A (t ) S (T ← S)
[Ref. p. 7-68 (17)
S
S
The quantity S represents the mean absorbed dose rate in T per unit radioactivity in S. If S is considered to be invariant with time, that is, if the analog of the human body and its implied geometric relationships are independent of age, then integration of Equation 17 yields the mean absorbed dose in T. D (T ) =
~
∑A
S
S (T ← S)
(18)
S
~ where AS denotes the time integral of the activity in the source region (the cumulated activity). Thus S may also be defined as the mean absorbed dose per unit cumulated activity. Methods for deriving the cumulated activity are discussed in Section 7.4.1. The S factor can be expressed in terms of equivalent dose by inclusion of the radiation weighting factors in its defining equation; i.e, SEE (T ← S ) =
∑∆
R
wR Φ R (T ← S) =
R
∑Y
R
ER wRΦ R (T ← S)
(19)
R
where wR is the radiation weighting factor (Section 7.3.1). The quantity SEE is a radiological protection quantity introduced by ICRP (1979) as the specific effective energy. However a more appropriate name would be the specific equivalent energy. The equivalent dose rate to tissue T can be written in terms of SEE as follows, •
H (T ,t ) =
∑ A (t ) SEE (T ← S) S
(20)
S
and the effective dose rate is defined as the sum of the weighted equivalent dose rates in a number of tissues (Section 4.5.2.3 and Equation 5, this Chapter). •
E (t ) =
∑w
T
•
H (T ,t )
(21)
T
7.3.5 Specific absorbed fractions for various radiations The three principal modes of nuclear transformation are beta decay, alpha decay, and isomeric transition. An additional process, spontaneous fission, is available to some heavy nuclides. The principle radiations involved in these modes of nuclear transformation are alpha particles, electrons (either negative or positive charge) and photons (electromagnetic radiation). The latter two radiations may arise from the nucleus as well as the orbital electrons of the newly formed radionuclide. These radiations differ significantly in their energy deposition pattern, as a result of different mechanisms through which they interact with matter, further details are given in Section 3.5.
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7.3.5.1 Electrons A continuous energy spectrum of electrons is associated with beta decay. The spectrum ranges in energy from zero to the maximum energy permitted by the difference in the energy level of the parent and daughter nucleus. Electrons of discrete energy are also observed in nuclear transformation, as a result of processes involving the orbital electrons. Under the auspices of the MIRD Committee, Berger has tabulated point-isotropic specific absorbed fraction data for monoenergetic electron sources ranging in energy from 0.025 to 5 MeV [71B1]. To facilitate numerical use, Berger presented the data in terms of a scaled point kernel F(r/r0) where r0 is the continuous slowing down approximation (csda) range. The point-isotropic specific absorbed fraction Φ(r) in terms of Berger's scaled point kernel is
Φ (r ) =
1 F (r / r0 ) 4πρr 2 r0
(22)
where ρ is the density of the medium. The tabulations were prepared for water as a surrogate medium for soft tissue. Table 7.10 presents the 90-percentile distance (x90) in water as a function of electron energy. The 90-percentile distance is defined to be the radius of a sphere around a point source within which 90 % of the emitted energy is absorbed. As can be seen from Table 7.10, electrons of energy up to about 2 MeV deposit their energy within a distance of less than 1 cm. Table 7.10. Deposition of electron energy (range-energy relationships). Percentile distance x90 in water for electrons from monoenergetic sources. The results for E0 ≤0.020 MeV are extrapolated (based on Table 9, page 15, 71B1). E0 [MeV] x90 [cm] E0 [MeV] x90 [cm] 4.0 1.57 0.70 0.207 3.5 1.36 0.60 0.169 3.0 1.16 0.50 0.131 2.6 0.99 0.40 0.096 2.2 0.82 0.30 0.063 2.0 0.74 0.20 0.0334 1.5 0.53 0.10 0.0106 1.2 0.41 0.05 0.00318 1.0 0.328 0.010 0.000194 0.90 0.287 0.005 0.000060 0.80 0.247 0.001 0.000008 Organs of the body are of dimensions sufficiently large relative to the electron range that the electron absorbed fraction may be taken as unity if the source is uniformly distributed in the organ. Thus the specific absorbed fraction for electrons is 1 / mT , if T = S
Φ (T ← S) =
0 ,
if T ≠ S
(23)
A notable exception to the above occurs for walled organs where the source resides in the contents, e.g., urinary bladder and the segments of the gastrointestinal tract. For organs whose contents contain an electron emitter, the specific absorbed fraction in the wall of the organ from its contents is usually taken to be given by
Φ (wall ← contents) = Landolt-Börnstein New Series VIII/4
1 2 mcontents
(24)
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[Ref. p. 7-68
where mcontents is the mass of the contents. This relationship is derived from the fact that the dose rate at the surface of a half space containing a uniform distribution of activity is one-half the equilibrium dose rate at locations within the contaminated half space far from the interface. It should be noted that the approach for walled organs may be very conservative, in that the critical cells are typically considered to be the basal cells of the epithelial layer, which lie at some depth in the tissue; in the gastrointestinal tract, they are further shielded by a mucus layer. Thus the dose rate in the wall may decrease rapidly from the value at the surface, particularly for low-energy electrons. Consideration of these details in the dosimetric models must await further description of the location of the cells at risk. 7.3.5.2 Alpha particles The point-isotropic specific absorbed fraction Φ(x) has not been tabulated in the literature for alpha particles since the range of alpha particles in tissue is sufficiently small that for organs of the body, an absorbed fraction of unity can be assumed. However, in some specific instances – such as alpha emitting short-lived radon daughters deposited on the airways of the lung – consideration must be given to the energy deposition pattern. The point-isotropic specific absorbed fraction Φ(x) can be defined as
Φ ( x) =
1 (dE / dx) x 4π ρ x 2 Eα
(25)
where (d E /d x) x is the stopping power of the alpha particle at the energy it has after travelling a distance x, and E α is the initial energy of the alpha emission. In order to avoid the discontinuity at x = 0, the quantity 4πx 2 ρΦ ( x) is tabulated for the point-isotropic specific absorbed fraction. The mass stopping 1 (dE / dx ) and the csda range as a function of energy in soft tissue are shown in Figure 7.12. power
ρ
5DQJH JFP
G(G[ ρ 0H9FP J
These data can be used with Equation 25 to compute the point-isotropic specific absorbed fraction. Specific absorbed fractions for source-target pairs in the body are the same as employed for beta radiation, that is Equation 23 is applicable to solid organs. For walls of the gastrointestinal tract Equation 24 is applied, however only 1 % of the alpha particles' energy is considered to penetrate the mucous lining of the tract.
Fig 7.12. Mass stopping power and range of alpha particles in soft tissue.
( 0H9
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7.3.5.3 Gamma-rays and characteristic X-rays Gamma-rays and X-rays are electromagnetic radiations of short wavelength, orders of magnitude shorter than visible light. A nucleus in an excited state from which it is energetically impossible to de-excite through emission of particulate radiation (emission of alpha or beta particles) may de-excite through the emission of one or more photons of electromagnetic radiation. Many nuclides formed in beta or alpha decay may be in an excited state, and thus gamma-ray emission often accompanies these decays. Electromagnetic radiations associated with changes in nuclear state are referred to as gamma radiation. A measure of the probability per unit distance (density distance) travelled by a photon that an interaction occurs is the mass attenuation coefficient. As the three interaction events, photoelectric effect, Compton effect and pair formation (see Chapter 3) are independent and mutually exclusive, the total mass attenuation coefficient µ /ρ is given as
µ / ρ =τ / ρ +σ / ρ +κ / ρ
(26)
where τ /ρ, σ /ρ and κ /ρ are the mass attenuation coefficient for the photoelectric effect, Compton effect, and pair formation interactions. Tabulations of these coefficients for various elements and compounds of general interest have been given [95H1]. Values for other compounds or absorbing media can be computed as
µ/ρ =
∑ w (µ / ρ ) i
i
(27)
where (µ /ρ)i is the tabulated value for the ith element, and wi is the fraction by weight of the ith element in the medium of interest. The equation is valid because the chemical binding energies between atoms in a molecule are very small, thus not significantly altering the electronic binding energies. The transfer of energy from the photon to secondary electrons is given by the mass energy-transfer coefficient and is denoted by µen /ρ. The mass energy-transfer coefficient is the weighted sum of the mass attenuation coefficients; i.e.,
µ en / ρ = fτ ( τ / ρ ) + fσ ( σ / ρ ) + fκ ( κ / ρ )
(28)
The weights fτ, fσ, and fκ indicate, for their respective interactions, the fraction of the photon energy which is converted into kinetic energy of secondary electrons and dissipated in the medium by collision losses. It is beyond the scope of this Chapter to detail the prescription for estimation of the weights. It is important to note that the weights reflect only the energy transferred as kinetic energy of charged particles and thus energy emitted as X-rays following photoelectric effect and the rest mass energy of the positronelectron pair in the pair formation process are excluded from the weight. It should be further noted that for the composition of body tissues and typical photon energies, the correction for bremsstrahlung energy loss by the secondary electrons is rather small. 7.3.5.4 Point-isotropic specific absorbed fraction The fraction of the energy emitted by a point- isotropic source that is absorbed per unit mass at a distance x from the source the point isotropic specific absorbed fraction Φ(x) can be expressed as
Φ ( x) =
µ en e − µ x B (µ r ) ρ 4π x 2 en
Landolt-Börnstein New Series VIII/4
(29)
7-36
7 Internal dosimetry of radionuclides
[Ref. p. 7-68
µρ FP J
µHQρ FP J
where x is the distance from the point source, µ is the linear attenuation coefficient at the source energy, µen / ρ is the mass energy-transfer coefficient at the source energy, and Ben(µr) is the energy-absorption buildup factor. The mass attenuation and mass energy-transfer coefficients for soft tissue as a function of photon energy are shown in Fig. 7.13. The build up factor is defined as the ratio of the absorbed dose obtained from a measurement to the absorbed dose calculated to be due to “uncollided” photons at the location. The scattered photons are of lower energy than the uncollided photons and hence subject to increased absorption as seen in Fig. 7.13. Several tabulations of energy-absorption buildup data are available in the literature for application to body tissues. Berger presented such data in MIRD Pamphlet No. 2 [68B1] for a point source in water. Published values applicable to 40 mean free paths (µr = 40) have been published by Spencer and Simmons [73S1], whereas Berger's data were applicable to only 20 mean free paths. For small values of µr, Ben(µr) is approximately unity and increases rapidly with increasing values of µr. The maximum value of the buildup factor occurs for photons of about 100 keV.
Fig 7.13. Mass attenuation and mass energytransfer coefficients for photons in soft tissue; [95H1].
( 0H9
7.3.6 Calculation of doses to soft tissues and the skeleton For the majority of tissues in which radionuclides deposit, following their entry into the blood, the assumption is made that both they and the sensitive cells are essentially uniformly distributed. On this basis the average tissue dose is calculated. This applies for example to the liver, spleen, kidneys, muscle, gonads and glands. In some cases, particles or colloids containing radionuclides may enter the blood and these will result in a heterogeneous distribution of activity in tissues in which they deposit, particularly for alpha emitters with a short path length (~ 50 µm). Most experimental evidence suggests, however, that a heterogeneous distribution of activity is no more likely to produce long term damage, such as cancer, than a homogeneous distribution of activity and the calculation of average tissue dose is justified [80I3, 99S1, 03M1]. In the case of the skeleton, however, it has been necessary to take account of the way radionuclides deposit in order to assess the dose to sensitive tissues. A generalised model to assess doses from bone seeking radionuclides was given in ICRP Publication 30 [79I1] (Table 7.11). The target tissues in the skeleton are taken to be the active red bone marrow (RBM), which is present in cavities in trabecular bone, and endosteal and epithelial cells assumed to lie within 10 µm of bone surfaces (BS). Energy deposited in the yellow marrow of cortical bone is not considered to cause any radiation effects. The radiation dose to the RBM and BS depends upon the pattern of deposition of the radionuclides in the bone (volume or surface deposition), the radiation it emits (α, β, γ) and the effective half-time. Some bone surface seeking elements are Ga, Zr, Th, Pu, Am and Cm and some volume seekers Ca, Sr, Ra and U. The most recent biokinetic models for Pu, Am and Cm allow for their progressive burial in the skeleton (Section 7.2.5.3). Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-37
The total endosteal area of the skeleton over which the dose is calculated is taken as 12 m2, half associated with cortical bone and half with trabecular bone. The layer on bone surfaces over which the equivalent dose is averaged has a mass of 120 g. The mass of the red bone marrow in cavities within trabecular bone is taken to be 1500 g. Table 7.11 gives the fraction of energy absorbed for α- and βemitters deposited in the skeleton. Thus, for an α-emitter such as 239Pu, which initially deposits on the bone surface, half the activity deposited in the skeleton is taken to be associated with trabecular bone surfaces and half to cortical bone surfaces. For activity on trabecular bone surfaces 25 % of the energy released will be absorbed by the sensitive 10 µm cell layer on bone surfaces (BS), 50 % will be absorbed in the red bone marrow (RBM); the remaining energy will be harmlessly dissipated in bone mineral. For activity on cortical bone surfaces 25 % of the energy released will be absorbed by cells on bone surfaces; the rest will be dissipated in yellow-marrow or in mineral bone. For radionuclides deposited in bone volume the fraction of energy deposited in sensitive cells, and hence the dose, will be less than for bone surface seekers as much of the energy will be dissipated in bone mineral. It should be noted that some recent analyses of radiation-induced bone tumours and information on bone cell development suggests that sensitive cells for bone tumour induction may reside at distances of more than 10 µm from bone surfaces and that a depth of 50 µm may be a more appropriate depth over which to calculate doses [00G1]. Preliminary calculations suggest that this may reduce the dose to cells near trabecular bone surfaces by a factor of about 2. Table 7.11. Fraction of energy deposited in target organs for α- and surfaces or in bone volume Source organ Target α-emitter α-emitter β-emitter on organ uniform in on bone bone surfaces volume surfaces Eβ ≥0.2 MeV Trabecular bone BSa 0.025 0.25 0.025 Cortical bone BS 0.01 0.25 0.015 Trabecular bone RBMb 0.05 0.5 0.35 Cortical bone RBM 0.0 0.0 0.0 a BS = Bone surfaces; b RBM = Bone marrow
β-emitters deposited on bone β-emitter on bone surfaces Eβ<0.2 MeV 0.25 0.25 0.5 0.0
β-emitter uniform in bone 0.025 0.015 0.35 0.0
7.4 Dose coefficients The previous Section described some of the fundamental principles used in the calculation of radiation dose. In practice many radiation protection professionals will not need to implement these methods in full, but will use published values of dose coefficients as a means of estimating doses to individuals or populations. A dose coefficient is defined as a committed dose per unit intake (Sv Bq−1), the term can be applied to either committed equivalent doses or committed effective doses. Dose coefficients are sometimes referred to as “dose per unit intake values”. ICRP has adopted lowercase letters to denote dose coefficients. Thus, a committed equivalent dose to a tissue T is denoted HT and the corresponding dose coefficient is denoted hT, similarly the dose coefficients for effective doses (E) are denoted e. Dose coefficients provide a means of converting intakes (Bq) into committed doses (Sv) from internal emitters and are therefore used in many branches of radiological protection such as environmental assessments, the protection of workers and nuclear medicine. As referred to earlier in this Chapter, ICRP has, over the last decade or so, published a number of documents giving dose coefficients for workers and members of the public, including children (Table 7.1). These Publications provide a consistent source of reference values. ICRP has also issued a publication giving doses to the embryo and foetus from intakes by the mother. In addition ICRP has given dose coefficients (usually expressed in mGy MBq−1) for use in nuclear medicine. More details are given in Section 7.4.2.4.
Landolt-Börnstein New Series VIII/4
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7 Internal dosimetry of radionuclides
[Ref. p. 7-68
7.4.1 Method of calculation The biokinetic models recommended by ICRP are described in Section 7.2. These models describe the fate of an inhaled or ingested radionuclide as it enters the body, its subsequent absorption and distribution among systemic tissues and its elimination from the body. Therefore, in most cases these models are combined to form a complete model for predicting the behaviour of the radionuclide from initial entry into the body until excretion or decay. For example, the complete model for inhaled plutonium would require that the Human Respiratory Tract Model (HRTM) be combined with the systemic model for plutonium (for activity taken up to blood) and the model for the gastrointestinal tract (for activity mechanically cleared from the lung and swallowed and for activity excreted into the GI tract in bile). These model constructs are referred to as compartment models, each compartment representing an apparent influence on the kinetic behaviour of the radionuclide. All the models recommended by ICRP to date describe the transfer of radionuclides (activity) between compartments of the model by linear first order processes. That is, the rate of biological removal of activity from a compartment at time t is taken to be the product of the activity in the compartment at time t and a transfer rate coefficient, usually denoted by k. Consider an isolated compartment i as in Fig. 7.14. Compartment i receives activity from and transfers activity to all other compartments in the model. The element kj,i of the transfer rate coefficient matrix describes the fraction of the activity in compartment i transferred to compartment j per unit time. The model is completely described by the transfer rate coefficients matrix and the initial activity (content at time zero) in each compartment. The nuclear transformations of many radionuclides form radioactive nuclei which must be considered in computing the dose coefficient. Thus the intake of a radionuclide, the parent, may result in a series of radionuclides being formed within the body. The kinetics of radioactive series must be superimposed on the kinetics described by the biokinetic models. The activity Am i of member m of the decay chain in compartment i is given by a set of coupled linear differential equations of the form of Equation 30. d m Ai (t ) = dt
inflow n
=
∑k
j =1, j ≠i
m i, j
+
A (t ) + λm m j
ingrowth m−1
∑F
m′ , m
m′ i
−
outflow
A (t ) − A (t ) (λm + m i
m′=1
(30)
n
∑k
m j ,i
)
j =1, j ≠i
where m = 1 for the parent radionuclide, k mj ,i is the fraction of the activity of chain member m in compartment i transferred to compartment j per unit time, λm is the decay constant of chain member m, Fm′, m is the fraction of the decays of chain member m′ which form member m (often referred to as the branching fraction) and by definition Aim′ (t ) = 0 if m = 1 . In computing dose coefficients the set of differential equations are generally solved as an initial value problem such as specified in Eq. 31. 0 for all i and m > 1 A (0 ) = 0 for m = 1 and i not a compartment of intake nonzero for m = 1 and i as a compartment of intake m i
(31)
For example, an intake by ingestion would assume that at time zero one unit of activity of the parent radionuclide is present in the stomach content and zero activity of the parent and daughter products is present in all other compartments of the model. Two approaches have been applied to describe the biokinetics of the daughter products. The so-called shared kinetics assumes that the behaviour of the daughter is the same as the parent. That is, kim, j = ki1, j , m = 2,..., n . If information indicates that the daughter product behaves differently from the Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-39
parent radionuclide in the body then independent kinetics are applied in Equation 30. The well recognized independent kinetic cases include the formation of noble gas radionuclides in the decay of solid daughter 83 parents, e.g., 35 Br→8336m Kr and radioiodine formed in the decay of radioisotopes of tellurium, e.g, Te→133 53 I . However other cases of importance include the long decay series associated with some of the
133 52
natural decay chains (such as 22890Th ) where fundamentally different behaviours of the chain members in the skeleton are encountered. k i, j
Isolated compartment i Ai (t)
k j,i
Fig 7.14. Isolated compartment exchanging activity with other compartments in the model.
The biological removal coefficient of the activity of chain member m from ith compartment is given n
by
∑k
m j ,i
which can be stated as a half-time by dividing ln(2) by the removal coefficient. This would be
j =1, j ≠i
the half-time for removal if there were no input into the compartment from other compartments. If the compartment is subject to continued input then a plot of the compartment activity as a function of time will reflect the removal rates from all compartments feeding the ith compartment. Thus in general one has to be careful in speaking about biological half-times for complex models. With the exception of a catenary model (a model in which the compartments only communicate with adjacent compartments; that is, the nonzero members of transfer coefficient matrix are k1,2, k2,3, k3,4, ..., kn−1,n) no closed form solution exists for Equation 30. The catenary form of Equation 30 is well-known in radiation protection as it describes the serial kinetics of a decay chain and its solution was formulated in earlier work by Bateman in 1910 [10B1] who was assisting Lord Rutherford in the early investigations of radioactivity. Catenary models were used in ICRP Publication 2 [59I1] and their solution has been extensively investigated by Skrable [74S1]. Since only the catenary system or very small systems (less than four compartments) can be solved exactly, it is necessary to approximate the solutions numerically. Applicable numerical approaches include analysis of eigenvalues and discrete variable methods (also called step-by-step methods or difference methods). With the advent of powerful desktop computing the discrete variable Runge-Kutta method can be readily applied to Equation 30 (see for example 88B1). However well-written variableorder, variable-step routines such as the solver developed by Gear [71G1] are more efficient than a fixedorder, variable-step Runge-Kutta routine. The calculations of Publication 30 [80W1] were carried out using Hindmarsh’s coding of the Gear method [74H1]. A collection of state-of-the-art solvers for the initial value problem for ordinary differential equation systems, including the Gear method, are contained in the widely available ODEPACK package [83H1]. Classical methods from linear algebra involve the Landolt-Börnstein New Series VIII/4
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7 Internal dosimetry of radionuclides
[Ref. p. 7-68
eigenvalues and eigenvectors of the matrix of transfer rate coefficients as described in good texts on linear algebra (for example 85J1). The methods are referred to as the matrix exponential method and eigenanalysis. Birchall and James [89B1] described the matrix exponential method and listed the source code for effective implementation of this method on a desktop computer. The eigenanalysis method has been implemented by, among others, Killough and Eckerman [84K1], Bertelli [87B1] and Polig [01P1]. A transportable library of state-of-the art numerical routines for eigenvalue problems is available in the LAPACK Package [00A1]. An advantage of the eigenanalysis approach is that the solution can be expressed analytically even though it is a numerical approximation. An imaginative hybrid numericalanalytical method has been developed by Leggett and co-workers that is extremely simple, flexible and highly efficient [93L1]. Some of these methods have been reviewed by Peace [03P1]. With the memory and speed available in today’s desktop computers and the availability of well written solvers, such as those within ODEPACK, the numerical aspects of compartment modelling are no longer a significant issue. However, having available a number of different solvers is useful in ensuring an appropriate approximation to the solution. The method of calculation of dose is described in Section 7.3 and is encapsulated in Equation 15. The number of nuclear transformations of the kth chain member in the ith compartment during the period t1 to t2 U ik (t1 ,t 2 ) is given as t2
∫
U (t1 , t 2 ) = Aik (u ) du k i
(32)
t1
where Aik (t ) is obtained as the solution to Equation 30. For protection of workers the integral of Equation 32 is evaluated from time zero (the time of the intake) to 50 years post intake and the integral is denoted ~ as U(50). The distinction between A typically used in the dosimetry of radiopharmaceuticals and U is that former has no restriction on the upper limit of the integration since it is typically short-lived radionuclides. The committed equivalent dose coefficient for tissue T, hT, for the worker is thus computed as hT =
∑∑U k
k S
(50) SEE (T ← S)
k
(33)
S
where the outer summation extends over the parent radionuclide and all members of its decay chain and the inner summation extends over all source regions in which activity may reside. Note that the compartments in the model must be assigned to the anatomical source regions S in computing US. For children both the parameters of the biokinetic models and S-factors (or SEE values) change with time. They must therefore be varied in an effectively continuous manner from the time of intake to age 70 years. Absorbed fraction data for photons and charged particles are available for six specified ages enabling sets of S-factors to be generated for newborn, 1, 5, 10, 15-years-olds and adults. At intermediate ages S-factors can be derived using an interpolation scheme. ICRP has used a linear interpolation in the inverse total body mass domain. Biokinetic models are specified by ICRP for six standard ages, 3-month-old, 1, 5, 10, 15-year-old, and adult. These models specify the transfer coefficients (ki,j) which determine the rates at which material is transferred between the different parts of the body (Equation 30). Transfer coefficients at intermediate ages can be derived using linear interpolation. Numerical methods, which are characterised by the variable time steps inherent in the method, can easily accommodate time-varying transfer coefficients and S-factors. For methods which apply an analytical solution the continuous variation of transfer rates must be modelled discretely using a time-stepping method. Starting at the age of intake the calculation is advanced by a time-step small enough for the interpolated rates to be held constant without undue loss of accuracy. The computed activities at the end of one time-step are used as the initial conditions for the next step. In this way the number of transformations within each step or interval is computed. S-factors calculated at the beginning or mid-point of a step are taken to apply to the Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-41
whole interval. The committed dose is then the sum of the doses computed in each time step. In essence the numerical methods are integrating dose rate (Equation 15) while analytical methods apply Equation 18 to a series of larger time steps. One of the advantages of the analytical methods is in calculations for adults where rate constants and S-factors are constant; the time step can then be the whole of the period of interest, e.g. 50 years. Numerical methods are advantageous, however, where the intake is complicated, perhaps varying in a difficult manner with time.
7.4.2 Sources of dose coefficients The recommendations of ICRP and other international bodies advance on different fronts at different rates. This means that it is not always straightforward to identify the most appropriate dose coefficients for a particular application (e.g. workplace, environment, nuclear medicine). For example, for inhalation by members of the public, dose coefficients are given for three lung absorption Types (F, M and S) for 31 important elements (Publication 72, 96I1). A default Type is specified for use when the chemical form is not known. For workers a comprehensive review of lung absorption characteristics for various chemical forms of radionuclides has not been undertaken since Publication 30 although updated dose coefficients using the HRTM have been issued [94I1]. The following subsections aim to help the reader identify the most appropriate dose coefficients at the time of publication of this review. The results given in ICRP dose compendia take into account the ingrowth of decay products in all regions of the body following an intake of unit activity of the parent nuclide. They do not take into account any activity of decay products in the initial intake. Thus doses from any radioactive decay products present at the time of the intake (perhaps in equilibrium with the parent) may need to be added to the dose from the parent nuclide. 7.4.2.1 Workers A compendium of dose coefficients (committed effective dose to 50 years after the intake) for workers for over 800 nuclides based on the HRTM model (Section 7.2.1), the ICRP Publication 30 model for the GI tract (Section 7.2.2) and the most recent ICRP systemic models (Section 7.2.5) has been issued in ICRP Publication 68 [94I1] which implemented the tissue weighting factors in Publication 60 [91I1] (Table 7.1). These results are also given in the Euratom Directive [96E1] and the IAEA Basic Safety Standards [96I1]. The previous complete source of doses to workers was ICRP Publication 30, published in four parts between 1979 and 1988 [79I1, 80I1, 80I2, 88I1]. Publication 30 concentrated on giving results as Annual Limits on Intake (Bq) rather than as dose coefficients. For workers, inhalation dose coefficients are based on an AMAD of 5 µm and specified distributions of time spent at two levels of exercise − sitting and light exercise [94I2]. In support of Publication 68, ICRP has issued Publication 78 [97I2] which contains bioassay predictions, such as daily urinary excretion and retention in lung, skeleton and whole-body, based on the same models used in Publication 68 for a limited range of radionuclides. This enables health physicists to use a consistent set of models in dose assessments. This document fills the role for ICRP Publication 68 that Publication 54 filled for Publication 30. Phipps et al [98P1] provide extended results and fitted functions for predicting bioassay quantities at times not addressed in the ICRP report. For particle sizes other than 5 µm AMAD, Ishigure et. al. calculated bioassay quantities for 0.1, 0.3, 1, 3 and 10 µm and have uploaded the results onto the National Institute of Radiological Sciences web site [02I4]. Further details of the methods of monitoring and dose assessment in the workplace are given in Sections 7.4-7.7 and in Chapter 10, Sections 10.3.2. and 10.3.3. A CD-ROM of dose coefficients for both members of the public and workers has been issued by ICRP [99I1]. It is consistent with the Publication 68 and it extends the results given in Publications 68 and 72 by giving inhalation dose coefficients for ten particle sizes (0.001, 0.003, 0.01, 0.03, 0.1, 0.3, 1, 3, 5, 10 µm AMAD) and ingestion coefficients. Effective doses and equivalent doses for all important tissues for a Landolt-Börnstein New Series VIII/4
7-42
7 Internal dosimetry of radionuclides
[Ref. p. 7-68
range of integration periods (1, 7, 30 days, 1, 5, 10, 20, 30 and 45 years) are given, together with the dose coefficients to age 70 years. The package also contains extensive help files and the text of Publications 68 and 72. 7.4.2.2 Members of the public Following the Chernobyl accident it was acknowledged that a set of internationally-agreed dose coefficients was required. Thus, ICRP provided age-specific biokinetic models for selected radionuclides in Publications 56, 67, 69, 71 and 72 [89I1, 93I1, 95I1, 95I2, 96I1] together with dose coefficients (committed effective dose to age 70 years) for six age groups: 3-month-old infants, 1-, 5-, 10-, 15-year old children and adults (Table 7.1). More details about the models themselves are given in Section 7.2. ICRP Publication 56 [89I1] considered H, C, Sr, Zr, Nb, Ru, I, Cs, Ce, Pu, Am and Np. Dose coefficients for selected radionuclides were based on the tissue weighting factors of ICRP Publication 26 [77I1] and the lung model of ICRP Publication 30. These results have now been superseded by other Publications. ICRP Publication 67 [93I1] considered S, Co, Ni, Zn, Mo, Tc, Ag, Te, Ba, Pb, Po, and Ra, and in addition revised some of the models given earlier in Publication 56. In particular the model for Sr was revised substantially and made consistent with the generic model structure used for Ba and Ra. By the time of the issue of ICRP Publication 67 the tissue weighting factors given in Publication 60 were available thus the dose coefficients are consistent with the most recent ICRP recommendations. Both equivalent (hT) and effective (e) dose coefficients were given and the results of ICRP Publication 56 were updated whether or not the systemic model was updated in Publication 67. Only ingestion was considered as a route of intake. Publication 69 [95I1] gave a similar range of dose coefficients for radionuclides of Fe, Se, Sb, Th and U. Following the publication of the model for the Human Respiratory Tract [94I2] ICRP Publication 71 reviewed the lung absorption characteristics of environmental forms of the 31 elements considered in Publications 56, 67, 69. ICRP Publication 71 also introduced biokinetic models for Ca and Cm. Inhalation dose coefficients based on the new model were given for all three default Types (F, M and S) and one of these three was recommended as a default for situations where the chemical form is unknown. Details of the HRTM model are given in Sction 7.2.1. For members of the public, dose coefficients are based on an Activity Median Aerodynamic Diameter, AMAD, of 1 µm and specified distributions of time spent at four levels of exercise (sleep, sitting, light exercise and heavy exercise [96I2]). Dose coefficients for ingestion were not given in Publication 71. A large compendium of inhalation and ingestion dose coefficients (effective dose only) for members of the public for over 800 nuclides based on the HRTM model (Section 7.2.1), the ICRP Publication 30 model for the gastrointestinal tract (Section 7.2.2) and the most recent systemic models (Section 7.2.5) was published in ICRP Publication 72 [96I1]. These results are also given in the Euratom Directive [96E1] and the IAEA Basic Safety Standards [96I2]. The CD-ROM of dose coefficients (Section 7.4.2.1, 99I1) also contains doses for members of the public consistent with ICRP Publication 72 [99I1]. As noted above, for 31 elements dose coefficients are given for three default absorption Types (F, M and S), while for the remaining elements it is assumed that compounds assigned to the Publication 30 classes (D, W and Y) are categorised as F, M and S respectively; thus for guidance about individual chemical forms of these elements one must refer to ICRP Publication 30. 7.4.2.3 Embryo and foetus ICRP has issued dose coefficients for the embryo and foetus in Publication 88 [01I1]. More details on the biokinetic models used for the dose calculations are given in Section 7.2.7, this subsection therefore covers only the phantoms used in foetal dosimetry and some aspects of the dosimetry. The pattern of energy deposition within the foetus is modelled in Publication 88 [01I1] using results from two separate sets of computer phantoms developed at Oak Ridge National Laboratories, USA (ORNL). The first is a series for the pregnant female developed by Stabin et al [95S1]; the second is for the foetus itself, developed by Eckerman [03E1]. Energy-dependent specific absorbed fractions (SAFs) Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-43
are given for both electrons (including beta particles and positrons) and photons. Due to the small organ masses and short ranges between organs, for beta radiation the beta spectrum is used and not the mean energy. In calculating organ and tissue doses for infants, children and adults, electrons have, in most cases, been assumed to be non-penetrating, i.e. their energy is taken to be absorbed entirely in the organ or tissue in which they are emitted (exceptions are the skeleton and walled organs). In the case of the foetus, however, the extremely small size of some tissues can mean that a substantial fraction of electron energy can be deposited outside the tissue where the electron is emitted [98U1]. Thus, activity in the foetal thyroid could deliver an electron dose to nearby tissues such as the brain. However, these so-called cross-fire doses are generally much lower than the doses to the source tissues themselves. 7.4.2.4 Nuclear medicine patients and volunteers in clinical research ICRP Publication 53 [87I1] presented biokinetic models and best estimates of biokinetic data for some 120 individual radiopharmaceuticals. It included absorbed dose coefficients for organs and effective dose equivalent coefficients (Publication 30 [79I1] terminology) calculated up to infinity, due to the administration of short-lived radionuclides for diagnostic or experimental purposes. The calculations used ICRP’s Publication 26 tissue weighting factors [77I1]. Some information on the range of variation to be expected in pathological states, for adults, children and the foetus were given. The absorbed dose coefficients are used in clinical diagnostic work for judging the risk associated with the use of specific radiopharmaceuticals, both for comparison with the possible benefit of the investigation and to help in giving adequate information to the patient. These estimates also provide guidance to ethics committees having to decide upon research projects involving the use of radioactive substances in volunteers who receive no individual benefit from the study. ICRP Publication 53 supplemented ICRP Publication 52, Protection of the Patient in Nuclear Medicine [88I3]. In Publication 80 [98I1], ICRP provided biokinetic models, absorbed doses to organs and effective dose coefficients, using the tissue weighting factors of ICRP Publication 60, for 10 new radiopharmaceuticals. It also provided recalculated dose coefficients for the 19 most frequently used radiopharmaceuticals from ICRP Publication 53, using ICRP Publication 60 dosimetry. An integrated index to all radiopharmaceuticals treated in ICRP Publications on nuclear medicine up to 1998 gave a listing of effective dose coefficients for adults. Recently Stabin and Siegel [03S1] have used the best current radiation decay data and computer phantoms to calculate dose coefficients for use in nuclear medicine. Decay data for over 800 radionuclides from the data service at Brookhaven National Laboratory were combined with absorbed fraction data from a number of currently available mathematical whole body and organ models to calculate the dose coefficients. Many more (816) radionuclides are considered than in the ICRP compendia and some alpha emitters are included. New models are also employed, and dose coefficients for bone and marrow have been updated with recently suggested modifications.
7.4.3 Dose coefficients for selected radionuclides 7.4.3.1 Doses to tissues following intakes of radionuclides Equivalent doses to tissues following inhalation of 239PuO2 by an occupationally exposed worker are given in Table 7.12. These doses are calculated using the HRTM [94I1], assuming inhalation Type S and a 5 µm AMAD aerosol, and the biokinetic model for plutonium given in Publication 67 [91I1]. The highest committed (50 year) doses are to the lung, as the site of entry into the body, and the skeleton (cells near bone surfaces) and liver, as the main sites of deposition from the blood. ICRP also calculates the committed effective dose which provides a method for comparing doses, and hence risks, from intakes of radionuclides with those from external radiation. This is discussed in detail in Chapter 4. Landolt-Börnstein New Series VIII/4
7-44
7 Internal dosimetry of radionuclides
[Ref. p. 7-68
Table 7.12. Committed equivalent doses to tissues, weighted committed equivalent doses and committed effective doses from inhalation by a worker of 239PuO2 (5 µm AMAD)a Tissue weighting Weighted committed Organ or tissue Committed equivalent factor equivalent dose [Sv Bq−1] dose [Sv Bq−1] −5 Cells near bone surfaces 9.1 × 10 0.01 9.1 × 10−7 −7 Colon 0.12 1.7 × 10 2.0 × 10−8 −5 Liver 0.05 1.9 × 10 9.5 × 10−7 −5 Lungs 0.12 4.7 × 10 5.6 × 10−6 −6 Red bone marrow 0.12 4.5 × 10 5.4 × 10−7 −7 0.05 Remainder 1.0 × 10−8 2.0 × 10 Skin 0.01 1.5 × 10−7 1.5 × 10−9 −7 Stomach 0.12 1.5 × 10 1.8 × 10−8 −6 Gonads 0.20 1.2 × 10 2.4 × 10−7 Committed effective 8.3 × 10−6 dose a Inhalation Type S Examples of equivalent dose coefficients to tissues from inhalation intakes of 131I, 137Cs and 239Pu by workers are given in Table 7.13. For an intake of 131I the main dose is to the thyroid, with a dose that is rather more than 1000 times that to other tissues. For 137Cs, which moves rapidly from the lungs to blood (Type F) and distributes throughout the body, most tissues receive a very similar dose. In the case of 239Pu oxide, which has a longer retention time in the lung (Type S) and deposits principally in the liver and skeleton there is a greater range of doses although the long retention time in the skeleton results in the highest tissue dose to cells near bone surfaces, as described above. Table 7.13. Doses to tissues following inhalation by a worker of 131I, 137Cs and 239Pu Committed equivalent dose [Sv Bq−1] Iodine-131 Tissue (Type F) Thyroid 2.1×10−7 Red bone marrow 5.5×10−11 Cells near bone surfaces 6.9×10−11 Colon 5.1×10−11 Lungs 8.1×10−11 Liver 2.4×10−11 Committed effective dose 1.1×10−8 a For inhalation of 5 µm AMAD aerosol b Assumes wR = 20 for α particles
Caesium-137 (Type F)
Plutonium-239 (Type S)b
6.3×10−9 6.3×10−9 6.6×10−9 8.1×10−9 6.1×10−9 6.5×10−9 6.7×10−9
1.5×10−7 4.5×10−6 9.1×10−5 1.7×10−7 4.7×10−5 1.9×10−5 8.3×10−6
Some examples of dose coefficients for different radionuclides following intakes by inhalation and ingestion are shown in Table 7.14. The variation in doses reflects differences in behaviour after intakes by inhalation or ingestion, variations in distribution and retention in tissues as well as the use of a radiation weighting factor wR of 20 in the calculation of equivalent dose to tissues from deposited α emitters. The lowest doses are from intakes of tritiated water and the highest doses from inhalation of the α emitters 224 Ra, 226Ra and 241Am.
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Table 7.14. Comparison of dose coefficients following inhalation or ingestion of various radionuclides by a worker. f1c Radionuclide Lung Dose coefficient [Sv Bq−1] a Type Inhalation Ingestion 3 b −11 H2O – 1.0 1.8×10 1.8×10−11 60 −9 Co M 0.1 3.4×10−9 7.1×10 90 −8 Sr F 0.3 3.0×10 2.8×10−8 95 −9 Zr F 0.002 3.0×10 8.8×10−10 95 −9 Nb M 0.01 1.3×10 5.8×10−10 106 −9 Ru F 0.05 9.8×10 7.0×10−9 131 −8 I F 1.0 1.1×10 2.2×10−8 134 −9 Cs F 1.0 9.6×10 1.9×10−8 137 −9 Cs F 1.0 6.7×10 1.3×10−8 144 −4 −8 Ce M 5×10 2.3×10 5.2×10−9 210 Po F 0.1 1.1×10−6 6.8×10−7 224 −6 Ra M 0.2 2.4×10 6.5×10−8 226 −5 Ra M 0.2 1.2×10 2.8×10−7 232 −4 −5 Th M 5×10 2.9×10 2.2×10−7 234 −6 U M 0.02 2.1×10 4.9×10−8 235 −6 U M 0.02 1.8×10 4.6×10−8 238 −6 U M 0.02 1.6×10 4.4×10−8 239 −5 −6 Pu S 1×10 8.3×10 9.0×10−9 241 −4 −5 Am M 5×10 2.7×10 2.0×10−7 242 −4 −6 Cm M 5×10 3.7×10 1.2×10−8 a Inhaled materials are classified as Type F, M or S (Fast, Moderate or Slow) which refer to their rates of absorption to blood from the respiratory tract (Section 7.2.1.2). AMAD = 5 µm. b Tritiated water is assumed to be completely absorbed from the lungs c Fractional absorption from the gut 7.4.3.2 Application of dose coefficients in risk calculations The doses to tissues calculated using the dosimetric models developed by ICRP together with risk coefficients for fatal cancer recommended by ICRP for the different tissues [91I1] can be used for assessing the consequences of intakes of radionuclides. An example is given in Table 7.15 for 239Pu inhaled as the oxide (inhalation Type F, AMAD = 5 µm) by a worker. Table 7.15. Estimation of risk of cancer death following inhalation of 239PuO2 (AMAD 5 µm) Risk for Tissue Cancer Risk coefficient Sv Bq−1 100 kBq inhaled [cancer deaths Sv−1]a inhaled −5 −3 Lungs Lung 1 in 31 4.7×10 6.8×10 −5 −3 Liver Liver 1 in 430 1.9×10 1.2×10 −5 −4 Cells near bone surfaces Bone tumour 1 in 280 9.1×10 4.0×10 Red bone marrow Leukaemia 1 in 550 4.5×10−6 4.0×10−3 Colon Colon 1 in 8,300 1.7×10−7 6.8×10−3 −6 −2 Committed effective dose 1 in 26b 8.3×10 5.0×10 a Risk coefficients for workers [91I1] b Based on risks calculated to individual tissues Landolt-Börnstein New Series VIII/4
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The ICRP dosimetric model for plutonium can be used to estimate the committed equivalent dose to tissues following an intake by inhalation. By weighting these doses by the appropriate risk coefficients [91I1], the consequences of an intake of 239PuO2 can be estimated (Table 7.15). The calculations have been undertaken using the HRTM [94I2] and assuming an AMAD of 5 µm (the default value for workers). The biokinetic model for plutonium given in ICRP Publication 67 (93I1, Figure 7.7) has also been used. Following the inhalation of 100 kBq of 239PuO2, for example, the risk of developing lung cancer would be (4.7×10−5) × (6.8×10−3) × 105 = 3.2×10−2 i.e. a 1 in 31 risk. The risks of leukaemia, liver and bone cancer would all be lower, about 1 in 550, 1 in 430 and 1 in 280 respectively. The overall risk of developing cancer would be about 1 in 26. These risks are clearly a maximum as they are based upon committed doses and depend upon the full risk to the tissues being expressed. They would therefore apply only to intakes received early in life. A similar, but approximate calculation could be carried out using the committed effective dose (8.3×10−6 Sv Bq−1) multiplied by the risk coefficient for whole body radiation exposure recommended by ICRP (5×10−2 Sv−1 ) given in Publication 60 [91I1]. For 100 kBq inhaled this would give an overall risk of 4.15×10−2 (i.e an overall risk of 1 in 24). This is very similar to the value given in Table 7.15, the difference being mainly due to the fact the ICRP uses rounded values of tissue weighting factors wT to calculate effective doses. If a more specific calculation was needed for an individual then it would be necessary to consider the accumulation of dose by the individual over time (i.e. year on year) and how that risk would be expressed. Using the same approach the overall risk of developing cancer following either the inhalation or ingestion of 100 kBq of some of the radionuclides for which dose coefficients are given in Table 7.14 are given in Table 7.16. There is a significant difference in the risk from intakes of α-emitters compared with β/γ-emitters, reflecting the much higher wR value of 20 for α particles. In general the risks of inhalation intakes are higher than for ingestion because of the lower absorption and faster rate of clearance from the gut than from the respiratory system. This is not the case for radionuclides such as tritium (as HTO) or caesium that are readily absorbed from the gut. For the same activity, the risk associated with ingestion of 239 PuO2 is nearly three orders of magnitude less than that following inhalation because of the low absorption in the gut (f1 = 10−5). For ingestion of 100 kBq of 131I the risk is about 1 in 38,000, the majority of the risk of radiation-induced cancers predicted being in the thyroid (more than 99 %). Following inhalation of the same amount of activity the risk is somewhat lower (1 in 78,000) as only about half of the activity inhaled is deposited in the respiratory system. Similar considerations apply to the inhalation and ingestion of 137Cs, although in this case, as the radionuclide distributes throughout body tissues the risk of cancer will be distributed amongst a number of tissues, with the greatest risks being for leukaemia and lung cancer. Of the radionuclides considered, 226Ra and 232Th are the most toxic with risks of about 1 in 7.4 and 1 in 10 respectively of developing fatal cancer following inhalation of 100 kBq. At the other extreme 3H is the least toxic with a risk of less than 1 in a million.
7.5 Internal monitoring This Section describes the general principles for individual monitoring. Sections 10.3.2 and 10.3.3 in Chapter 10 give more detailed information on in vivo measurements by whole and partial body counting as well as by analyses of excreta. This Section is reproduced from ICRP Publication 78, 97I2, paras. 58 76.
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Table 7.16. Comparison of risk of radiation-induced cancer death associated with inhalation or ingestion of 100 kBq of some radionuclides. Risk of cancer death Radionuclide Lunga Type f1c Inhalation Ingestion 3 b 7 H2O 1.0 1 in 1.4×10 1 in 1.4×107 90 Sr F 0.3 1 in 9,300 1 in 9,700 95 Zr F 0.002 1 in 86,000 1 in 170,000 95 Nb M 0.01 1 in 150,000 1 in 310,000 106 Ru F 0.05 1 in 22,000 1 in 18,000 131 I F 1.0 1 in 78,000 1 in 38,000 137 Cs F 1.0 1 in 37,000 1 in 19,000 144 Ce M 1 in 9,600 1 in 22,000 3×10−4 210 Po F 0.1 1 in 310 1 in 870 224 Ra M 0.2 1 in 64 1 in 3,000 226 Ra M 0.2 1 in 7.4 1 in 1,200 232 Th M 1 in 10 1 in 1,300 5×10−4 239 Pu S 1 in 26 1 in 18,000 1×10−5 241 Am M 1 in 12 1 in 1,100 5×10−4 242 Cm M 1 in 46 1 in 15,000 5×10−4 a Inhaled materials are classified as Type F, M or S (Fast, Moderate or Slow) which refer to their rates of absorption to blood from the respiratory tract (Section 7.2.1.2). AMAD = 5 µm b Tritiated water is assumed to be completely absorbed from the lungs c Fractional absorption from the gut
7.5.1 Methods of individual monitoring The purpose of this Section is to describe briefly the main measurement techniques, their advantages and their limitations. In most cases, individual monitoring for intakes of radionuclides may be achieved by body activity measurements, excreta monitoring, air sampling with personal air samplers, or a combination of these techniques. The choice of measurement technique will be determined by several factors: the radiation emitted by the radionuclide; the biokinetic behaviour of the contaminant; its retention in the body taking account of both biological clearance and radioactive decay; the required frequency of measurements; and the sensitivity, availability, and convenience of the appropriate measurement facilities. Routine monitoring programmes usually involve only one type of measurement if adequate sensitivity can be achieved. For some radionuclides, only one measurement technique is feasible, e.g. urine monitoring for intakes of tritium. For radionuclides, such as plutonium isotopes, that present difficulties for both measurement and interpretation, a combination of techniques has to be employed. If different methods of adequate sensitivity are available, the general order of preference in terms of accuracy of interpretation is: body activity measurements; excreta analysis; personal air sampling. Results of monitoring of the working environment (area monitoring) may provide information that assists in interpreting the results of individual monitoring, e.g. information on particle size, chemical form and solubility, time of intake. The results of workplace monitoring for air contamination may sometimes be used to estimate individual intakes. However the interpretation of the results of measurements from air sampling in terms of intake is not simple and may be misleading. The most common form of representative sampling is by using fixed samples at a number of selected locations intended to be reasonably representative of the breathing zone of the worker. When such a method is routinely used for quantitative determinations of intake, the representativeness of the results should be determined using a special monitoring programme, often involving personal air samples. Landolt-Börnstein New Series VIII/4
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Monitoring in relation to a particular task or event may often involve a combination of techniques so as to make the best possible evaluation of an unusual situation, for example, a programme of both body activity and excreta measurements and, in some circumstances, personal air sampling. In some cases of suspected incidents, screening techniques (such as measuring nose blow samples or nasal smears) may be employed to give a preliminary estimate of the seriousness of the incident. In these cases the regional deposition for ET1 can be used to confirm that an intake has occurred and to give a rough estimate of the intake. 7.5.1.1 In Vivo measurements The IAEA [96I2] has given guidance on the direct measurement of body content of radionuclides. Direct measurement of body or organ content provides a quick and convenient estimate of activity in the body. It is feasible only for those radionuclides emitting radiation that can escape from the body. In principle, the technique can be used for radionuclides that emit: X- or γ-radiation; positrons, since they can be detected by measurement of annihilation radiation; energetic β-particles that can be detected by measurement of bremsstrahlung; some α-emitters that can be detected by measurement of the characteristic X-rays. Many facilities for the measurement of radionuclides in the whole body or in regions of the body consist of one or a number of high efficiency detectors housed in well-shielded, low-background environments [96I2]. The geometrical configuration of the detectors is arranged to suit the purpose of the measurement, e.g. the determination of whole-body activity or of activity in a region of the body such as the thorax or the thyroid. The skull or knees may be used as a suitable site for measurement of radionuclides in the skeleton. Care must be taken to remove surface contamination before body activity is measured. For routine measurements, determination of whole-body content is often adequate for radiological protection purposes. Total body activity will then consist of systemic activity and activity in the gastrointestinal and respiratory tracts. However, in special investigations, or in interpretation of unusual measurements, it may be advantageous to determine the distribution within the body either by profile scanning or by analysis of the relative response of detectors placed at different positions along the body. Commonly encountered fission and activation products, such as 131I, 137Cs and 60Co, can be detected with comparatively simple equipment at levels that are adequate for radiation protection purposes. Such simple equipment may consist of a single detector, viewing the whole body or a portion of the body, or, for iodine isotopes, a small detector placed close to the thyroid. The advantage of simple equipment is that it may be operated at the place of work, thereby avoiding the time required to visit a remote wholebody monitoring facility. Measurements may then be made more frequently so that any unusually large intake would be recognised soon after it had occurred. In contrast, high sensitivity techniques are needed for monitoring a few radionuclides at the levels that are required for protection purposes. Examples are the α-emitting radionuclides such as plutonium isotopes. Until recently, most body activity measurement facilities, whether high-sensitivity or simple systems, used thallium-activated sodium iodide detectors. These have the advantage that crystals of large volume can be manufactured and so provide high efficiency for detection of γ-rays. Interpretation of a γ-ray energy spectrum obtained from a mixture of radionuclides may, however, raise some difficulties. The components of the spectrum can be resolved by a multiple linear regression analysis technique, but this requires previous calibration of the detection equipment with standard sources of the required radionuclides dispersed in a matrix in such a way as to simulate the distribution and attenuation within the body. The increasing availability of high-efficiency germanium detectors is leading to their use in situations where workers may be exposed to mixtures of γ-ray emitting radionuclides. The superior energy resolving power of these detectors simplifies the interpretation of spectra obtained from complex mixtures of radionuclides. The activity present in a wound can be easily detected with conventional β-γ detectors if the contaminant emits energetic γ-rays. In the case of contamination with α-emitting radionuclides, detection is more difficult since the low energy X-rays that follow the α-decay will be severely attenuated in tissue; Landolt-Börnstein New Series VIII/4
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this effect is more important the deeper the wound. It is often necessary to localise the active material and this requires a well-collimated detector. Wound monitors most have an energy discrimination capability if a good estimate is to be made of contamination with mixtures of radionuclides. 7.5.1.2 Analysis of excreta and other biological materials In some cases, excreta monitoring may be the only measurement technique for those radionuclides which have no γ-ray emission or which have only low energy photon emissions. Excreta monitoring programmes usually involve analysis of urine, although faecal analysis may be required in some circumstances, for example where an element is preferentially excreted via faeces or to assess clearance of Type-S material from the respiratory tract. Other samples may be analysed for specific investigations. Examples are nose blow or nasal smears as routine screening techniques or blood, in the case of suspected high level contamination. The collection of urine samples involves three considerations. Firstly, care must be taken to avoid adventitious contamination of the sample. Secondly, it is usually necessary to assess the total activity excreted in urine per unit time from the sample provided. For most routine analyses, a 24 h collection is preferred but, if this is not feasible, it must be recognised that smaller samples may not be representative. Tritium is a particular case for which it is usual to take only a small sample and to relate the measured activity concentration to the concentration in body water. Thirdly, the volume required for analysis depends upon the sensitivity of the analytical technique. For some radionuclides, adequate sensitivity can be achieved only by analysis of several days’ excreta. The analysis of faecal samples for routine monitoring involves uncertainty in interpretation owing to daily fluctuations in faecal excretion. Ideally, therefore, collection should be over a period of several days. However, this may be difficult to achieve in practice and interpretation may need to be based on a single sample. Faecal monitoring is more often used in special investigation, particularly following a known or suspected intake by inhalation of Type M or insoluble S compounds. In these circumstances measurement of the quantity excreted daily may be useful in the evaluation of clearance from the lungs and in the estimation of intake. Early results may be useful in identifying exposed individuals. Radionuclides that emit γ-rays may be determined in biological samples by direct measurement with scintillation or semiconductor detectors. Analysis of α- and β-emitting radionuclides requires chemical separation followed by appropriate measurement techniques. Measurement of so-called total α or β activity may occasionally be useful as a simple screening technique, but there is no method that will determine accurately all the α and β activity in the sample. The technique may be used in routine monitoring situations where intakes are expected to be very low compared with annual limits. The results would not be interpreted quantitatively, but would be used to provide confirmation of satisfactory conditions, an unusual result indicating the need for further investigation which would include radiochemical analysis. Total activity measurements may also be useful following a known contamination event or to identify those samples that merit early attention. Measurements of total α or β activity cannot be used in quantitative evaluations of intake or committed effective dose, unless the radionuclide composition is known. Measurement of activity in exhaled breath is a useful monitoring technique for some radionuclides such as 226Ra and 228Th since the decay chains of both these radionuclides include gases which may be exhaled. It can also be used to monitor 14CO2 formed in vivo from the metabolism of 14C-labelled compounds. 7.5.1.3 Air sampling A Personal Air Sampler (PAS) is a portable device specifically designed for the estimation of intake by an individual worker from a measurement of time-integrated concentration of activity in air in the breathing zone of the worker. A sampling head containing a filter is worn on the upper torso close to the breathing zone. Air is drawn through the filter by a calibrated air pump carried by the worker. Ideally, sampling Landolt-Börnstein New Series VIII/4
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rates would be representative of typical breathing rates for a worker (~1.2 m3 h−1). However, sampling rates of current devices are only about 1/10 of this value. The activity on the filter may be measured at the end of the sampling period to give an indication of any abnormally high exposures. The filters can then be retained, bulked over a longer period, and the activity determined by radiochemical separation and high sensitivity measurement techniques. An estimate of intake during the sampling period can be made by multiplying the measured integrated air concentration by the volume breathed by the worker during the period of intake. There are three important requirements for a PAS device. Firstly, the sampler should collect sufficient material for the activity corresponding to a significant intake to be measurable in a reasonable counting time. This will depend mainly on the lowest committed effective dose that the PAS is required to detect. Typically, in a routine monitoring programme, the requirement will be to detect annual intakes that in total give rise to committed effective doses greater than 1/10 of the annual dose limit. Secondly, the volume of air aspirated by the sampler should be sufficient to provide a statistically accurate representation of the activity concentration in the breathing zone of the worker. PAS monitoring is most often used for radionuclides such as plutonium, for which a very small number of particles may contain activities that would correspond to a significant intake. The statistics of sampling small numbers of events then becomes the critical factor in determining sampling accuracy. Thirdly, the particle collection characteristics of the sampler should be known. These depend on the aspiration efficiency of the sampling head and the collection efficiency of the filter. The aspiration efficiency is the ratio of the particle concentration in the air entering the sampler to that in the ambient air. It is usually close to unity for particles of aerodynamic diameter less than about 1 µm, but the inertia of larger particles will give a tendency to under- or over-sample according to conditions. Similar effects apply to particles entering the nose and mouth and are taken into account in the ICRP Human Respiratory Tract Model [94I2] (the aspiration efficiency of the respiratory tract is termed inhalability). A PAS does not provide information on particle size. Nevertheless, it is important either to determine the particle size distribution of the inspirable material or to make realistic assumptions about it, since it can have a marked effect on deposition fractions in the respiratory tract, and hence on dose estimates. This is particularly important now that the recommended default AMAD of 5 µm is intended to be realistic rather than conservative in terms of dose estimation [95D1, 97A1]. All samplers are size selective to a greater or lesser extent, under- or over-sampling at particular particle sizes, and this can result in errors in intake estimation. The aspiration efficiency of a PAS should therefore be determined to indicate whether corrections are necessary. An investigation of the aspiration efficiency of a PAS gave values close to unity up to an aerodynamic diameter of 30 µm under workplace conditions [86M1]. It has been suggested that samplers should be designed to collect the inspirable fraction rather than the total aerosol [81V1]. Use of such samplers would be acceptable, but would require modification of analysis procedures, since the ICRP Respiratory Tract Model implicitly assumes that the total aerosol concentration is known. Static air samplers (SAS) are commonly used to monitor workplace conditions, but can underestimate concentrations in air in the breathing zone of a worker, typically by a factor of up to about 10 [80M1]. Nevertheless, if SAS devices are sited appropriately, a comparison of PAS and SAS measurements can be used to define a PAS:SAS air concentration ratio which can be used in the interpretation of SAS measurement for dose assessment purposes. It should however be recognised that the use of SAS is a relatively indirect method for assessing doses, and use of the results to estimate individual dose requires a careful assessment of exposure conditions and working practices. Apart from their potential use for dose estimation, SAS devices can also provide useful information on radionuclide composition, and on particle size if used with a size analyser such as a cascade impactor.
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7.6 Monitoring Programme (Section 7.6 is reproduced from ICRP Publication 78, 97I2, paras. 81 - 88)
7.6.1 Need for a monitoring programme ICRP Publication 75 [97I1] recommends that the emphasis in any particular monitoring programme should be on the formal assessment of doses to those workers who are considered likely to receive routinely a significant fraction of the relevant dose limit or who work in areas where exposures could be significant in the event of an accident. The results of workplace monitoring should give an indication of the likelihood of doses from intakes exceeding 5 mSv a year. Experience has shown that workers involved in the following operations would normally require individual monitoring: • handling large quantities of gaseous and volatile materials, e.g. tritium and its compounds in large scale production processes, in heavy water reactors, and in luminising; • uranium mining and processing and fabrication of uranium and mixed oxide fuels; • processing of plutonium and other transuranic elements; • processing and use of thorium, and • production of large quantities of radionuclides and radiopharmaceuticals. Selection of the type of monitoring programme depends upon the frequency of contamination of the workplace. In situations where contamination events are very infrequent, it is unlikely that routine individual monitoring would be required. Workplace monitoring should be undertaken and the results used to trigger a programme of individual monitoring in relation to special events. However, for the processes listed above, if contamination of the workplace occurs frequently, a routine individual monitoring programme would be appropriate. For workers who are not routinely employed in areas that are designated as controlled areas in relation to the control of airborne contamination and who are unlikely to have significant intakes of radionuclides, routine monitoring of the workplace will usually be sufficient to provide assurance that intakes are adequately controlled.
7.6.2 Routine monitoring The required frequency of measurements in a routine monitoring programme depends upon the retention and excretion of the radionuclide, the sensitivity of the measurement techniques available, and the acceptable uncertainty in the estimate of intake and committed effective dose. The measurement technique should be selected so that uncertainties in the measured value are small in relation to the major source of uncertainty which usually lies in the unknown times of intakes. The frequency of measurements within a routine monitoring programme should be chosen so as to reduce the uncertainty arising from the unknown time of intake to an acceptable level.
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7.6.3 Special or task-related monitoring Special monitoring refers to monitoring carried out in actual or suspected abnormal situations. Taskrelated monitoring is carried out to provide information about a particular operation. Since both special and task-related monitoring relate to distinct events, either real or suspected, one of the problems encountered in interpretation of routine monitoring results does not apply, viz. the time of intake is known. Furthermore, there may be more information about the physical and chemical form of the contaminant.
7.6.4 Confirmatory monitoring One method of confirming that working conditions are satisfactory is to carry out occasional individual monitoring. Such measurements can be interpreted only qualitatively, but unexpected findings would give grounds for further investigation. Confirmatory monitoring of this type is most useful for those radionuclides that are retained in the body for long periods, and occasional measurements provide a check on the build-up of the activity within the body.
7.6.5 Wound monitoring When skin is broken, punctured or abraded, radioactive material can penetrate to subcutaneous tissue and thence be taken up by body fluids. Depending upon the radionuclides and the amount of activity it may be necessary to undertake a medical investigation and a programme of special monitoring. In these circumstances, the amount of radioactive material at the site of the wound should be determined taking into account self-attenuation of the radiation in the foreign material and in tissue, as an aid to decisions on excision. If an attempt is made to remove material from the wound, measurements should be made of the removed material and of any activity remaining at the wound site, so as to maintain an activity balance. Subsequently, a series of measurements should be made to determine uptake to body tissues. These measurements may consist of in vivo measurements, or urine or faecal excreta monitoring, as appropriate for the particular radionuclides. If whole-body measurements are made, it may be necessary to shield any activity remaining at the wound site. Uptake can be assessed from the data given in Section 7.8. If medical intervention to prevent uptake or enhance excretion is considered, then it should be noted that any treatment will modify the biokinetic behaviour described by the models given in Section 7.1 and the data in Section 7.7 cannot be used directly to assess committed effective doses when treatment has been administered. When therapy is used following an accidental intake, a programme of special monitoring should be undertaken to follow the distribution and retention of the particular contaminant in the person, and these data should be used to make a specific assessment of committed effective dose for that person.
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7.7 Dose Assessment (Section 7.7 is reproduced from ICRP Publication 78, 97I2, paras. 103 - 109). Examples for dose estimation from results of in vitro measurements are given in Section 10.3.3.9.
7.7.1 Estimation of intake and dose For special or task-related monitoring when the time of intake is known, the intake can be estimated from the measured results using the predicted values of measured quantities as illustrated by Figures 7.15 to 7.25. If only a single measurement is made, the intake can be determined from the measured quantity M by Intake =
M m(t )
(34)
where m(t) is the predicted value at the time of intake t. The intake can be multiplied by the dose coefficient to give the committed effective dose; this can then be compared with the dose limit or any predetermined investigation level based on dose. If the measurement indicates that an investigation level has been exceeded, further investigation is required. The nature of the investigation will depend upon the circumstances and the extent to which the investigation level is exceeded. The following should be considered: • repeated measurements to confirm or refine the initial evaluation, and • the use of additional monitoring techniques. If a series of measurements is available, the data in Figures 7.15 to 7.25 provide the time course of the predicted activity (at least over a period of 10 days). The predicted values can then be scaled to obtain the best fit to the measured data points. The best fit is usually taken to be that fit which minimizes the sum of the squares of the residuals, a residual being defined as the number of standard deviations separating a measurement from the fitted curve. The intake is then equal to the value by which the predicted values are scaled. For routine monitoring, it is assumed that intake took place in the middle of the monitoring interval of T days. For a given measured quantity M obtained at the end of the monitoring interval, the intake is Intake =
M m(T 2 )
(35)
and the dose from intake in the monitoring interval is obtained by multiplying the intake by the dose coefficient. The dose or intake can be compared with the dose limit or of the activity corresponding to that limit. Alternatively, the dose or intake can be compared with pre-determined investigation levels. An intake in a preceding monitoring interval may influence the actual measurement result obtained. If more than about 10 % of the actual measured quantity may be attributed to intakes in previous intervals, for which intake and dose have already been assessed, a correction should be made. For a series of measurements in a routine monitoring programme, the following procedure may be observed:
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• determine the magnitude of the intake in the first monitoring interval; • predict from the graphs in Section 7.8 the contribution to the subsequent measurement from this intake; • subtract this contribution from all subsequent data, and • repeat above for the next monitoring interval. Alternative techniques for assessing committed effective dose from a series of measurement values are described in the literature, e.g. [96P1]. If a measured value in a routine monitoring programme exceeds a pre-determined investigation level, further investigation is required. The nature of the investigation will depend upon the circumstances and the extent to which the investigation level is exceeded. The following should be considered: • • • •
repeated measurements to confirm or refine the initial estimate; the use of additional monitoring techniques; review of the working conditions and the circumstances of the exposure; if default parameter values were used in the original assessment, investigation of the particle size and chemical form of the actual contaminant and selection of more appropriate values, if necessary, and • in cases of substantial intakes, removal of the contaminated person from work with radioactive materials and investigation of the actual retention and excretion characteristics, in order to refine the dose assessment.
7.7.2 Control of worker doses The limit on the annual effective dose to a worker applies to the sum of the effective doses from external exposure and committed effective dose from intakes of radionuclides. For practical purposes, the total effective dose ET can be calculated from the formula: ET = H p (d ) +
∑e j
j ,inh
(50) ⋅ I j ,inh +
∑e
j ,ing
(50) ⋅ I j ,ing
(36)
j
where Hp(d) is the personal dose equivalent at a depth d in the body, normally 10 mm for penetrating radiation, ej.inh(50) is the committed effective dose per unit activity intake by inhalation from radionuclide j, integrated over 50 years, Ij.inh is the intake of radionuclide j, by inhalation, ej,ing(50) is the committed effective dose per unit activity intake by ingestion from radionuclide j, integrated over 50 years , and Ij,ing is the intake of radionuclide j by ingestion. Strictly, personal dose equivalent is an operational quantity measured in the workplace using personal dosemeters, whereas the committed effective dose quantities are calculated using measurements of other parameters (e.g. air concentrations) in the workplace. However, for practical purposes the two kinds of quantity can be combined in the assessment of the total effective dose. In the assessment of committed effective doses from internal radionuclides it is often helpful to work in terms of the secondary quantities: Annual Limit on Intake (ALI, Bq); and Derived Air Concentration (DAC, Bqm−3). The ALI is the intake which would lead to a committed effective dose of 20 mSv (the average annual limit on effective dose). ALI =
0.02 e(50)
(37)
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The DAC is the activity concentration in air which would lead to an intake of one ALI assuming a standard breathing rate (1.2 m3 h−1) and annual working hours (2,000). DAC =
ALI 1.2 × 2000
(38)
These values should not be seen as limits in the way that the 5-year-averaged effective dose is limited, but rather as helpful guides to whether the limits are likely to be approached or exceeded.
7.8 Monitoring data for radionuclides In this Section illustrative graphs of predicted values of measured quantities (whole-body retention, specific organ retention, daily urinary or faecal excretion) are given in Figs 7.15 to 7.25 as a function of time following a single intake by inhalation, ingestion and injection. The data for the following radionuclides are included: 3H, 60Co, 90Sr, 106Ru, 131I, 134Cs, 137Cs, 144Ce, 234, 235, 238U, 239, 240Pu, and 241Am. For inhalation, results are generally given for a single lung clearance Type which is representative of chemical forms present in the workplace (see Table 7.4). For tritium, a graph for inhalation of tritiated water is given which is treated as Class SR-2. In the case of 239, 240Pu and 241Am graphs for both Type M and Type S forms are given. For ingestion, f1 values recommended for materials in the workplace are applied (see Table 7.6). For direct entry into the blood the graphs are applicable to soluble (transportable) forms of radionuclides that have been directly injected into the bloodstream or have entered the body by inhalation, ingestion or through skin/wound contamination. The Human Respiratory Tract Model in ICRP Publication 66 [94I2] was used to calculate particle deposition and respiratory tract clearance of the deposited particles. The subject exposed to the aerosols was the ICRP reference worker doing light work: defined as light exercise with the ventilation rate of 1.5 m3 h−1 for 5.5 h + sitting with the rate of 0.54 m3 h−1 for 2.5 h. The following ICRP default values [94I2] for the physical characteristics of the radioactive aerosols were used. • • • •
Activity Median Aerodynamic Diameter (AMAD) = 5 µm geometric standard deviation of particle size = 2.5 particle density = 3 g cm−3 particle shape factor = 1.5
The GI tract model in ICRP Publication 30 [79I1] was used. The rate constant λB for the absorption of the materials from the small intestine to the blood was obtained from the value of f1, the fraction of materials absorbed into blood from the small intestine, using the equation:
λB = f1λSI / (1 − f1 )
(39)
where λSI is the rate constant of material transfer from the small intestine to the upper large intestine. If f1 value is 1, 0.99 was taken for calculation, which is in line with the ICRP publications. The latest ICRP biokinetic models at 2003 were used, which are given in the ICRP publications listed in Table 7.1. In the graphs of Figs 7.15 to 7.25 body or organ retention for day 1 means the content at the end of day 1 etc. For excreted activities, the value at day 1 represents the activity excreted during the first day after intake, corrected for radioactive decay to the end of day 1. One exception to this is for the intake of tritiated water; the activity concentration in urine was calculated by dividing the whole body activity at the time of sampling by the volume of body water, 42 litres. In the context of in vivo measurements, the following definitions are relevant. Whole-body retention is the sum of systemic material (including that in the urinary bladder) and material retained within the respiratory and gastrointestinal tracts. The lung retention is taken to be the sum of the contents of the thoracic lymph nodes and the bronchial, bronchiolar, and alveolar-interstitial regions. Landolt-Börnstein New Series VIII/4
7-56
7 Internal dosimetry of radionuclides
[Ref. p. 7-68
Hydrogen-3 (half-life = 12.3 y) Possible chemical forms of 3H to which workers are exposed include tritium gas (HT), tritiated water (HTO) and organically bound tritium (OBT) [89I1]. Tritium emits low energy β−-particles (0.0057 MeV in average) with 100 % yield and is readily detected by liquid scintillation counting of a urine sample. A typical detection limit readily achievable in monitoring programme is 100 Bq/l for urine samples [97I2]. Since the activity concentration of HTO in urine is assumed to be equal to that in body water, the analysis of HTO in a urine sample is used to give activity concentration in body water at the time of sample collection. 0
a
10 H-3 (Water)
H-3 (OBT), f1=1
10-1
Fraction of ingested activity [-]
Fraction of inhaled/ingested/injected activity [-]
100
10-2 10-3 10-4 10-5 10-6 10-7 10-8 100
Whole body Concentration in urine 101
102 103 Time after intake [d]
104
b
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Urine 0
10
1
2
10 10 Time after intake [d]
3
4
10
Fig. 7.15. Predicted whole-body retention, daily urinary excretion or concentration in urine of 3H as a function of time following acute intake of unit activity of 3H via a all intake routes for tritiated water, b ingestion of organically bound tritium (OBT), the f1 value of which is 1.0.
Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-57
Cobalt-60 (half-life = 5.27 y) Insoluble compounds of cobalt, e.g. oxides, hydroxides, halides and nitrates are assigned to Type S (f1=0.05 for workers, 0.01 for adult members of the public) and all other compounds to Type M (f1=0.1) by ICRP [94I1]. Cobalt-60 emits two high-energy γ-rays (1.173, 1.332 MeV) per disintegration, which are highly penetrable radiations and therefore readily detected by photon detectors positioned outside the body. A typical detection limit readily achievable in monitoring programme is 50 Bq of 60Co in the whole body and 100 Bq in the lungs [97I2]. Gamma-ray spectrometry on biological samples permits detection of 1 Bq/l of 60Co in urine and 1 Bq per sample of faeces [97I2]. 0
100
10 Co-60, Type M, 5 µm
Co-60, f 1=0.1
Fraction of ingested activity [-]
Fraction of inhaled activity [-]
10-1 10-2 10-3 10-4 10-5 10-6
Whole body Urine Faeces Lungs
10-7
a
10-8 100
101
102 103 Time after intake [d]
104
b
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Urine Faeces 0
10
1
2
10 10 Time after intake [d]
3
4
10
0
10
Fraction of injected activity [-]
Co-60
c
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Urine Faeces 0
Landolt-Börnstein New Series VIII/4
10
1
2
10 10 Time after intake [d]
3
4
10
Fig. 7.16. Predicted whole-body/lung retention, or daily urinary/faecal excretion of 60Co as a function of time following acute intake of unit activity of 60Co via a inhalation of particulate aerosols of Type M compounds with the AMAD of 5 µm, b ingestion of compounds whose f1 value is assumed to be 0.1, c injection of 60Co in soluble forms.
7-58
7 Internal dosimetry of radionuclides
[Ref. p. 7-68
Strontium-90 (half-life = 29.1 y) All compounds of strontium possibly present in the work place, except for strontium titanate (SrTiO3), are assigned to Type F (f1=0.3) by ICRP [94I1]. Strontium titanate is assigned to Type S (f1=0.01) [94I1]. Strontium-90 emits β−-particles (0.20 MeV in average) with 100 % yield but does not emit energetic photons. Internally deposited Sr-90 is therefore measured by β counting of a urine sample following chemical separations. A typical detection limit readily achievable in monitoring programme is 1 Bq/l of 90 Sr in urine [97I2]. The decay product of 90Sr, 90Y is radioactive (half-life = 64 h), which emits high-energy β−-particles (0.99 MeV in average) with 100 % yield per disintegration of 90Sr. Strontium-90/yttrium-90 in the body can sometimes be measured by photon detectors positioned outside the body via the bremsstrahlung produced, though the minimum detectable activities are relatively high [99I2]. 10
0
0
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
a
10
10
Sr-90, f 1=0.3
Fraction of ingested activity [-]
Fraction of inhaled activity [-]
Sr-90, Type F, 5 µm 10
Whole body Urine Faeces Skeleton 0
10
1
2
10 10 Time after intake [d]
3
10
4
b
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Urine Faeces Skeleton 0
10
1
2
10 10 Time after intake [d]
3
4
10
0
Fraction of injected activity [-]
Sr-90
c
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Urine Faeces Skeleton 0
10
1
2
10 10 Time after intake [d]
3
10
4
Fig. 7.17. Predicted whole-body/skeletal retention, or daily urinary/faecal excretion of 90Sr as a function of time following acute intake of unit activity of 90Sr via a inhalation of particulate aerosols of Type F compounds with the AMAD of 5 µm, b ingestion of compounds whose f1 value is assumed to be 0.3, c injection of 90Sr in soluble forms.
Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-59
Ruthenium-106 (half-life = 1.01 y) Oxides and hydroxides of ruthenium are assigned to Type S (f1=0.05), halides to Type M (f1=0.05) and all other compounds to Type F (f1=0.05) by ICRP in Publication 68 for workers [94I1]. Though 106Ru does not emit energetic photons, the radioactive decay product 106Rh (half-life = 30 s) emits γ-rays of 0.512 MeV (20.6 % per disintegration of 106Ru), 0.622 MeV (9.8 %) and 1.050 MeV (1.5 %). They are penetrable radiations and therefore readily detected by photon detectors positioned outside the body. A typical detection limit readily achievable in monitoring programme is 200 Bq of 106 Ru in the whole body [97I2]. Gamma-ray spectrometry on biological samples permits detection of 5 Bq/l of 106Ru in urine [97I2]. 0
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
a
10
10
10 Ru-106, Type M, 5 µ m
Whole body Urine Faeces 0
10
1
2
10 10 Time after intake [d]
3
0
Ru-106, f1=0.05
Fraction of ingested activity [-]
Fraction of inhaled activity [-]
10
10
4
b
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Urine Faeces 0
10
1
2
10 10 Time after intake [d]
3
10
4
0
Fraction of injected activity [-]
Ru-106
c
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Urine Faeces 0
Landolt-Börnstein New Series VIII/4
10
1
2
10 10 Time after intake [d]
3
10
4
Fig. 7.18. Predicted whole-body retention, or daily urinary/faecal excretion of 106Ru as a function of time following acute intake of unit activity of 106Ru via a inhalation of particulate aerosols of Type M compounds with the AMAD of 5 µm, b ingestion of compounds whose f1 value is assumed to be 0.05, c injection of 106Ru in soluble forms.
7-60
7 Internal dosimetry of radionuclides
[Ref. p. 7-68
Iodine-131 (half-life = 8.04 d) Elemental iodine vapour is assigned to Class SR-1 (10 % deposition in ET1, 40 % in ET2, 50 % in BB), with Type F clearance [94I1, 95I2]. Methyl iodide gas is assigned to Class SR-1 (70 % deposition in ET2 and the lungs), with Type V clearance [95I2]. For workers particulate aerosols of iodine compounds are all assigned to Type F (f1=1.0) [94I1]. Iodine-131 emits γ-rays of 0.284 MeV (6.1 % yield), 0.364 MeV (81.2 %), and 0.637 MeV (7.3 %). The principal γ-ray emissions at 0.364 MeV are used for measurement of 131I by photon detectors positioned just outside the thyroid, in which iodine is highly accumulated. A typical detection limit readily achievable in monitoring programme is 100 Bq of 131I in the thyroid [97I2]. Gamma-ray spectrometry on biological samples permits detection of 1 Bq/l of 131I in urine [97I2]. 10
0
10
0
I-131, f1=1
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
a
10
10
Fraction of ingested activity [-]
Fraction of inhaled activity [-]
I-131, Vapour 10
Thyroid Urine Faeces 0
10
1
2
10 10 Time after intake [d]
3
10
4
b
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Thyroid Urine Faeces 0
10
1
2
10 10 Time after intake [d]
3
10
4
0
Fraction of injected activity [-]
I-131
c
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Thyroid Urine Faeces
0
10
1
2
10 10 Time after intake [d]
3
10
4
Fig. 7.19. Predicted thyroid retention, or daily urinary/faecal excretion of 131I as a function of time following acute intake of unit activity of 131I via a inhalation of iodine vapour, b ingestion of compounds whose f1 value is assumed to be 1.0 (0.99 was taken in this publication for computational reasons [95I2]), c injection of 131I in soluble forms.
Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-61
Caesium-134 (half-life = 2.06 y) All compounds of caesium possibly present in work place are assigned to Type F (f1=1.0) by ICRP [94I1] although it is recognised that other forms may be present in the environment [95I2]. Caesium-134 emits γ-rays of 0.563 MeV (8.4 % yield), 0.569 MeV (15.4 %), 0.605 MeV (97.6 %), 0.796 MeV (85.4 %) and 0.802 MeV (8.7 %), which are penetrable radiations and therefore readily detected by photon detectors positioned outside the body. A typical detection limit readily achievable in monitoring programme is 50 Bq of 134Cs in the whole body [97I2]. Gamma-ray spectrometry on biological samples permits detection of 1 Bq/l of 134Cs in urine [97I2]. 0
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
a
10
10
10
Cs-134, Type F, 5 µm
Whole body Urine Faeces 0
10
1
2
10 10 Time after intake [d]
3
10
0
Cs-134, f 1=1
Fraction of ingested activity [-]
Fraction of inhaled activity [-]
10
4
b
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Urine Faeces 0
10
1
2
10 10 Time after intake [d]
3
10
4
0
Fraction of injected activity [-]
Cs-134
c
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Urine Faeces 0
Landolt-Börnstein New Series VIII/4
10
1
2
10 10 Time after intake [d]
3
10
4
Fig. 7.20. Predicted whole-body retention, or daily urinary/faecal excretion of 134Cs as a function of time following acute intake of unit activity of 134Cs via a inhalation of particulate aerosols of Type F compounds with the AMAD of 5 µm, b ingestion of compounds whose f1 value is assumed to be 1.0 (0.99 was taken in this publication for computational reasons [95I2]), c injection of 134Cs in soluble forms.
7-62
7 Internal dosimetry of radionuclides
[Ref. p. 7-68
Caesium-137 (half-life = 30.0 y) All compounds of caesium possibly present in work place are assigned to Type F (f1=1.0) by ICRP [94I1] although it is recognised that other forms may be present in the environment [95I2]. Though 137Cs does not emit energetic photons, the radioactive decay product 137mBa (half-life = 2.55 min) emits γ-rays of 0.662 MeV (85.0 % per disintegration of 137Cs), which are penetrable radiations and therefore readily detected by photon detectors positioned outside the body. A typical detection limit readily achievable in monitoring programme is 50 Bq of 137Cs in the whole body [97I2]. Gamma-ray spectrometry on biological samples permits detection of 1 Bq/l of 137Cs in urine [97I2]. 0
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
a
10
10
10
Cs-137, Type F, 5 µm
Whole body Urine Faeces 0
10
1
2
10 10 Time after intake [d]
3
10
0
Cs-137, f 1=1
Fraction of ingested activity [-]
Fraction of inhaled activity [-]
10
4
b
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Urine Faeces 0
10
1
2
10 10 Time after intake [d]
3
10
4
0
Fraction of injected activity [-]
Cs-137
c
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Urine Faeces 0
10
1
2
10 10 Time after intake [d]
3
10
4
Fig. 7.21. Predicted whole-body retention, or daily urinary/faecal excretion of 137Cs as a function of time following acute intake of unit activity of 137Cs via a inhalation of particulate aerosols of Type F compounds with the AMAD of 5 µm, b ingestion of compounds whose f1 value is assumed to be 1.0 (0.99 was taken in this publication for computational reasons [95I2]), c injection of 137Cs in soluble forms.
Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-63
Cerium-144 (half-life = 284 d) Oxides, hydroxides and fluorides of cerium are assigned to Type S (f1=0.0005) and all other compounds to Type M (f1=0.0005) by ICRP in Publication 68 [94I1]. Cerium-144 emits γ-rays of 0.080 MeV (1.6 % yield) and 0.134 MeV (10.8 %). The radioactive decay product of 144Ce, 144Pr emits γ-rays of 0.697 MeV (1.5 % per disintegration of 144Ce). Because of their low abundances, detection limits of in vivo counting for 144Ce are relatively high; a typical detection limit that can be readily achieved is 10 kBq of 144Ce in the whole body [88I1]. Detection limits lower than this value are required for routine monitoring. Urine monitoring is not recommended, because cerium in the body is tenaciously retained and hardly excreted (biological half-life = 3500 d [79I1]). 10
0
10
0
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
a
10
10
Ce-144, f1=0.0005
Fraction of ingested activity [-]
Fraction of inhaled activity [-]
Ce-144, Type M, 5 µm 10
Whole body Urine Faeces 0
10
1
2
10 10 Time after intake [d]
3
10
4
b
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Faeces
0
10
1
2
10 10 Time after intake [d]
3
10
4
0
Fraction of injected activity [-]
Ce-144
c
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Urine Faeces 0
Landolt-Börnstein New Series VIII/4
10
1
2
10 10 Time after intake [d]
3
10
4
Fig. 7.22. Predicted whole-body retention, or daily urinary/faecal excretion of 144Ce as a function of time following acute intake of unit activity of 144Ce via a inhalation of particulate aerosols of Type M compounds with the AMAD of 5 µm, b ingestion of compounds whose f1 value is assumed to be 0.0005, c injection of 144Ce in soluble forms.
7-64
7 Internal dosimetry of radionuclides
[Ref. p. 7-68
Uranium-234 (half-life = 2.44×105 y), -235 (half-life = 7.04×108 y), -238 (half-life = 4.47×109 y) In Publication 68 [94I1], ICRP assigned most hexavalent compounds of uranium, e.g. UF6, UO2F2 and UO2(NO3)2 to Type F (f1 = 0.02), less soluble compounds, e.g. UO3, UF4, UCl4 and most other hexavalent compounds to Type M (f1 = 0.02) and highly insoluble compounds, e.g. UO2 and U3O8 to Type S (f1 = 0.002). Principal isotopes of uranium (234U, 235U, 238U) are α-emitting radionuclides and do not emit energetic photons except for 235U. Internally deposited uranium-isotopes are therefore measured by α-spectrometry on biological samples following radiochemical separation. A typical detection limit is 10 mBq/l in urine and 10 mBq in faeces [97I2]. Uranium-235 emits γ-rays of 0.144 MeV (10.5 % yield), 0.186 MeV (54.0 %) and 0.205 MeV (4.7 %). They are used for lung counting of 235U. A typical detection limit is 200 Bq [97I2]. For routine monitoring, the detection limits for α-spectrometry are adequate, but those for lung counting would not permit detection of intakes at annually limited levels [97I2]. 10
0
10
0
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
a
10
10
U-234/235/238, f 1=0.02
Fraction of ingested activity [-]
Fraction of inhaled activity [-]
U-234/235/238, Type M, 5 µm 10
Lungs Urine Faeces Skeleton 0
10
1
2
10 10 Time after intake [d]
3
10
4
b
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Urine Faeces Skeleton
0
10
1
2
10 10 Time after intake [d]
3
10
4
0
Fraction of injected activity [-]
U-234/235/238
c
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Whole body Urine Faeces Skeleton 0
10
1
2
10 10 Time after intake [d]
3
10
4
Fig. 7.23. Predicted whole-body/lung/skeletal retention, or daily urinary/faecal excretion of 234/235/238U as a function of time following acute intake of unit activity of 234/235/238 U via a inhalation of particulate aerosols of Type M compounds with the AMAD of 5 µm, b ingestion of compounds whose f1 value is assumed to be 0.02, c injection of 234/235/238U in soluble forms.
Landolt-Börnstein New Series VIII/4
Ref. p. 7-68]
7 Internal dosimetry of radionuclides
7-65
Plutonium-239 (half-life = 2.41×104 y), -240 (half-life = 6.54×103 y) In Publication 68 [94I1], ICRP assigned insoluble oxides of plutonium, e.g. high-fired PuO2, a common chemical form in nuclear industry, to Type S (f1=0.00001) and all other compounds to Type M (f1=0.0005). Among Type M compounds, f1-value of nitrates is assumed to be 0.0001 [94I1]. Plutonium-239/240 are α-emitting radionuclides and do not emit energetic photons. Internally deposited 239/240Pu are therefore measured by α-spectrometry on biological samples following radiochemical separation. A typical detection limit is 1 mBq/l in urine and 1 mBq in faeces [97I2]. Emission of low energy characteristic X-rays (0.014 - 0.020 MeV) are used for lung counting of 239/240Pu. A typical detection limit is 2 kBq, which is not adequate for routine monitoring [97I2]. 100
10
0
Pu-239/240, Type S, 5 µm
Pu-239/240, Type M, 5 µm
a
10-2
Fraction of inhaled activity [-]
Fraction of inhaled activity [-]
10-1
10-3 10-4 10-5 10-6 10-7
Lungs Urine Faeces Skeleton
10-8 100
101
10
102 103 Time after intake [d]
104
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
b
0
10
10
Lungs Urine Faeces Skeleton
0
10
1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Skeleton Urine Faeces
0
Landolt-Börnstein New Series VIII/4
10
1
2
10 10 Time after intake [d]
3
10
4
10
4
Pu-239/240, f1=0.0001
Fraction of ingested activity [-]
Fraction of ingested activity [-]
c
-1
3
0
Pu-239/240, f1=0.0005 10
2
10 10 Time after intake [d]
10
4
d
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Skeleton Urine Faeces
0
10
1
2
10 10 Time after intake [d]
3
7-66
7 Internal dosimetry of radionuclides 10
0
10
[Ref. p. 7-68
0
e
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Fraction of injected activity [-]
Fraction of ingested activity [-]
Pu-239/240, f 1=0.00001 10
Skeleton Urine Faeces
0
10
1
2
10 10 Time after intake [d]
3
10
4
f
10
-1
10
-2
10
-3
10
-4
10
-5
10
-6
10
-7
10
-8
10
Pu-239/240
Skeleton Urine Faeces 0
10
1
2
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Fig. 7.24. Predicted lung/skeletal retention, or daily urinary/faecal excretion of 239/240Pu as a function of time following acute intake of unit activity of 239/240Pu via a inhalation of particulate aerosols with the AMAD of 5 µm of Type M compounds, b Type S compounds, c ingestion of compounds whose f1 value is assumed to be 0.0005, d 0.0001, e 0.00001, f injection of 238/2390Pu in soluble forms.
Am-241 (half-life = 4.32×102 y) All compounds of americium possibly present in work place are assigned to Type M (f1=0.0005) by ICRP [94I1]. Based on several experimental results, ICRP considers that the trace contaminant 241Am that has grown from 241Pu in matrices of nuclear fuels behaves similarly to the bulk materials [95I2]. For this reason, Type S as well as Type M is taken in this publication. Americium-241 is α-emitting radionuclide accompanied with low-energy (0.060 MeV) γ-ray emission with 35.7 % yield. Internally deposited 241Am can be measured by both indirect and direct methods. A typical detection limit of α-spectrometry following radiochemical separation is 1 mBq/l in urine and 1 mBq in faeces [97I2]. These detection limits are adequate for both special and routine monitoring. Typical detection limits of in vivo measurements are 20 Bq for lungs and 20 Bq for skeleton. These detection limits are not necessarily adequate for routine monitoring [97I2].
Landolt-Börnstein New Series VIII/4
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7 Internal dosimetry of radionuclides
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Fig. 7.25. Predicted lung/skeletal retention, or daily urinary/faecal excretion of 241Am as a function of time following acute intake of unit activity of 241Am via a inhalation of particulate aerosols with the AMAD of 5 µm of Type M compounds, b Type S compounds, c ingestion of compounds whose f1 value is assumed to be 0.0005, d injection of 241 Am in soluble forms.
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7 Internal dosimetry of radionuclides
7.9 References 10B1 Bateman, H.: Proc. Cambridge Philos. Soc. 16 (1910) 423. 56M1 Marinelli, L.D., Miller, C.E.: Gamma Ray Activity of Contemporary Man. Science 124 (1956) 122. 59I1 ICRP: Recommendations of the International Commission on Radiological Protection. Report of committee II on permissible dose for internal radiation. ICRP Publication 2. Oxford: Pergamon Press, 1959. 68B1 Berger, M.J.: MIRD Pamphlet No. 2. Energy Deposition in Water by Photons from Point Isotropic Sources; J Nucl. Med. 9: Suppl. No. 1 (1968) 15-25. 69S1 Snyder, W.S., Ford, M.R., Warner, G.G., Fisher, H.L.: Estimates of absorbed fractions for monoenergetic photon sources uniformly distributed in various organs of a heterogeneous phantom. J. Nucl. Med. 10 (suppl 2) (1969) 5. 71B1 Berger, M.J.: Distribution of absorbed dose around point sources of electrons and Beta particles in water and other media. J. Nucl. Med. 12 (suppl 5) (1971) 5. 71G1 Gear, C.W.: Numerical initial value problems in ordinary differential equations, Englewood Cliffs, N.J.: Prentice Hall, 1971. 73S1 Spencer, L.V., Simmons, G.L.: Improved Moment Method Calculations of Gamma-Ray Transport: Application to Point Isotopic Sources in Water; Nucl. Sci. Eng. 50 (1973) 20-31. 74H1 Hindmarsh, A.C.: GEAR: Ordindary differential equations system solver, Lawrence Livermore National Laboratory Report UCID-30001, Rev. 3, 1974. 74S1 Skrable, K., French, C., Chabot, G., Major, A.: General equation for the kinetics of linear first order phenomena. Health Phys. 27 (1974) 155. 75I1 ICRP. Report on the Task Group on Reference Man. ICRP Publication 23. Oxford: Pergamon Press, 1975. 76L1 Loevinger, R., Berman, M.A.: Revised schema for calculating the absorbed dose from biologically distributed radionuclides: MIRD Pamphlet No 1, Revised. Soc. Nucl. Medicine, NY, 1976. 77I1 ICRP: Recommendations of the International Commission on Radiological Protection. ICRP Publication 26. Ann. ICRP 1 (3) (1977); reprinted (with additions) in 1977. 77U1 UNSCEAR: United Nations Scientific Committee on the Effect of Atomic Radiation Sources and Effect of Ionizing Radiation. 1977 Report to the General Assembly, with Annexes. Annex C – Radioactive Contamination due to Nuclear Explosions, 1977. 78S1 Snyder, W.S., Ford, M.R., Warner, G.G.: Medical Internal Radiation Dose Committee (MIRD) Pamphlet No. 5 (revised). New York, USA: Society of Nuclear Medicine, 1978. 79I1 ICRP: Limits for intakes of radionuclides by workers. ICRP Publication 30, Part 1. Oxford: Pergamon Press, 1979. 80I1 ICRP: Limits for intakes of radionuclides by workers. ICRP Publication 30, Part 2. Oxford: Pergamon Press, 1980. 80I2 ICRP: Limits for intakes of radionuclides by workers. ICRP Publication 30, Part 3. Oxford: Pergamon Press, 1980. 80I3 ICRP: Biological effects of inhaled radionuclides. ICRP Publication 31. Ann. ICRP 4 (1/2). Oxford: Pergamon Press, 1980. 80M1 Marshall, M., Stevens, D.C.: The Purposes, Methods and Accuracy of Sampling for Airborne Particulate Radioactive Materials; Health Phys. 39 (1980) 409-423. 80W1 Watson, S.B. and Ford, M.R. A User’s Manual to the ICRP Code- A Series of Computer Programs to Perform Dosimetric Calculations for the ICRP Committee 2 Report, Oak Ridge National Laboratory Report ORNL/TM-6980 (1980). 81V1 Vincent, J.H., Armbruster, I.: On the quantitative definition of the inhalability of airborne dust. Ann. Occup. Hyg. 24 (1981) 245. 83H1 Hindmarsh, A.C.: ODEPACK, a systematized collection of ODE solvers, in: Scientific Computing, Stepleman, R.S., et al. (eds.), Amsterdam: North-Holland, 1983, p. 55. Landolt-Börnstein New Series VIII/4
7 Internal dosimetry of radionuclides 83I1 83S1 83S2
84K1 84L1 85J1 86I1 86M1 87B1 87C1 87I1 87S1 88B1 88I1 88I2 88I3 88N1 88Z1 89B1 89I1 89M1 89T1 91I1 91I2 93I1
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ICRP.: Radionuclide transformations: Energy and intensity of emissions. ICRP Publication 38. Ann. ICRP 11-13. Oxford: Pergamon Press, 1983. Stather, J.W., Greenhalgh, J.R.: The metabolism of iodine in children and adults. NRPB-R140, Chilton, 1983. Stieve, F.E.: Exchange and transfer mechanisms of radioactive compounds between the mother and the developing off-spring in utero – Review of the experimental literature. In: Streffer, C., Patrick, G. (eds). Effects of prenatal irradiation with special emphasis on late effects. EUR 8067 EU, Commission of the European Communities, 1983, p. 159. Killough., G.G., Eckerman, K.F.: A Conversational Eigenanalysis Program for Solving Differential Equations. Proceedings of the 17th Midyear Topical Symposium of the Health Physic Society, Ed (R.L. Kathren, D.P. Higby, and M.A. McKinney) (1984). Leggett, R.W., Dunning jr., D.E., Eckerman, K.F.: Modelling the behaviour of chains of radionuclides inside the body. Radiat. Prot. Dosim. 9 (2) (1984) 77. Jacques, J.A.: Compartmental analysis in biology and medicine. Ann Arbor: The University of Michigan Press, 1985. ICRP: The metabolism of plutonium and related elements. ICRP Publication 48. Ann. ICRP 16 (2/3). Oxford: Pergamon Press, 1986. Mark, D., Vincent, J.H., Stevens, D.C., Marshall, M.: Investigation of the entry characteristics of dust samplers of a type used in the British nuclear industry. Atmos. Environ. 20 (1986) 2389. Bertelli, L., Lipsztein, J.L.: A mathematical simulation for the study of radionuclide kinetics in the human body. Radiat. Prot. Dosim. 18 (4) (1987) 209. Cristy, M., Eckerman, K.F.: Specific absorbed fractions of energy at various ages from internal photon sources. ORNL/TM-8381/v1-7. Oak Ridge, TN: Oak Ridge National Laboratory, 1987. ICRP: Radiation dose to patients from radiopharmaceuticals. ICRP Publication 53. Ann. ICRP 18 (1-4). Oxford: Pergamon Press, 1987. Stieve, F.E.: Placental transfer of other nuclides. In: Gerber, G.B., Métivier, H., Smith, H. (eds). Age-related factors in radionuclide metabolism and dosimetry. Dordrecht: Martinus Nijhoff, 1987, p. 315. Berkovski, V., Likhtarev, I., Ratia, G., Bonchuk, Y.: Internal dosimetry support system: Multipurpose research computer code. Radiat. Prot. Dosim. 79 (1-4) (1988) 371. ICRP: Individual monitoring for intakes of radionuclides by workers: Design and interpretation. ICRP Publication 54. Ann. ICRP 19 (1-3) (1988). ICRP: Limits for intakes of radionuclides by workers: An addendum. ICRP Publication 30, Part 4. Ann. ICRP 19 (4). Oxford: Pergamon Press, 1988. ICRP: Protection of the patient in nuclear medicine. ICRP Publication 52. Ann. ICRP 17 (4). Oxford: Pergamon Press, 1988. NEA/OECD: Gastrointestinal absorption of selected radionuclides. A report by an NEA expert group. Paris: Nuclear Energy Agency/OECD, 1988. Zankl, M., Veit, R., Williams, G., Schneider, K., Fendel, H., Petoussi, N., Drexler, G.: Radiat. Environ. Biophys. 27 (1988) 153. Birchall, A., James, A.C.: A microcomputer algorithm for solving first-order compartmental models involving recycling. Health Phys. 56 (6) (1989) 857. ICRP: Age-dependent doses to members of the public from intake of radionuclides. ICRP Publication 56, Part 1. Ann. ICRP. 20 (2). Oxford: Pergamon Press, 1989. Marshall, M., Stevens, D.C.: The purposes, methods and accuracy of sampling for airborne particulate radioactive materials. Health Phys. 39 (1989) 409. Tanaka, G., Nakahara, Y., Nakajima, Y.: Nippon Acta Radiol. 49 (1989) 344. ICRP: 1990 Recommendations of the ICRP. ICRP Publication 60. Ann. ICRP 21 (1-3). Oxford: Pergamon Press, 1991. ICRP: Addendum 1 to Publication 53 – Radiation dose to patients from radiopharmaceuticals. ICRP Publication 62. Ann ICRP 22 (3). Oxford: Pergamon Press, 1991. ICRP: Age-dependent doses to members of the public from intake of radionuclides: Part 2, Ingestion dose coefficients. ICRP Publication 67. Ann. ICRP 23 (3/4). Oxford: Elsevier Science Ltd, 1993.
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7-70 93L1 94I1 94I2 94Z1 95D1 95H1 95I1 95I2 95S1
96D1 96E1 96I1 96I2 96I3 96P1 97A1 97I1 97I2 98I1 98N1 98P1 98S1 98U1 99C1
7 Internal dosimetry of radionuclides Leggett, R.W., Eckerman, K.F., Williams, L.R.: An elementary method for implementing complex biokinetic models. Health Phys. 64 (3) (1993) 260. ICRP: Dose coefficients for intake of radionuclides by workers. ICRP Publication 68. Ann. ICRP 24 (4). Oxford: Elsevier Science Ltd, 1994. ICRP: Human respiratory tract model for radiological protection. ICRP Publication 66. Ann. ICRP 24 (1-3). Oxford: Pergamon Press, 1994. Zubal, I.G., Harrell, C.R., Smith, E.O., Rattner, Z., Gindi, G., Hoffer, P.B.: Med. Phys. 21 (1994) 299. Dorrian, M.D., Bailey, M.R.: Particle size distribution of radioactive aerosols measured in the workplace. Radiat. Prot. Dosim. 60 (1995) 119. Hubbell, J.H., Seltzer, S.M.: Tables of X-ray mass attenuation coefficients and mass energyabsorption coefficients, NISTIR 5632, Gaithersburg, MD: National Institute of Standards and Technology, 1995. ICRP: Age-dependent doses to members of the public from intake of radionuclides: Part 3: Ingestion dose coefficients. ICRP Publication 69. Ann. ICRP 25 (1). Oxford: Elsevier Science Ltd, 1995. ICRP: Age-dependent doses to members of the public from intake of radionuclides: Part 4: Inhalation dose coefficients. ICRP Publication 71. Ann. ICRP 25 (3-4). Oxford: Elsevier Science Ltd, 1995. Stabin, M.G., Watson, E.E., Cristy, M., Ryman, J.C., Eckerman, K.F., Davis, J.L., Marshall, D., Gehlen, M.K.: Mathematical models and specific absorbed fractions of photon energy in the nonpregnant adult female and at the end of each trimester of pregnancy. Oak Ridge, TN: Oak Ridge National Laboratory. ORNL/TM-12907, 1995. Dimbylow, P.J.: Proc. Voxel phantom development 6-7 July 1996, Dimbylow, P.J. (ed.), Chilton, UK: National Radiological Protection Board, 1996, p. 1. EC. Council Directive 96/29EURATOM of 13 May 1996, laying down the basic safety standards for the protection of the health of workers and the general public against the dangers arising from ionising radiation. Off. J. Eur. Commun. 39, No. L159 (1996). ICRP: Age-dependent doses to members of the public from intake of radionuclides: Part 5: Compilation of ingestion and inhalation dose coefficients. ICRP Publication 72. Ann. ICRP 26 (1). Oxford: Elsevier Science Ltd, 1996. IAEA: International basic safety standards for protection against ionising radiation and for the safety of radioactive sources. Jointly sponsored by FAO, IAEA, ILO, NEA/OECD, PAHO and WHO. Vienna, IAEA: Safety Series 115 (1996). IAEA: Direct methods for measuring radionuclides in man. Vienna, IAEA: Safety Series 114 (1996). Piechowski, J., Menoux, B.: Assessment of radioactive systemic uptakes by deconvolution of individual monitoring results. Health Phys. 70 (1996) 537. Ansoborlo, E., Boulard, D., Leguen, B.: Particle size distribution of Uranium aerosols measured in the French nuclear fuel cycle. Radioprotection 32 (1997) 219. ICRP: General principles for the radiation protection of workers. ICRP Publication 75. Ann. ICRP 27 (1). Oxford: Pergamon Press, 1997. ICRP: Individual monitoring for internal exposure of workers – replacement of ICRP Publication 54. ICRP Publication 78. Ann. ICRP 27 (3/4). Oxford: Pergamon Press, 1997. ICRP. Radiation Dose to Patients from Radiopharmaceuticals – Addendum 2 to ICRP Publication 53, also includes Addendum 1 to ICRP Publication 72. ICRP Publication 80. Ann. of the ICRP. 28(3). Pergamon Press, Oxford (1998). NRPB: Standards for intakes of radionuclides. Doc. NRPB 9 (4) (1998). Phipps, A.W., Jarvis, N.S., Silk, T.J., Birchall, A.: Time-dependent functions to represent the bioassay quantities given in ICRP Publication 78. NRPB-M824, Chilton, 1998. Spitzer, V.M. and Whitlock, D.G: Atlas of the Visible Human Male. Sudburg, MA: Jones and Bartlett (1998). Ulanovsky, A.V., Eckerman, K.F.: Absorbed fractions for electron and photon emissions in the developing thyroid: Fetus to five year old. Radiat. Prot. Dosim. 79 (1-4) (1998) 419. Caon, M., Bibbo, G., Pattison, J.: Phys. Med. Biol. 44 (1999) 2213. Landolt-Börnstein New Series VIII/4
7 Internal dosimetry of radionuclides 99E1 99I1 99I2 99S1 99W1 00A1 00G1 00X1 01I1 01P1 01S1 01Z1 02I1 02I2 02I3 02I4 02N1 02P1 02S1 03C1 03E1 03G1 03H1 03K1 03M1 03P1 03S1
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EPA.: Cancer risk coefficients for environmental exposure to radionuclides: Federal Guidance Report No 13, EPA Number: 402R99001. Publisher Info National Service Center for Environmental Publications, PO Box 42419 Cincinnati, OH 45242-2419, 1999. ICRP: The ICRP database on dose coefficients: Workers and members of the public (CD-ROM). Distributed by Elsevier Science Ltd, Oxford. ISBN 0-08-043876, 1999. IAEA: Assessment of occupational exposure due to intakes of radionuclides. Safety Guide No. RS-G-1.2, Vienna: IAEA, 1999. Stather, J.W.: Dosimetry and effects of radioactive hot particles. Radiat. Res. Vol. 2: Dublin: Proceedings of 11th ICRR, 1999. WHO: Guidelines for iodine prophylaxis following nuclear accidents: Update 1999, Geneva: WHO/SDE/PHE/99.6, 1999. Anderson, E., Bai, Z., Bischof, C., Blackford, L.S., Demmel, J., Dongarra, J., Du Croz, J., Greenhaum, A., Hammarling, S., McKenney, A., Sorensen, D.: LAPACK Users Guide, SIAM, Philadelphia, Third Edition (2000). Gössner, W., Masse, R., Stather, J.W.: Cells at risk for dosimetric modelling relevant to bone tumour induction. Radiat. Prot. Dosim. 92 (1-3) (2000) 209. Xu, X.G, Chao, T.C., Bozkurt A.: Health Phys. 78 (2000) 476. ICRP: Doses to the embryo and fetus from intakes of radionuclides by the mother. Corrected version issued May 2002. ICRP Publication 88. Ann. ICRP 31 (1-3) (2001). Polig, E.: Modeling the distribution and dosimetry of internal emitters: A review of mathematical procedures using matrix methods. Health Phys. 81 (5) (2001) 492. Saito, K., Wittmann, A., Koga, S., Ida Y., Kamei, K., Zankl, M.: Radiat. Environ. Biophys. 40 (2001) 69. Zankl, M., Wittmann, A.: Radiat. Environ. Biophys 40 (2001) 153. ICRP: Guide for the practical applications of the ICRP Human Respiratory Tract Model. Supporting Guidance 3. Ann. ICRP 32 (1-2) (2002). ICRP: The ICRP database of dose coefficients: Embryo and fetus (CD-ROM2). Distributed by Elsevier Science Ltd, Oxford. ISBN 0-08-044188-2, 2002. International Commission on Radiological Protection: ICRP Publication 89. Oxford, UK: Pergamon Press, 2002. Ishigure, N., Nakano, T., Enomoto, H., Matsumoto, M: Graphic Database on Predicted Monitoring Data for Intakes of Radionuclide. http://www.nirs.go.jp:8080/anzendb/RPD/gpmd.php (2002). Nipper, J., Williams, J., Bolch, W.: Phys. Med. Biol. 47 (2002) 3143. Petoussi-Henss, N., Zankl, M., Fill, U., Regulla, D.: Phys. Med. Biol. 47 (2002) 89. Stather, J.W., Phipps, A.W., Harrison, J.D., Eckerman, K.F., Smith, T.J., Fell, T.P., Noßke, D.: Dose coefficients for the embryo and foetus following intakes of radionuclides by the mother. J. Radiol. Prot. 22 (2002) 7. Charles, M.W., Mill, A.J., Darley, P.J.: Carcinogenic risk of hot-particle exposures. J. Radiol. Prot. 23 (2003) 5. Eckerman, K.F., Ulanovsky, A.V., Kerr, G.D.: Electron and photon absorbed fractions in the developing Fetus. ORNL/TM Report, 2003. Guilmette, R.A., Durbin, P.W.: Scientific basis for the development of biokinetic models for radionuclide-contaminated wounds. Radiat. Prot. Dosim. 105 (1-4) (2003) 213. Harrison, J.D., Smith, T.J., Phipps, A.W.: Infant doses from the transfer of radionuclides in mothers’ milk. Radiat. Prot. Dosim. 105 (1-4) (2003) 251. Kramer, R., Vieira, J.W., Khouri, H.J., Lima, F.R.A., Fuelle, D.: Phys. Med. Biol. 48 (2003) 1239. Métivier, H.A.: New model for the human alimentary tract. Radiat. Prot. Dosim. 105 (1-4) (2003) 43. Peace, M.S.: Practical experience of the application of ICRP models in internal dose assessment. Radiat. Prot. Dosim. 105 (1-4) (2003) 33. Stabin, M.G., Siegel, J.A.: Physical models and dose factors for use in internal dose assessment. Health Phys. 85 (3) (2003).
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Acknowledgements The authors are very grateful for the contributions from their colleagues in the development of this review. They would particularly wish to thank Henri Métivier, John D Harrison, Nina Petoussi-Henß and François Paquet. They are also grateful for the excellent technical assistance from Karen Roberts in the preparation of the manuscript.
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8 Decontamination
The first part of this Chapter describes decontamination techniques of large volume systems, segmented parts and walls and floors. Advantages and disadvantages of the different methods are discussed and examples of decontamination procedures and their results in laboratory- and large scale are demonstrated. Considerations are focussed on nuclear facilities and power plants during operation and decommissioning. The second part deals with decontamination of the human skin. In case of contamination of the human skin by radionuclides suitable measures have to be initiated to keep the dose to the skin below the limits recommended by the ICRP. For purposes of dose estimates numerical values of the equivalent dose rate in Sv/s at an activity per area of 1 Bq/cm2 are given for 128 radionuclides. In addition first aid and specific decontamination procedures are described as simple decontamination appliances immediately after contaminations or for decontamination of specific body regions and organs below reference values for residual contamination.
8.1 Decontamination of materials List of Abreviations AGR ALARA AMDA AP APAC APACE BWR CAN-DECON CANDU CEA CEC CEGB CITROX CORD DF EDTA EPRI Framatome HX LOMI
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Advanced Gas Cooled Reactor As Low As Reasonable Achievable Automated Mobile Decontamination Application Alkaline Permanganate Alkaline Permanganate Ammonium Citrate Alkaline Permanganate Ammonium Citrate EDTA Boiling Water Reactor CANdu DECONtamination CANada Deuterium Uranium Pressurized Heavy Water Reactor Commissariat à l'Energie Atomic (France) Community of European Countries Central Electricity Generating Board CITric Acid OXalic Acid Cyclic Oxidation Reduction Decontamination Decontamination Factor Ethylene-Diamine-Tetraacetic Acid Electric Power Research Institute (USA) Company's Name Heat Exchanger Low Oxidation-state Metal Ions
8-2 MOPAC NP NPP OZOX PWR SGHWR WAGR
8 Decontamination
[Ref. p. 8-34
Modified Permanganat Ammonium Citrate Nitric Acid Permanganate Nuclear Power Plant OZone OXalic Acid Pressurized Water Reactor Steam Generating Heavy Water Reactor Windscale Advanced Gas Cooled Reactor
8.1.1 General approaches to decontamination 8.1.1.1 Contamination Contamination is the deposition of the radioactive elements or compounds from a contaminant media or gas, by chemical, physical or other ways, on the surfaces of components, systems, and structures in nuclear facilities. The characteristics of the contamination are strongly correlated with the nature and features of the surface and of the contaminant media. In metallic surfaces there often exists a chemical similarity with the contaminant element (as for example for metallic cations in the water) that can cause its diffusion into the metallic sub-layer, hence becoming very difficult to remove. Contamination can be classified into three types: - free contamination; i.e. can be removed by simple blowing, vacuum or similar methods; - loose contamination; i.e. can be removed by common cleaning techniques; - fixed contamination; i.e. cannot be removed without removal of surface layers. The following presents typical contamination patterns encountered in nuclear facilities [88Int]. In reactor systems, the radioactive contamination on the internal surfaces is caused by deposition from the reactor coolant of neutron activated particles and dissolved elements, and of fission products and transuranics released following a failure of the fuel cladding. These deposits become part of the oxide layer, which forms on the inside of the piping. This layer has a complex structure, which depends on a variety of parameters such as coolant chemistry, temperature of formation, system materials, operating time, etc. Over long periods of time, the radionuclides in the layer can diffuse slightly into the base metal or penetrate the pipes along grain boundaries. In general for water cooled reactors, two types of oxides form on the internal surface of reactor piping: an adherent inner layer which is formed by in-situ corrosion of the base metal, and a relatively loose outer layer which is formed by deposition or precipitation of crud from the coolant. 8.1.1.2 Characteristics of oxide layer in BWRs and PWRs Occupational dose in BWRs and PWRs is mostly caused by corrosion-originated nuclides: 60Co, 58Co, 54 Mn, 51Cr and 59Fe. Depending on fuel failure rate, the other fission species would contribute to plant dose rates. Most part of these species is included in oxide layer inside pipes and equipment. Decontamination usually dissolves or removes the radioactive species together with the oxide layer. Characteristics of the oxide layers are quite different between BWRs and PWRs. Iron occupies 80-90 % of metal elements in the BWRs oxide layer. BWRs use stainless steels and carbon steel for the reactor cooling and the feedwater systems. PWRs use great amount of nickel-base alloys for steam generators. Metal fraction of Ni and Cr is 60-70 % in PWRs oxide layer. BWRs oxide layer grows in oxidising water chemistry and consists of α-Fe2O3, Fe3O4, and NiFe2O4. Reducing environment of PWRs forms (NixFe3-x-yCryO4)-type Cr-rich oxide layers. Table 8.1 compares the oxide characteristics of Japanese BWRs and PWRs. Usually oxide forms indistinct double layer. The inner layer grows from base metal and deposits tightly on the base metal surface. The outer layer contains fuel surface crud released with shear stress by primary coolant flow. In some specific cases, the outer layer crud is easily removed with ultrasonic vibration or high-pressure water jet. Landolt-Börnstein New Series VIII/4
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Table 8.1 Crud characteristics of Japanese BWRs and PWRs [00Hir1] Primary coolant chemistry Characteristics of primary system
Surface area percentage of materials Outer layer Form
Crud analysis
Inner layer
Metal fraction
BWRs Oxidising SS: 38-42 % Carbon steel: 16-20 % Zircaloy: 40-44 % Ni-base alloy: <1 % Fe3O4, NiFe2O4 (dominant) Fe3O4 (dominant), α-Fe2O3, NiFe2O4, FeCr2O4 Fe: 80-90 % Ni: 7-10 % Cr: 1-10 %
PWRs Reducing SS:4-6 % Zircaloy: 25-28 % Ni-base alloy 65-70 % Other alloys <1 % Fe3O4 (dominant), NiFe2O4, FeCr2O4 FeCr2O4, Fe2CrO4 Fe: Ni: Cr:
20-40 % 25-60 % 15-45 %
8.1.1.3 Other types of contamination In fuel reprocessing facilities, the acid process stream in the dissolution and separation steps of the process, tends to inhibit the formation of an internal oxide layer and deposition of radionuclides is thus limited. Nevertheless, after separation, the phase, which carries the uranium and plutonium, can form pasty and heavy deposits in the pipes and tanks. These deposits are often very difficult to remove. In other types of nuclear facilities such as hot cells and mixed oxide fuel fabrication plants, low levels of contamination can exist in process vessels, cells, etc. as a result of normal operation. In UO2 fuel fabrication plants low levels of activity are present from the processing of UO2. 300
Specific activity [Bq /g]
250
60
Co 137
Cs
200 150
Fig. 8.1 Penetration in concrete of 137Cs and 60Co in samples of Gundremmingen KRB A Reactor (concrete samples of floors were taken from the decommissioned nuclear power plant). DS = decontamination seal; filled columns and solid line: Cs-137; open columns and dasheddotted line Co-60.
100 50 0 DS 0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 Penetration depth [mm ]
In most nuclear facilities, many external surfaces become contaminated as a result of leakage and spillages from process systems and from demolition, maintenance and waste management activities. Airborne or waterborne activity can deposit out forming contamination layers on floors, equipment, instrumentation, etc. Thus surfaces can become contaminated by physical or chemical mechanisms. Of particular concern is the potential contamination of concrete surfaces by waterborne contamination. Unless the surface of the concrete is sealed, water-soluble radionuclides, such as 137Cs, can penetrate deeply into concrete. The only method for removing such contamination would be to cut or Landolt-Börnstein New Series VIII/4
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[Ref. p. 8-34
chip away the surface layer of concrete containing the radioactivity. To measure the penetration of the contamination in Gundremmingen KRB-A, concrete surfaces were drilled and analysed by gamma spectroscopy in the radiation protection laboratory to determine the nature and depth of the contamination. 137Cs and 60Co, and traces of 134Cs, were found to be present. The isotope 137Cs accounted for about 90 % of the radioactivity. As shown in Fig. 8.1, most of the activity is embedded in the decontamination seal, but some contamination with a high proportion of 137Cs penetrates to a greater depth than contamination in which 60Co is predominant [84Ebe]. For ventilation systems, the surface contamination is usually loose, although adherence is aided by oil films often found on the inside of ducts particularly downstream of fans. Since the exhaust systems operate at negative pressures they tend to draw in dusts and aerosols, which may contain activity. Deposition tends to be heavier in sections of the ducting where the direction or velocity of the fluid changes or at edges of joints or flanges. For motors, instrumentation and walls etc., loose airborne contamination is usually the major problem. This contamination can generally be removed if it is accessible. If motors and other delicate equipment need to be reused, ultrasonic and freon decontamination processes are sometimes used. 8.1.1.4 Decontamination Decontamination is the removal by chemical, physical or other methods, of surface radioactive material from both internal and external surfaces of components, systems and structures in nuclear facilities. Usually decontamination and cleaning are considered as separate processes even though they can often be the same physical process; the difference is the degree of cleaning and the emphasis on species removed. Decontamination is the removal of radioactive dirt and oxides from surfaces, whilst cleaning usually refers to the removal of non-radioactive materials. Decontamination should be considered to be a part of cleaning because, in general, only a small part of the material removed during decontamination is radioactive. It is important to note that decontamination is not the elimination of the radioactivity, just the removal to a different location. The term decontamination is widely used in reference to surfaces commonly in contact with contaminated agents (such as reactor coolant, off gases, etc.) whereas the term cleaning refers to surfaces only lightly contaminated by aerosols or by purge liquids, etc. The concept of decontamination was introduced at the birth of the nuclear industry and was used to describe the reduction of radiation levels on the surfaces of components, systems and structures in order to allow their maintenance, repair, and control works. The importance of decontamination, and the consequent development of new decontamination processes varied as the problems of reduction of radiation levels and man-sievert expenditures (including costs) affected the exploitation of nuclear stations and facilities. In the early 1960s, decontamination was already a common practice in the nuclear industry. In the mid 1970s, with the support of regulatory agencies and industries, decontamination processes became more sophisticated and a complete evaluation including environmental concerns, costs, legal and public requirements became a common feature of decontamination practices. In the late 1970s, a new emphasis was placed on the decommissioning of nuclear facilities and this introduced a new concept in decontamination, not only to reduce radiation levels, which is normally the major objective of decontamination, but also to facilitate waste management and, if possible, to permit reuse of the material or components. 8.1.1.5 The use of decontamination in decommissioning The techniques used in decontamination for decommissioning purposes have two main differences in comparison with common in-service decontamination techniques. The techniques can be allowed to affect the integrity of the base materials. This is because, in principle, the component or system will not be reused. The techniques should generate the minimum quantity of secondary wastes.
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The basic approach to decommissioning work must be to answer the question: whether, when and how decontamination is to be carried out. Hence, the decision to carry out decontamination is based on a compromise between the advantages and disadvantages of separating the radioactivity or dealing with the fully active installation or component. Decontamination processes could be carried out merely to ease the handling of materials or to bring the radioactivity down to a level where unrestricted release is possible. The latter objective presupposes the existence of such a level, agreed upon with the relevant authorities. Some examples are presented below of the above argument [90Ber]: a) b) c)
A system or a component is to be worked on for inspection, maintenance or modification, and it is so contaminated that the radiation levels in the work areas are above acceptable values with reference to the established regulations or to the ALARA principle; A facility, system or component is to be re-used for other purposes, which require it to be “free” from contamination; A facility, system or component is to be dismantled and the wastes arising from such proceedings are to be collected, conditioned, and disposed of in a safe and economical way according to national practices and regulations, and consistent with an operational national waste management system.
In case (a), the objective is only to obtain a reduction in the level of the contamination without damaging the components. Here the problem is to weight the overall decontamination costs, including management of the necessary provisions for safe work in radiation areas, such as reduced individual working hours, shields, remote operation, telemanipulation. In case (b), complete decontamination is the main goal and the problem is a question of feasibility i.e. is it “possible” to obtain the required level of decontamination, and of cost i.e. is it less expensive as a whole, to decontaminate, or to dismantle everything and use new components? Case (c) is common to all decommissioning strategies for nuclear facilities, although in some instances cases such as (a) and (b) may arise. In this case, it may be necessary to consider in more detail, factors that can influence the decision on whether to carry out decontamination in the first place. If decontamination is the preferred option, then there is also a wide range of processes available to choose from. 8.1.1.6 Identification of decontaminable components From the above, it can be concluded that decontamination is a useful tool in decommissioning work but it is not possible to generalise which parts should be decontaminated and, if so, how the decontamination should be carried out. In reality a priority list of components and systems required for decontamination needs to be drawn up [85Lör]. Components for decontamination should be identified as early as possible in order to avoid wasting time and money on unnecessary decontamination work. Simple decontamination to remove weakly adhesive contamination is still useful because it reduces radiation exposure and facilitates subsequent handling. One must first compare the two approaches, namely, decontamination for unrestricted release, or direct transfer to some type of repository for radioactive materials. The exposure of personnel and the respective costs of each approach must be considered. In addition to the main criteria above other factors should be taken into consideration, these are: • The type and degree of contamination; • The geometry of the components; • The mass of the components to be decontaminated.
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It is possible to determine the type and degree of contamination but careful consideration must be given to whether components exposed for long periods to high pressures and temperatures (e.g. primary circuit) can in fact be decontaminated, whereas components contaminated only by contact with air, whether through moisture or other factors, clearly are decontaminable. In between lies a grey area, which needs further investigation, based on operating data and/or tests, before a proper decision can be taken. The geometry of the components plays a major role. Decontamination processes are mostly not able to ensure even removal or cope with particular corners and cavities and these are the precise points at which substantially higher contamination can occur. With complicated geometries, the measurements required to produce evidence that the limits have been observed may not be possible or may only be carried out with great difficulty. The mass of components of the same type to be decontaminated, i.e. components that will be subjected to the same decontamination process, is important in two ways. Firstly, there is little likelihood that very large masses will be transported to a final repository, simply because they contain a few grams or kilograms of radioactive material. Secondly, the decontamination process used must be suitable for such large masses. When selecting the process, one must ensure that the time needed for decontamination remains within acceptable limits, since personnel costs, which are an important factor, increase in proportion to that time. Where the amounts received from the sale of decontaminated material are significant the process need not in principle spare the material; on the contrary it can be quite aggressive, since it is not planned to re-use the material in its original form. An estimate of the mass of material that might need to be decontaminated was made for two German nuclear power stations, one with a pressurised water reactor and one with a boiling water reactor. An effort was made to make a distinction on the basis of the above-mentioned criteria. The material was divided into three categories: nondecontaminable material, decontaminable material and material on which no decision can yet be taken. Components, which today are still very difficult to decontaminate, or where decontamination is not viable can naturally be reclassified at a later date on the basis of experience and in the light of progress in decontamination techniques. 8.1.1.7 Effectiveness of decontamination, decontamination factor The efficiency of different decontamination processes has to be evaluated. The common parameter is called the “decontamination factor”. This is “a numerical representation of the effectiveness of a decontamination process” and it is calculated as ratio between predecontamination and post-decontamination measurements, i.e.: DF = Mb / Ma
(8.1.1)
Where DF = decontamination factor (generally greater than 1) Mb = measurement “before” decontamination (at a reference point) and Ma = measurement “after” decontamination (at the same reference point as Mb). In terms of a percentage, the decontamination factor can be expressed as: DF (%) =
Mb − Ma Mb
(8.1.2)
In the following, only the definition (8.1.1) will be considered. With regard to the kind of measurements, which can be considered, the decontamination factor can be defined (or calculated) by two different methods [85Duc]. The first method is to use radiation measurements. This is called the “radiation DF” and is defined as: radiationDF =
Ib Ia
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Where Ib = dose (radiation) rate “before” decontamination (at a reference point) and Ia = dose (radiation) rate “after” decontamination (at the same reference point as Ib). This definition is widely used in decontamination for operating plant, where the radiation measurements referred to are dose rate area measurements; for this case the term “man-sievert DF” is also used. In many cases the radiation measurements can be taken from monitor counters located near or over the surface to be measured. The second methods is to use activity measurements. This is called the “decon DF” and is defined as: deconDF =
Ab Aa
(8.1.4)
where Ab = activity “before” decontamination (at a reference point) and Aa = activity “after” decontamination (at the same reference point as Ab) This definition is widely used for off-line decontamination where the surface activity can be properly measured. In laboratory studies and research, the decon DF is more widely used than the radiation DF. As mentioned before, each numerical value of DF must be referred to a single measurement before decontamination and a single measurement after decontamination. This means that the decontamination factor can be of relevance only for single points or for very small surfaces or components (which can be measured with a single operation). In any other case, where many measurements have been carried out it is necessary to calculate an “average DF”. The activity of the background is a parameter which is strongly dependent on the procedures and instrumental techniques used for the activity measurements. For evaluation of Decontamination Factors according to (8.1.3) shielding measurement devices are necessary to provide correct data. 8.1.1.8 Decontamination techniques (processes) Decontamination techniques may be classified in several different ways depending on the purpose of decontamination e.g. to save man-sieverts, for restoring the component/system, for decommissioning, the kind of decontamination media e.g. chemical, mechanical, electrochemical, etc., and on the nature of the surface required for decontamination (e.g. metal, concrete, painted surfaces, etc.). The most widely used criterion refers to the kind of decontamination media. Nevertheless different classifications have been proposed since in many cases it is not easy to clearly define the “decontamination media”. Some processes may combine several different decontamination media e.g. electropolishing, which combines chemical and electrical actions, or water jets used with detergents, which combine mechanical and chemical actions. In the following paragraphs some of the classifications of decontamination techniques proposed by different studies are presented: The US-Department of Energy (DOE) “Decommissioning Handbook” [80Man], in 1980, classified decontamination techniques into four categories: (i) chemical decontamination, (ii) manual and non-chemical decontamination, (iii) electropolishing, and (iv) ultrasonic/chemical decontamination • Chemical decontamination: Alkaline Permanganate (AP), Ammonium Citrate (AC), EDTA, Oxalic Acid (OX), Citrox, Sulphamic Acid, Hydrochloric Acid, Nitric Acid, Sulphuric Acid, Phosphoric Acid, Oxalic Peroxide (OP), Sulphox, Can - Decon, NS - 1 • High-pressure water lance • Electropolishing: in-tank and in-situ • Ultrasonic decontamination
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Further Classifications were given by: The Electric Power Research Institute (EPRI) [82Gar], in 1982, The US-Nuclear Regulatory Commission (NRC) [81Nel], in 1981, The Commissariat A l'Energie Atomique-France (CEA) [82Com] in 1982, and the International Atomic Energy Agency (IAEA) [83Int], Vienna, in 1983. 8.1.1.9 Decontamination and secondary waste generation The problem of secondary waste arising from decontamination works is of major concern in terms of quantities and characteristics. This problem is of particular relevance for chemical decontamination processes for decommissioning. In this context it became one of the leading factors for the real acceptance of the process. On-line chemical processes like Can-Decon, Cam-Derem, LOMI, CORD/OZOX and EMMA include phases of solution cleaning in the process itself and generate relatively low volumes of ion-exchange resins as secondary waste. With electropolishing decontamination using standard electropolishing solutions, such as concentrated phosphoric acid, the treatment of the spent solution by reprocessing and recycling the solutions in order to reduce the final volume of wastes for storage has to be considered.
8.1.2 Decontamination techniques for large volume closed systems 8.1.2.1 Reactor decontamination in BWRs and PWRs 8.1.2.1.1 Chemical decontamination principle Chemical decontamination reagent work to dissolve spinel oxide involving Fe3+, Ni2+ and Cr3+ on the inner surfaces of pipes and equipment. The three-valent ions have low solubility. The oxide can be dissolved as follows: Reducing dissolution: Fe-containing oxide is effectively dissolved with reducing reagent and acid following the reducing dissolution scheme: Fe2O3 + 6 H+ + 2 e− → 2 Fe2+ + 3 H2O Fe3O4 + 8 H+ + 2 e− → 3 Fe2+ + 4 H2O Acid dissolution: Acid can dissolve the spinel oxide. But acid alone also attacks the base metal: Fe3O4 + 8 H+ → Fe2+ + 2 Fe3+ + 4 H2O Fe + 2H+ → Fe2+ + H2 Oxidising dissolution: Cr-containing oxide is dissolved following the oxidising dissolution scheme with permanganate ion: Cr2O3 + 2 MnO4− + H2O → 2 HCrO4− + 2 MnO2 A pH-potential diagram is illustrated in Fig. 8.2. Chemical decontamination is an oxide-dissolving technique used not in the passive state area, but in the metal corrosive area (indicated with hatched line in the Fig. 8.2). Nearly all decontamination processes like Can-Decon, NS-1, LOMI and CORD reduce Fe3+ to Fe2+. Fig. 8.1 shows the Redox-potential as a function of pH-value of nitric-acid-permanganate (NP) and alkaline permanganate (AP). HP (permanganic acid) is in the same potential range of NP. All three oxidation methods (HP, NP and AP) oxidize Cr3+ to Cr6+ but have no ability to dissolve Fe2O3 and Fe3O4. NP solution is advantageous in the field of Fe/Cr/Ni - austenitic materials, AP in the field of Ni-alloys. Landolt-Börnstein New Series VIII/4
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H2CrO4
Cr2 O 27HCrO -4
1.6
Potential [ V vs.SHE ]
1.2
-
CrO 24
NP b
8-9
log C = - 4 Cr : Fe : Ni : Mn :
3+ 2+ 0.8 Fe / Fe
AP NS -1
0.4 0
LND 101A a
LOMI
Ni 2+/ Ni
Cr2 O3
- 0.4
Fe2 O3 Passive state
- 0.8
Fe 2+/Fe
Cr 3+/ Cr 2+
Fe3 O4
Cr 2+/ Cr
-1.2 -1.6 -2
Fig. 8.2 pH-potential diagram of chemical decontamination; [00Hir]
Cr (OH) 2
0
2
4
6 pH
8
10
12
14 16
Application of oxidation decontamination Metal fraction of Cr in the oxide layer strictly affects decontamination performance. In the case of lower chromium content than 7 %, reducing reagent achieves successful result. But in the cases of Cr content higher than 7 %, very few reducing reagent works well without the steps of oxidising decontamination. The oxidising reagent dissolves Cr and breaks spinel structure before the reducing step. 8.1.2.1.2 Testing material compatibility during and after decontamination Before application of any decontamination process it has to be qualified in laboratory tests. Table 8.2 lists evaluation issues to confirm materials compatibility during laboratory test. This basis qualification covers the influence of the solvent during decontamination and the post operation behaviour in the NPP systems. In addition during application on site test coupons can be inserted to the decontamination circuit to monitor the corrosion and IGSCC. Occasionally, actual pipes are taken from the plant and used for own compatibility evaluation. Materials compatibility during decontamination General corrosion is evaluated with weight loss measurement and surface/cut surface observation of test coupons. General corrosion is a key issue of carbon steel and low alloy steels rather than stainless steels and nickel base alloys. Flowrate of decontamination liquid, temperature and inhibitor effect should be taken into account. No galvanic corrosion effect is reported on the present existing decontamination methods for BWRs and PWRs. To evaluate pitting or inter-granular corrosion, new test coupons might be inappropriate. Probes from actual tube materials that experienced plant operation history should be tested. Corrosion in crevices caused by residual decontamination reagents should be evaluated.
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Table 8.2 Material testing for decontamination reagent evaluation [00Kat]
During decontamination
Evaluation issues General corrosion Pitting, crevice corrosion Inter-granular corrosion Long term compatibility against IGSCC Crack growth rate of SCC
After decontamination
Corrosion with residual decontamination reagent Effect of decontamination repetition Recontamination
Materials Carbon steel Carbon steel Low alloy steel Stainless steel Ni base alloy Stainless steel Ni-base alloy Stainless steel Ni-base alloy All reactor materials All reactor materials All reactor materials
Material behaviour after decontamination is one of the most concerning issues. Results of a systematic survey conducted in western countries are described below. United States In 1980s, London Nuclear Ltd., GE and other organisations tested material compatibility to apply reactor decontamination mainly for BWRs and Candu reactors. The tests were mainly on IGSCC susceptibility of type 304 stainless steel (SS) and nickel base alloys (Inc 600). Can-Decon, PNS-Citrox, and LOMI were tested as listed in Table 8.3 [00Kat1]. Can-Decon slightly increased SCC susceptibility of type 304 stainless steel, therefore the process was modified to a process called “CANDEREM”. 8.1.2.1.3 LOMI, Can-Decon/CANDEREM and CORD / CORD UV LOMI, Can-Decon/CANDEREM and CORD / CORD UV were the most important processes for the last 10 years, LOMI mainly in the US, CANDEREM in Canada and US, CORD UV in Western Europe and Japan. In recent years in the US a revival of the CITROX process could be observed instead of applying LOMI and CANDEREM. LOMI was developed by CEGB with the target to dissolve only iron oxide (hematite) located on fuel elements within the core of the SGHWR. By adding an AP and NP oxidation step the application range of the process was extended mainly to BWR. Until 1990s, Can-Decon was applied to BWR recirculation systems and other actual plants at least 34 times. After the evaluation regarding IGSCC, CANDEREM was developed. The CANDEREM uses EDTA and citrate but no oxalate, which was the main chemical of Can-Decon. Both processes were developed by AECL. The CANDEREM is used at a temperature higher than 100 °C. In 1991, the CORD method developed by Siemens/KWU was applied to NPP Isar, a BWR in Germany. The CORD process strictly controls pH-value and corrosion potential using oxalic acid as reducing reagent. As a preoxidation step permanganic acid (HP) is applied. Many plants including Japanese BWRs have applied CORD since 1991 to the present. Since 1994 the CORD process was improved to the CORD UV process, in which the decontamination chemicals are destroyed during the process to carbon dioxide and water. There has been no negative information regarding IGSCC with the CORD/CORD UV process. Many plants have used this process. There is a lot of experience available on this decontamination technique. Questions how circulate and to heat the solutions are well experienced. As example Fig. 8.3 shows the Fukushima Daiichi, Unit 2 Flow Diagram for decontamination [00Wil]. In total, more than 400 full system decontaminations have been performed with the Cord process. Landolt-Börnstein New Series VIII/4
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Table 8.3 Material-testing results for decontamination reagent evaluation in the US [00Kat1] Decon reagent
Test method Tube test
304 SS
Can-Decon (LND-101A)
SSRT
Can-Decon (LND-104, LND101A)
Double Ubend 4-pointsupporting bend test SSRT
Can-Decon (Nutek L-106, LND-101)
Can-Decon (LND-104) Can-Decon, LOMI
SSRT Stress beam Test
PNS-Citrox
SSRT
LOMI, Can-Decon PNS-Citrox
SSRT Stress beam Test
LOMI
Tube test Crack growth rate measurement SSRT
Can-Decon (LND-101A)
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Water chemistry BWR (DO: 8ppm)
Results
304 SS
BWR (DO: 0.2ppm)
304SS, Inconel 600
PWR (DH: >3ppm)
No effect on SCC susceptibility in 100 h decon, but 500 h decon. Increased on sensitised 304 SS. No susceptibility increase to SCC
304SS, Inconel 600 304SS, Low alloy steel, (SA533B) Inconel 600
PWR (DH: >3ppm) BWR (DO: 0.2ppm)
No susceptibility increase to SCC
304SS, Low alloy steel, (SA533B) Inconel 600 304SS, Inconel 600
BWR (DO: 0.2ppm)
Sensitized 304SS increased susceptibility to SCC in some cases.
BWR (DO: 0.2ppm)
Can-Decon and PNSCitrox increased susceptibility of sensitised 304SS to SCC
304 SS 316 NG SS
BWR (DO: 0.2ppm)
No susceptibility increase to SCC
304SS
BWR (DO: 0.2ppm)
No susceptibility increase to SCC
No acceleration on propagation rate of existing crack
Can-Decon increased 304SS SCC susceptibility. No effect with LOMI
Corrosion
Reporting organisation London 50µm IGA was observed Nuclear Ltd. on sensitised [85EPR] 304 SS. No descrip- Ontario tion Hydro [85EPR1] Pitting on 304 SS and 1.5µm IGA symptom on Inconel 600 No IGA
London Nuclear Ltd. [85EPR2] London Nuclear Ltd. [85EPR3] G.E. [86EPR]
Can-Decon caused 200µm IGA on 304SS and 100µm IGA on Inconel 600. No IGA with LOMI. G.E. 80µm IGA on 304SS and [86EPR1] 60µm IGA on Inconel 600 IGA were observed as the same extent as the above two columns No description
G.E. [86Man]
G.E. [86Man]
Slight IGA G.E. symptom was [86Man] detected in some cases
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[Ref. p. 8-34 Existing Recirculation System
Vent UV Modules
Temporary Decon. Equipment
Spray Ring
Chemical Injection Tank Chemical Injection Tank
RPV Pump 20% speed Reactor Recirculation System Pump
CRD Housing Decon Loops: 200 m3/ h /each
Demineralizer
Main Circulation Pump
Heater
Cooler
Fig. 8.3 Fukushima 1 Unit 2 Flow diagram; [00Wil]
8.1.2.2 Fuel assemblies and decontamination Fuel surface crud usually contains a 100 to 1000 times higher inventory of radioactivity than recirculation system oxide layers. A decontamination reagent is easily decomposed by gamma rays and neutrons from the fuel even in reactor lay down periods. In spite of these difficult conditions fuel elements may have to be decontaminated for two reasons. In case the crud deposits cause too high pressure drops to achieve a homogeneous reactor coolant water flow cleaning is required. In addition, the treatment of spent fuel elements depends on a certain degree of cleanliness. The SGHWR (100 MWe, closed in 1991, UK) and Candu reactors often performed the decontamination of fuel assemblies in situ. The LOMI developed for this purpose has self-regenerating ability in the presence of radiolysis. Pressure-tube-type reactors are usually designed for on-power refueling. This requires more time than required for BWRs and PWRs to discharge all fuel from the core. In spite of that, Fugen (165 MW, Japan) a pressure tube type reactor, was decontaminated after discharging the fuel to obtain a higher decontamination factor and to reduce radioactive waste. NPP Paks found a thermal-hydraulic anomaly in the reactor core caused by corrosion product deposits. Consequently, the coolant flow through the fuel assemblies was insufficient resulting in a temperature asymmetry in the reactor core. The fuel assemblies were removed from the core and successfully cleaned applying the CORD UV process. 8.1.2.3 Decontamination of sodium cooled systems These systems may be decontaminated effectively by acid solutions. Within the research and development programme of the CEC an inorganic acid-based process has been evaluated and tested by CEA for the RAPSODIE reactor in France [89Cos]. Decommissioning operations began in 1987. Preliminary cleaning and water rinsing after isolation of the main vessel eliminated all traces of residual sodium. Main contamination nuclides were 137Cs, 63Ni and 54Mn. After steam-cleaning to remove the residual sodium, the specimens were highly rusted. One of the first reagents to provide satisfactory results was a mixture of nitric acid and sulphuric acid at 85 °C. In order to improve the effectiveness of the decontamination, the aggressiveness of the reagent was Landolt-Börnstein New Series VIII/4
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enhanced by adding cerium (IV) in sulphate form. This reagent is sufficiently oxidising (Redox potential Eh = 1.610 V) iron, chromium and nickel in austenitic steels. Together with an alkaline washing excellent decontamination results were obtained: The estimated initial contamination level of 5500 Bq cm−2 was reduced to less than 10 Bq cm−2. The low residual contamination values allowed to estimate the pipes to be melted down for reutilization and release and an authorisation has been applied for. The dose rates were uniform throughout the facility, ranging from 1 to 15 µGyh−1 8.1.2.3 Gas cooled reactors (WAGR) An important feature of the Windscale AGR dismantling programme was the removal and disposal of the four heat exchangers [89Cro]. Each heat exchanger contains tube banks (or superheater banks), with plain Cr/Mo low alloy steel tubes and evaporator and economiser banks with finned mild steel tubes. Contamination was found to be incorporated in this oxide layer and to consist predominantly of 137Cs with some 134Cs and 60Co. Average values of 1.6 × 103 Bq cm−2 for Cs and 0.2 × 103 Bq cm−2 for Co were measured, giving a total for the superheater of about 2 ×1011 Bq. To remove the activity it was necessary to remove the oxide layer from the tubes. The decontamination was performed by spraying with a 3000-litre mixture of 0.5 molar (3.15 % by weight) nitric acid and 0.0025 molar citric acid at ambient temperature. In total the radiation level was removed by 70 %, activity were removed, the manSv-uptake was reduced remarkably and the targets reached.
8.1.3 Decontamination techniques for segmented parts 8.1.3.1 Chemical decontamination The chemical decontamination of an item removed from a nuclear plant or facility is generally carried out by immersion in a tank containing the chemical reagent [92Com]. The size of the tank depends on the dimensions of the item to be decontaminated. A common size is one, which is 1-2 m square with a depth of 0.5-1 m. Tanks for water rinsing are always installed. In sequential multistep processes the availability of several tanks can be useful in order to reduce the time needed. Chemical decontamination is characterised by the following parameters: • Type and nature of the chemical reagent; • Temperature of the process; • Duration of the process. The effectiveness of decontamination can be improved by increasing the duration of the treatment and the temperature. Optimum results are usually obtained with the solvents at elevated temperatures (up to 120 °C). During the decontamination process, as the concentration of the contaminants in the solution increases, the item being cleaned may become re-contaminated. This problem can be minimised by cleaning the least contaminated items first and by cleaning or replacing the solution if the concentration of contaminants exceeds certain levels. It should be noted that strong corrosive attack of the base metal may not result in high decontamination factors. These can however be achieved without significant corrosion of the base metal. Strong solutions of nitric and phosphoric acid used in the USA Bonus programme resulted in the removal of up to 0.2 mm of the inner wall of pipes, but only an average decontamination factor of approximately 100 was achieved. Results to date with non-aggressive processes indicate that decontamination factors as high as 2000 can be achieved without significant corrosion of the base metal [81Nel].
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Some multistep processes are commonly used for removing highly adhesive contamination layers. In many cases chemical decontamination can be used as a single step in complex processes e.g., before electropolishing, items covered with thick oxide layers are submitted to chemical decontamination in order to soften the oxide. 8.1.3.1.1 Chemical reagents Since the 1950s several chemical reagents have been commonly used for cleaning contaminated items. Lists with more than 100 chemical products can be read in dedicated decontamination handbooks [88Int, 81Nel, 82Com]. Some examples of chemical reagents used are given in Table 8.4. Table 8.4 Some reagents typically used in chemical decontamination of reactor components depending on type of material to be decontaminated. Materials Reagents Aluminium HNO3, Na-EDTA + 2 % detergent, 10 % citric acid, sulphamic acid …. Brass HNO3, 5 % AC Carbon steel Inhibited HCl, inhibited sulphamic acid, EDTA, citric + oxalic acids, APAC, CITROX, Copper Phosphorous, nitric and acetic acids Nickel & Alloys AP, 25 % HNO3 + 25 % HF, AP + AC, AP + CITROX, EDTA Monel 25 % sulphamic acid Stainless steel AP- CITROX, EDTA-CITROX, 30 % HF + 20 % HNO3, 70-80 % H3PO4, 0.4 M Cr SO4 + 0.5 M H2SO4, AP-HNO3, AP-AC, AP-OX, APACE Zircalloy OX + H2O2 + glucosic acid, EDTA, 8 M HNO3, … AP: alkaline permanganate, AC: ammonium citrate, OX: oxalic acid, CITROX: citric + oxalic acid, APACE: AP + AC + EDTA. There is a wider range of solvents to choose from for decommissioning programmes since corrosion of the base metal is of little concern. Certain solvents exhibit a time dependency in the mixing, heating, recirculation and draining cycle that affects both the chemical solution stability and the solubility of contained contamination. Each process under consideration would have to be evaluated for the effect of a loss-of-flow accident and associated cooling of the solvent. Factors considered would include toxic or explosive gas generation, excessive plate-out and excessive corrosion. The selected process must include appropriate emergency procedures, e.g. emergency draining, gas detection, and emergency ventilation. 8.1.3.1.2 Spent decontaminant solutions The selection of the chemical reagent directly affects the features of the secondary wastes arising from the process. It is obvious that continuously renewing the solution increases the decontamination effectiveness [85Pas] but the quantity of spent solution to treat and to dispose of also increase dramatically. In latter years the regeneration of chemicals have become a fundamental step in all chemical decontamination processes. Several conventional chemical processes can be used for regenerating the spent solutions either on their own or in combination and they include: ion exchange, evaporation/distillation or electrodialysis. The problem of limiting the secondary wastes arising from the decontamination process, sometimes leads to the selection of other similar processes like electropolishing or ultrasound with chemicals rather than chemical decontamination. As stated previously, only a detailed cost/benefit analysis can provide the actual criteria for selecting the best option for decontamination. Landolt-Börnstein New Series VIII/4
Ref. p. 8-34]
8 Decontamination
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8.1.3.2 Electrochemical decontamination 8.1.3.2.1 In-Tank Electrochemical decontamination can be considered in principle to be a chemical decontamination assisted by an electrical field. Nevertheless the best electropolishing is a process widely used in nonnuclear industrial applications to produce a smooth polished surface on a variety of metals and alloys. It can be considered the opposite of electroplating as metal layers are removed from a surface rather than added as a coating. Usually the object to be electropolished is immersed in a tank of electrolyte and is used as the anode in the electrolytic cell. The passage of electric current results in the anodic dissolution of surface material and, for normal operating conditions, a progressive smoothing of the surface. A progressive dissolution of the surface material occurs within a certain range of voltage and current density [83Int]. If the voltage and current densities are too low, the surface is attacked non-uniformly, causing etching rather than polishing. Similarly voltages that are too high cause severe pitting of the surface [87Pav].
(+)
DC Power supply
( -)
Anode
Cathode Hydrogen Oxygen
Containment Tank
Electrolyte
aa aaaa aa
aa a Fig. 8.4 Electropolishing device Surface to be decontaminated
If the anode is a contaminated material such as metal or alloy, all radioactive contamination on the surface (or entrapped within surface imperfections) can be removed and released into the electrolyte by this surface dissolution process [85Pas]. The process produces a very smooth (0.02-0.03 mm), non-reactive and non-adsorbing surface resistant to recontamination during further operations. Experience has shown that electropolishing is an effective technique for removing both fixed and smearable radionuclide contamination. Moreover it is fast and easily controlled. In general, there are two methods of application for electropolishing. The most common method is immersing the item to be decontaminated in a tank filled with a suitable electrolyte. The second method involves the use of “in-situ” mobile devices that are able to electropolish part of the surface of the item, which, because of size or installation, cannot be electropolished in a tank. Phosphoric acid is normally used as the electrolyte in electropolishing because of its stability, safety and applicability to a variety of alloy systems. Moreover, the non drying nature of phosphoric acid helps minimise airborne contamination, and the good complexing characteristics of phosphoric acid for metal ions is a significant factor in minimising recontamination from the electrolyte. Representative operating conditions for decontamination using phosphoric acid electrolytes are: solution temperatures of 5 to 25 °C, phosphoric acid concentrations of 40 to 85 %, electrode potentials of 8 to 12 V and current densities of 5 to 25 A cm−2. Landolt-Börnstein New Series VIII/4
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The direct current power supply converts alternating current in to direct current which generates the current flow between anode and cathode required for the electrochemical reaction. Voltage requirements range from 0 to 24 V with sufficient amperage to provide the required current densities. From experience gained in non-nuclear industrial applications, electropolishing normally uses phosphoric acid as the electrolyte (sulphuric acid is an alternative). However, during the last decade a variety of different electrolytes have been tested and developed with particular reference to the problem of reducing the secondary wastes arising from the process. As a consequence new processes usually called electrodecontamination or electropickling decontamination have been proposed. These consider the use of basic solutions as well as organic acid mixtures. Finally, among the electrodecontamination processes for “in situ” applications, electrobrushing using an electrobrush continuously fed by an electrolyte should be mentioned. Electrolytic decontamination can be used to remove fixed or imbedded contamination on iron-based alloys, including stainless steel, as well as on copper, aluminium, lead, and molybdenum. However the effectiveness of the decontamination can be limited by the presence of foreign materials on the surface of the items to be decontaminated. Materials such as oil, grease, oxide (rust), and paint or other coatings should be removed before decontamination. In decontaminating (mainly for decommissioning purposes) reactor coolant systems, the systems are usually covered by oxide layers that in principle work as a barrier for electropolishing. This problem can be overcome by increasing the electropolishing time. Nevertheless some new processes consider the periodic switching of polarity between cathode and anode, as well as changing voltage and current in order to increase the removal of the surface materials [86Gau]. Generally, at least two (stainless steel) tanks are required for performing electropolishing. One tank contains the electrolyte, electrodes, and parts to be decontaminated (as anode). The other tank holds the water used for rinsing the parts after decontamination. Power supply amperage capacities up to 2700 A are common. The cathode is normally a piece of copper, or stainless steel, positioned in the electrolyte within 30-100 mm from the item to be decontaminated. In addition for special items, the walls of the tank for immersion electropolishing can also serve as the cathode. To control vapours released from the electrolyte during the electropolishing process an extraction hood is located alongside the electropolishing tank. Provision for heating and agitating the electrolyte and rinse tank is also required. Studies on “in-tank” electropolishing became of relevance in the early 1970s in the USA where they were used to decontaminate hot-cells, glove-boxes, and tools contaminated by alpha emitters. Decontamination carried out in conjunction with Rockwell Hanford Operations and United Nuclear Industries in the USA, show that components heavily contaminated with PuO were decontaminated from 1 million dpm per 100 cm2 to background in less than 10 minutes [78All]. Typical decontamination times range from 5 to 30 min, corresponding to the removal of 10 to 50 mm of surface material at a current density of 2-15 A/dm−2. It is usually necessary to remove the anode contacts at least once during a cycle in order to decontaminate the area under the contacts. Since the early 1980s commercial use of electropolishing in decontamination of reactor coolant water systems and components for decommissioning purposes was made at the KRB power station (reactor A) in Germany [83Eic, 01Eic]. Electrochemical decontamination by electropolishing causes a steady increase of dissolved iron in the phosphoric acid. If the content of iron exceeds 100 g dm−3, a precipitation of iron phosphate occurs and this stops the efficiency of the decontamination process. Therefore the acid has to be exchanged or regenerated periodically. The regeneration of phosphoric acid is based upon the reaction of Fe2+ with oxalic acid (see Fig. 8.5). Electrochemical decontamination of steel, however, generates a high percentage of Fe in the phosphoric acid, which cannot be precipitated to iron oxalate. The high content of Fe3+ is reduced to Fe2+ by subsequent pickling. When a high portion of Fe2+ is obtained, the phosphoric acid has to be mixed with an aqueous solution of oxalic acid. The activity (mainly 60Co) is mostly separated from the solution by precipitation together with the iron. The iron oxalate is dried and stored for subsequent processing. The initial concentration of the phosphoric acid can be achieved by an evaporation process [89Sta].
Landolt-Börnstein New Series VIII/4
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8 Decontamination Oxalic Acid
Decontamination of 1 Mg Steel Material + -
8-17
Iron Oxalate
Precipitation Thermolysis
H 3 PO 4 Chemical, Electrochemical
Vaporization
Waste: 15 kg Iron Oxide
Fig. 8.5 Flow chart of the phosphoric acid regeneration
Several electrolytes were investigated and proposed as alternatives to phosphoric and sulphuric acid. The need for new electrolytes was initially motivated by the incompatibility of phosphoric and sulphuric acids with the existing treatment facilities and by the possibility of generating secondary liquid wastes which are more easier to process. This regeneration process was extensively proved at KRB Gundremmingen in Germany where more than 200 m3 of phosphoric acid (concentration 20-40 %) have been regenerated. The iron oxalate can be converted to iron oxide by pyrolytic decomposition. With this method, KRB-A-Reactor materials like pipes, pumps, and housings were decontaminated. Decontamination Factors of more than 100000 were reached. 8.1.3.2.2 “In-situ” In the case of “in-situ” electrochemical decontamination, the surface of the item to be decontaminated is flooded with electrolyte through a gap between the cathode of the device and the item's surface. The inside of tanks, housings and other relatively open vessels, without internal components, can be decontaminated without removal using an expandable “bladder” with a conductive surface that serves both as the cathode and as a displacement device to minimise the electrolyte volume. Several devices have been developed in the USA for application of “in-situ” electropolishing [82Gar1]: The internal cathode device consists of a perforated, tubular, copper or stainless steel cathode section with insulator-spacers at each end and has the provision for pumping the electrolyte and feeding power through the insulator at one end. The perforated tubular section permits flow of the electrolyte to the pipe surface being decontaminated, thus accomplishing the electropolishing action. An improved device with four module heads was designed and used in Germany for the Obrigheim power plant [84KWU]. The pump stream device consists of a perforated, disc-shaped, copper or stainless steel cathode facing the surface to be decontaminated, with an insulated handle for flow of electrolyte and supply of power. The electrolyte flows out of the end of the device in a stream and impinges on the surface being decontaminated. 8.1.3.2.3 Electrobrushing Electrobrushing is an electrodecontamination process for selected areas. The item to be decontaminated is used as the anode, while an electrobrush serves as the cathode. The brush itself is a cellulose sponge wetted by a continuous feed of an electrolyte, such as 5 % sulphuric acid solution inhibited with 1 g dm−3 ethyl quinolinlium. Decontamination is carried out by scrubbing at a current of 15 to 40 A at 15 to 20 V, and decontamination factors of around 30 are reported at a rate of 0.6 m2h−1 [81Nel].
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8 Decontamination
[Ref. p. 8-34
The disadvantages of this process include the production of large volumes of aqueous radioactive waste and excessive attack of the surface by the electrolyte. In addition, if the decontamination is performed manually rather than remotely, the radiation exposure to operators may be high. Single electropolishing and brushing processes can be combined. Subsequently, in decontaminating cast steel components at the KRB Gundremmingen power station in Germany, it was found that brushing prior to electropolishing resulted in a 35 % reduction of the time required for the galvanic decontamination process and the reduction of dissolved iron in the electrolyte was also found to be a benefit of prebrushing [89Sta]. 8.1.3.3 Jetting decontamination techniques The impingement of either a liquid or a solid media (or a liquid solid mixture called slurry) can be successfully used for decontamination. Extensive use of jetting methods to clean surfaces and items has been made in many industries and for varied applications. As a result of this several decontamination methods have been studied, developed and as a result of this several contaminated materials. Generally, jetting decontamination processes have a high flexibility and can therefore be applied to both large surfaces e.g. floors and walls, and relatively small-contaminated items and systems. Contaminated glove-box internals and several types of tools have also been cleaned using jetting processes. The problem of amount and characteristic of secondary wastes is one of the main concerns for jetting processes. The amount of waste can be strongly reduced by recirculating, treating and rinsing the impinging jetting media. Particular care must be applied to processes using abrasives. A variety of nozzles and lance configurations can be used for high-pressure water cleaning depending on the configuration of the item to be decontaminated. A straight jet can be used on the end of a long handled lance to reduce worker exposures for decontamination of accessible tank interiors, walls or floors. Self-propelling mole nozzles on a flexible high-pressure lance or hose can be used to decontaminate the inside of tubing or pipes. Because high pressure water cleaning is very effective for the removal of smearable surface contamination, water lances have been successfully used to decontaminate pump internals, valves, cavity walls, spent fuel pool racks, reactor vessel walls and heads, fuel handling equipment, feedwater spargers, floor drains, sumps, interior surfaces of pipes and storage tanks. Although decontamination factors of up to several hundred are commonly achieved, the normal factor for most applications is up to 50. Decontamination factors of 2 to 50 with water as the agent and of 40 with the addition of a proper cleaning agent were achieved at a particular site [79Rem]. 8.1.3.3.1 Abrasive jetting Abrasive jetting is a very effective decontamination method in which an abrasive medium is propelled by a jet of air (dry blasting) or water (wet blasting) against the surface to be cleaned [82Gar, 81Rem]. Typical abrasives are sand, alumina and metals, metal oxides and sawdust. Sand is the most common abrasive because it is inexpensive and a good scouring agent. Abrasive cleaning can be wet or dry. Abrasive particles are impelled at high velocity against the surface to be cleaned by air, water or a mixture of the two, as in the following air abrasive blasting, water abrasive blasting, air slurry blasting. Alternatively the particles may be carried in a viscous matrix and rubbed against the surface (liquid honing or abrasive slurry cleaning) or the abrasive may be in the form of stones, which can be used to grind or hone the surface to be cleaned. For effective cleaning either highpressure air at about 1 MPa or water at pressures similar to those used in hydrolaser systems are used, depending on the application. The abrasive can also be attached to a flexible backing to form a type of sandpaper or emery cloth or it can be forced against the surface by centrifugal action.
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8 Decontamination
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The different types of abrasive can be grouped into three general classifications depending on hardness: Hard abrasive materials: For decontamination, hard abrasive materials (harder than the material to be removed) are commonly used [64Ame]. In addition, there are three special sub-classifications: Cleanable abrasives, soluble abrasives and system compatible abrasives. The last one being important only for the decontamination of items required for reuse. A cleanable abrasive simplifies the problem of secondary wastes by markedly reducing the quantity of contaminated material to be disposed of after decontamination. One type of cleanable abrasive that can be used on non-stainless steel surfaces is steel shot. To facilitate cleaning the particles should be smooth; unfortunately smooth particles also reduce the abrasive action. A soluble abrasive is a material, which can be used in solid form in an air or water jet and can later be removed from the system by dissolving it in a liquid and flushing it out of the system. Boron oxide, which dissolves in water to form boric acid, is such a material. There are a number of methods of application of abrasive jetting, which depend on the carrier fluid, flow velocity and kind of abrasive used. Abrasive blasting is often carried out in special cabinets [91Bru]. A specific case of abrasive blasting is dry ice blasting which uses ice pellets (2-3 mm in diameter) produced by CO2 flashing at 40 °C. The use of dry ice pellets as an abrasive media strongly reduces the volume of secondary wastes produced [70Ayr]. The vacuum blasting method is a modification of air abrasive blasting for “in-situ” use where the discharge nozzle is surrounded by a concentric hood. An air exhaust line is attached to the hood and the blast air, debris and spent abrasive are drawn out through the exhaust line. The debris and abrasive are separated and the abrasive is reused. Smaller, hand-held units are also available. Water abrasive blasting has been used successfully to decontaminate a wide variety of contaminated components. Methods of remote application have been developed for “in-situ” cleaning to supplement the more common manual application methods. Decontamination factors of 200 to 300 are commonly achieved. 8.1.3.3.2 Freon jetting Systems have been developed to remove loose contamination from surfaces and equipment using commercial freon (trichlorotrifluoroethane) cleaning solvents [88Int]. Freon has a low viscosity and surface tension, which allows it to penetrate into cracks and crevices and remove contamination, including that associated with grease, oil, etc. Freon is non-flammable and chemically inert and can therefore be used to clean many types, of equipment without damaging delicate components. Most radioactive contaminants are insoluble in freon and can be removed by filtration or distillation, allowing recycling of the freon. Freon decontamination is carried out by directing a high pressure (15 MPa) jet of the liquid onto the surface to be cleaned. The decontamination is usually carried out inside a glove box [81McV], but experimental units have been developed for “in-situ” cleaning. The freon liquid and particles of contaminant are collected in a sump; the liquid is then filtered to remove the contaminants, cleaned and recycled. The freon is distilled as required to remove any radioactive material, which has dissolved in the liquid. The use of this decontamination technique is often limited due to legislative and regulatory restrictions in the industrial use of freon and freon-compounds due to their potential effect on the environment. 8.1.3.4 Ultrasonic decontamination Ultrasound consists of longitudinal mechanical waves and has been used over a long period of time for cleaning dirty surfaces in non-nuclear industry. In particular, ultrasonic cleaning has been used with good success for removing oil, grease, dirt and scale from a variety of items of various sizes and configurations. As a result of this ultrasound was used as one of the first methods for surface decontamination purposes in the nuclear industry [81McV]. Landolt-Börnstein New Series VIII/4
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8 Decontamination
[Ref. p. 8-34
The process is particularly appropriate for decontaminating items with complex surfaces where other decontamination techniques are not suitable. It has been successfully used to clean dirt from holes, cracks and crevices in parts made of metal, glass, and a variety of plastics. Its most common application is for decontamination of tools and items that are evenly contaminated and it is carried out by immersion in an ultrasonic tank containing water (or water with chemical additives) [82Gar]. New applications have been developed replacing water with more aggressive chemicals in order to combine the cleaning effect of both the chemical and ultrasound. 8.1.3.4.1 Ultrasonic technique The two main components of ultrasonic cleaning systems are the ultrasonic generator and the transducer (or vibrator). The ultrasonic generator converts normal 50-60 Hz power to a high frequency supply usually in the range of 18 to 25 kHz. The high frequency alternating current is then fed to a transducer to produce vibration in the liquid. The transducer is generally made of piezoelectric material (i.e. material that will elongate or contract depending upon the polarity). As a consequence of the vibration, compressive and rarefacted phases are present in the solution. In the rarefaction phase, cavities are generated (mainly originating from nucleation points) and during the compressive phase these collapse causing a phenomenon called “cavitation”. When an item is immersed in the solution, the collapse of cavities causes scrubbing on its surface and hence produces a cleaning action. It is important to note that the presence of many nucleation points causes cavitation mainly on the surface of the items. The cavitation occurs even if the surfaces are inhomogeneous and complicated or located in inaccessible zones. Calculations indicate that during cavitation localised peak pressures as high as 70 MPa can be reached. These conditions produce a strong cleaning action on any surface upon which they act [82Gar]. Two factors play a fundamental role in the action of ultrasound: the cavitation threshold and the scrubbing factor. The cavitation threshold is the pressure difference inside the fluid, which allows the cavitation phenomenon to take place, and is directly correlated with the ultrasonic power applied to the solution. 8.1.3.4.2 Ultrasound in conjunction with chemicals It is well known that in order to increase the effectiveness of the process in terms of scrubbing effects, or decontamination factors, an appropriate liquid should be selected. In ultrasonic cleaning, specific chemical agents are commonly added and the ultrasonic cleaning combines the effect of cavitation of a liquid at the surface to be cleaned with the chemical action of the liquid. The physical and chemical properties of the liquid are important. Decontamination factors as low as 2 have been obtained by using pure water at room temperature. The addition of a cleaning or wetting agent in the amount of 2 to 5 % by weight, with an increase in temperature to about 80 °C, can greatly increase the cleaning ability of water. The addition of small amounts of citric acid, or other chemicals, can also enhance the cleaning ability. Reports on the effectiveness of ultrasonic decontamination from the 40 or more nuclear plants where it has been used are mixed. Some plants use it on a regular basis with good success, obtaining decontamination factors in the range of 5 to 100 [80Man]. Other plants report little success, and some have stopped using ultrasonic cleaning entirely [82Gar]. KWU-Service used ultrasonic decontamination to clean primary recirc pumps at several power plants (Biblis A-B, Neckarwestheim, Unterweser, Borselle and Atucha). This equipment has also been used in more recent power plants. At the KWU Centre in Karlstein, Germany, a large amount of scaffolding and small tools were decontaminated up to the release limit [84KWU].
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8.1.3.4.3 Decontamination by mobile ultrasonic tools A specially designed ultrasonic hand-held wall cleaner and floor cleaner has been designed at the Argonne National Laboratory in the USA for decontamination of flame-sprayed zinc on hot cell liners [70Ayr]. 8.1.3.5 Decontamination by foams Foam decontamination uses liquid foam generated from an acid or acid mixture, using air, nitrogen or an inert gas. The foam also contains various chemical additives such as inhibitors, foam stabilisers and surfactants [82Com]. Typical acids used are hydrochloric, nitric, hydrofluoric, sulphuric, and phosphoric and organic acids can also be considered. The foam is produced in a foam generator and the density can vary considerably − typically 1 litre of acid solution is used to produce 20 dm3 of foam. Foams have been successfully used in decontamination. The use of foams started more than 40 years ago [60Ayr] and at that time, inhibited hydrochloric acid, together with special surfactants, was used. Decontamination factors in the range of 5 to 50 are obtained with a single foam application using 7-10 % phosphoric acid [82Gar]. 8.1.3.6 Decontamination by gels A gel medium is defined as a semisolid system obtained by flocculating and immobilising particles in a continuous medium. The problem in utilising this method of decontamination is the generation and maintenance of adequate gel systems in combination with decontaminating chemicals. For example, it has been found difficult to develop an adequate gel system using alkaline permanganate solutions [82Gar]. Gels can be made from either organic based or inorganic based systems and contain in the gel formulation decontaminating chemicals which are normally acids such as phosphoric, sulphuric, or nitric [80Des]. High decontamination factors (in excess of 50) have been obtained in the laboratory for mild steel, stainless steel, aluminium, copper and Plexiglas. Large decontamination tests by gel spraying decontamination have been performed at CEA-CEN-Cadarache, France, on 17 Mg of steel from the German ISAR-BWR (consisting of 11 Mg of frame and 6 Mg of pipes) [89Brun]. The chemicals used were sulphuric acid (2 mol dm−3) and hydrofluoric acid (1.6 mol dm−3) and during the decontamination 100 litres of gels were used. 8.1.3.7 Decontamination by pastes Cleaning pastes are widely used for treating metal surfaces, particularly for stainless steel, and can be extremely effective for decontamination. They consist of a filler, carrier and use an acid or mixture of acids as the active agent. The concept of decontamination pastes follows the method used for the older paste systems, however new acid systems particularly effective in removing radioactive contaminants have been developed. These pastes, when applied in a thin layer on contaminated surfaces, can provide effective decontamination, together with generating relatively small quantities of waste. 8.1.3.8 Mechanical decontamination techniques Mechanical techniques include many decontamination methods based on the use of mechanical tools or devices to remove the surface contamination [56USS]. Mechanical devices are commonly used for cleaning industrial tubing and piping and can be adapted for the decontamination of similar items in a nuclear installation. Landolt-Börnstein New Series VIII/4
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8 Decontamination
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8.1.3.9 Decontamination by strippable coatings Decontamination by strippable coatings consists of the application of a coating over the surface to be decontaminated. This coating is then left on the surface for a set period (from few hours to a few days) and then removed/stripped resulting in removal of the contamination. Strippable coating formulations usually consist of high molecular weight, film forming, synthetic polymers such as polyethylene, polyvinylacetate, polyvinylchloride, acrylics, etc., dispersed as an emulsion in an aqueous base. These coatings usually contain an active agent e.g., an acid or mixture of acids, which attack the contaminants on the surface to which the coatings are applied. The coatings may be applied with a brush, spray system, roller or other similar method. In some cases, it may be necessary to apply two or more coats to ensure that the coating has sufficient strength to be readily removed from the surface without tearing. The coatings are applied in varying thicknesses from 0.5 to 2 mm [79TMI]. Usually the coating is then manually stripped off the surface in sheets, compacted and placed in waste containers. 8.1.3.10 Melting Melting is considered as a decontamination process since it can be used to reduce the specific contamination. The method completely destroys components and is effective only for contaminants that are volatile or more soluble (e.g. plutonium) in the slag than the molten metal. The decontamination efficiency varies widely depending on the radioisotope present. The radionuclides remaining in the molten material are distributed homogeneously and effectively immobilised, thus reducing the possibility of the spread of contamination. The melting should take place in a suitable refinery, which has filters on the gas exhaust system to protect the environment [85Pfl]. Melting is extensively used in Germany. From 1984 to 1989 more than 2000 Mg of low-level contaminated scrap (<74 Bqg-1) have been melted and recycled [90Sap].
8.1.4 Decontamination techniques for building surfaces Decontamination processes to be used for contaminated concrete depend greatly on the characteristics of the concrete surface to be cleaned. They can vary from very simple hand based processes, to jackhammers or drilling removal techniques. The former is normally used for cleaning painted or smooth surfaces covered by loose contamination and the latter for decontaminating concrete in which the contamination has penetrated deeply. The following techniques a are in use: Brushing, washing and scrubbing: These are widely and frequently used at nuclear facilities to clean smooth surfaces, because they are simple and inexpensive. They are generally considered together, because they are related and in many decontamination works are used jointly or sequentially. Smearable contamination can be removed by wiping with a dry or damp cloth if the surface is smooth or impervious. To increase the effectiveness of decontamination detergents and solvents are added to the solutions especially if the loose contamination is associated with grease or oil. Abrasive powders or pads can be used if the contamination is associated with rust if it is embedded near the surface. Vacuum cleaning: Vacuum cleaning is one of the most widely and frequently used decontamination processes to clean smooth concrete surfaces. It is also used to collect dust resulting from brushing decontamination processes such as scarifying, spalling, etc. The process is very Simple and can be efficiently used for loose particles on both wet and dry surfaces. Scarifying and grinding techniques: Scarifying and grinding processes [80Bar] have been widely used for a long time for the decontamination of concrete walls and floors of different nuclear facilities. They are particularly appropriate for the removal of thin concrete contamination layers (typically less than 10 mm).
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Thermal scarifying: Thermal scarifying processes are based on the removal of concrete layer by thermal energy introduced by flame (or plasma) impingement on the surface to be decontaminated. In principle two different thermal scarifying phenomena may take part in the process: spattering and melting [85Ebe]. Both spattering and melting use almost all the thermal energy of the flame, so that no heat penetrates the material. The energy remains in the spattered and molten material, which becomes separated from the concrete. After flame scarifying, the loose particles and remnants of combustion must be removed from the treated surface. Circular brushes or cleaning machines, which can be fitted with steel wire brush rollers or beater rollers, are recommended for this task. Flame scarifying has long been used to treat surfaces in order to produce a clean, dry base for paint and other surface coatings. Spalling: When a floor or wall is deeply contaminated, a thick concrete layer must be removed in order to decontaminate it completely. Removal of the surface radioactivity in this manner, in comparison with demolishing the entire structure, eliminates the need to dispose of large quantities of non-radioactive concrete, which may arise with other volume removal techniques. To remove these thick concrete layers hard mechanical processes should be used. Surface breakers, pneumatically or hydraulic operated drilled bits, and water cannons are typically used in spalling processes.
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8.2 Decontamination of skin 8.2.1 Introduction Radiation protection rules in national legislation generally include action levels or limits for protection measures in case of surface contamination at the workplace and of objects (see Section 8.1). Frequently limits for decontamination measures of skin are not included. However, radiation protection in practice requires at least reference values to avoid unsuitable or detrimental decontamination measures which may lead to skin lesions or increased incorporations. Based on new calculations of equivalent dose rates for the skin [85Hen], recommendations of reference values have been developed, among others, by the German Commission on Radiological Protection [92SSK] which serve as a basis for the following specifications. The described measures in case of skin contamination are generally based on the following principle: In the event of contamination, the resulting radiation dose must be kept as low as reasonably achievable, economic and social factors being taken into account, and considering individual situations also below the dose limits for the skin, in accordance with the recommendations of the International Commission on Radiological Protection ICRP [91ICR] (see also Sect. 4.8). Adherence to this principle requires in working areas, where contamination cannot be excluded, suitable measures to keep the skin dose after contamination as low as possible. The following specifications serve this purpose. They are no recommendations for measures in emergency and disaster situations and do not include medical treatment of contaminated wounds. They are rather addressed to technical, medical and scientific installations where sealed and unsealed radioactive sources are handled, in accordance with the respective radiation protection rules of their national legislation. The following areas are mainly involved: • • • •
Nuclear power plants Nuclear fuel cycle installations Scientific and industrial laboratories Hospitals, medical laboratories and practices of doctors in nuclear medicine.
8.2.2 Transport of radioactive substances via the skin 8.2.2.1 Anatomy of the skin The skin (cutis) consists of epidermis and subcutis. The epidermis is the avascular external skin layer. The subcutis consists of a tissue layer with connective tissue septa in which fat cell clusters and nerves are located. In case of skin contamination the epidermis is primarily concerned. Due to permeation of radioactive substances, radioactivity may enter the transfer compartment (see Chapter 7) via the subcutis and thus lead to internal contamination of the organism. The epidermis consists of multi-layer corneal squamous-cell epithelium of a thickness between 30 µm and 2 mm, depending on the body region: The external layer of the body skin, i.e. the upper layer of the epidermis, is the stratum corneum, on the surface of which flat, denucleated corneal cells peel off in fine scales. It is followed by the stratum lucidum consisting of denucleated cells which is only produced at thick parts of the epidermis - palm and sole. It is followed by the stratum granulosum, the so-called granular cell layer. Then follows the stratum spinosum, the so-called prickle cell layer and the stratum basale, the basal-cell layer. Stratum spinosum and stratum basale are also called stratum germinativum (germ cell layer), because here the corneal cells scaled off at the surface of the epidermis are substituted by cell division. Therefore, the cells of the stratum germinativum are the radiation-sensitive cells of the epidermis. Their radiation dose (equivalent dose) has to be determined in case of skin contamination, their dose level determines the introduction and conclusion of decontamination measures. Landolt-Börnstein New Series VIII/4
Ref. p. 8-34]
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8.2.2.2 Transport procedure The detriment to health due to skin contamination from radioactive substances depends mainly on • • • •
Type of the radionuclide and its chemical compound Activity per area and/or specific activity Solubility of radioactive substance Situation and size of possibly affected part of skin.
In principle the healthy skin is protected best against percutaneous incorporation of radioactive substances. Therefore, all persons handling unsealed radioactive sources should give special care to their skin, particularly at their hands, and keep it in a good and healthy condition, because fissured skin surface may become openings for entry of radioactive substances and consequently for incorporations of radionuclides. The intact skin is an effective but not completely dense barrier against radioactive substances. Whereas solid particles, preferred by rubbing, may enter the skin mechanically, liquids are subject to capillary forces and diffusion processes. As long as a liquid wets the skin surface, a transport of material occurs from the surface into the skin and through the skin. This leads to both a transient deposit of radioactivity in the corneal layer and transport into deeper skin layers (subcutis) together with uptake into blood and subsequent internal radiation exposure (see Chapter 7). The capacity of the corneal skin layer to take up radioactive liquid is exhausted within several minutes. This layer takes up about one micro litre liquid per square centimeter wet surface [92Pra]. Activity from this procedure taken up into the corneal layer is therefore proportional to the specific activity of the liquid and to the size of the affected skin surface. However, the corneal layer may have a special affinity to some substances. In case of a permanent contamination this may lead to radioactive enrichment in the corneal layer. The transport through the skin (permeation) from a liquid on the surface is proportional to the specific activity in the liquid and to the size of the affected skin surface and additionally to the time period during which the contaminated liquid remains on the skin. In case of inorganic substances dissolved in water, organic acids, salts, or lipophilic compounds, the substance transported through the skin per square centimeter and hour is equal to the amount contained in 0.001 to 0.1 µl [92SSK]. The permeation rate is significantly higher with gases dissolved in water or easy volatile substances.
8.2.3 Skin dose at contamination 8.2.3.1 Calculation of the equivalent dose to the skin The equivalent dose of the contaminated skin can be calculated as follows [85Hen]: ln 2 − ⋅t ⎞ 86400 ⎛⎜ T1/2 ⎟ & H S = AF ⋅ T1/ 2 ⋅ ⋅h = ⋅ 1− e ⎟ s ln 2 ⎜ ⎠ ⎝
(8.2.1) ln 2 ⎛ − ⋅t ⎞ T1/2 ⎟ & 5 ⎜ H S = 1.25 ⋅ 10 ⋅ AF ⋅ T1/ 2 ⋅ 1 − e ⋅h ⎜ ⎟ s ⎝ ⎠ with HS equivalent dose of contaminated skin in Sv AF activity per area at the surface in Bq/cm2 physical half-life in days (d) T1/2 t time of contamination in days (d) h&S equivalent dose rate conversion coefficient in Sv/s per Bq/cm2 (see Section 8.2.3.2) Landolt-Börnstein New Series VIII/4
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[Ref. p. 8-34
If a dwell time of one week is assumed for contamination, equation (8.2.1) simplifies for long-lived radionuclides, i.e. for radionuclides with significantly longer half-life than 7 days, as follows: HS = 604800 · AF · h&S
(8.2.2)
8.2.3.2 Equivalent dose rate conversion coefficients
The equivalent dose rate conversion coefficient for skin contamination is defined as follows (see equation 8.2.1): The numerical value of the equivalent dose rate conversion coefficient corresponds to the equivalent dose rate in Sv/s in the contaminated skin at an activity per area of 1 Bq/cm2. For determining these values [85Hen] the mean skin dose was calculated by integration over a skin depth between 50 and 100 µm, and the contribution of gamma-, beta-, electron radiation (Auger electrons) and alpha-particles was considered. Contamination of the whole skin surface was assumed to calculate the contribution of gamma radiation to skin dose. The assumption that the radiation-sensitive layer (stratum germinativum, Sect. 8.2.2.1) is situated mainly in a skin depth of 50 to 100 µm leads to somewhat higher equivalent dose rate conversion coefficients than those obtained for the reference skin depth of 70 µm [91ICR]. Radioactive substances may enter the corneal layer, however, the activity concentration decreases significantly with depth of the corneal layer (exponentially with a half-value thickness of about 2 µm). As a consequence this permeation has only little influence to dose in the radiation-sensitive skin layer (stratum germinativum). Therefore the ambient activity distribution in the corneal layer was not considered in the calculation of the equivalent dose rate conversion coefficents. The equivalent dose rate conversion coefficient values for 128 radionuclides are summarised in Table 8.5. Table 8.5 Equivalent dose rate conversion coefficients (Sv/s/(Bq/cm2)) for contaminated skin (averaged over a depth of 50-100 µm) [85Hen]
Radionuclide
Radiation Electrons/BetaParticles
Gamma-Radiation
Alpha-Particles
Na-24
4.3E-10
6.6E-11
-
Cr-51
7.1E-12
1.4E-12
-
Mn-54 Mn-56
2.1E-12 4.2E-10
1.6E-11 3.0E-11
-
Fe-55 Fe-59
2.7E-10
1.2E-12 2.2E-11
-
Co-56 Co-58 Co-60
8.8E-11 7.0E-11 2.4E-10
6.2E-11 1.9E-11 4.5E-11
-
Ni-59 Ni-65
4.1E-10
1.6E-12 9.7E-12
-
Landolt-Börnstein New Series VIII/4
Ref. p. 8-34]
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Radionuclide
Radiation Gamma-Radiation
Alpha-Particles
Zn-65 Zn-69 Zn-69m
Electrons/BetaParticles 7.5E-12 4.3E-10 2.4E-11
1.3E-11 1.1E-16 7.7E-12
-
Se-75
3.0E-11
9.3E-12
-
Rb-86 Rb-88 Rb-89
4.2E-10 1.7E-10 3.9E-10
1.8E-12 1.1E-11 3.6E-11
-
Sr-89 Sr-90 Sr-91 Sr-92
4.2E-10 3.9E-10 4.2E-10 3.7E-10
1.6E-15 1.3E-11 2.4E-11
-
Y-90 Y-90m Y-91 Y-91m Y-92 Y-93
4.2E-10 5.9E-11 4.2E-10 2.4E-11 3.8E-10 4.1E-10
4.3E-16 1.1E-11 6.6E-14 1.0E-11 4.6E-12 1.5E-12
-
Zr-93 Zr-95 Zr-97
1.8E-13 2.8E-10 4.2E-10
1.4E-11 3.3E-12
-
Nb-93m Nb-95 Nb-95m Nb-97
4.0E-09 4.5E-10 4.2E-10
4.7E-13 1.4E-11 2.8E-12 1.2E-10
-
Mo-93 Mo-99 Mo-101
4.1E-10 4.6E-10
2.5E-12 2.9E-11 2.4E-11
-
Tc-99 Tc-99m Tc-101
2.6E-10 5.1E-11 4.3E-10
2.2E-12 5.8E-12
-
Ru-103 Ru-105 Ru-106
1.5E-10 4.2E-10 -
8.8E-12 1.5E-11 -
-
Rh-103m Rh-105 Rh-106
3.2E-10 3.9E-10
3.2E-13 1.4E-12 3.8E-12
-
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[Ref. p. 8-34
Radionuclide
Radiation Gamma-Radiation
Alpha-Particles
Ag-110m Ag-111
Electrons/BetaParticles 1.2E-10 4.2E-10
5.1E-11 4.6E-13
-
Sb-124 Sb-125 Sb-126 Sb-127 Sb-129 Sb-130 Sb-131
3.7E-10 1.8E-10 3.5E-10 4.1E-10 3.9E-10 4.9E-10 4.1E-10
3.2E-11 8.8E-12 5.3E-11 1.3E-11 2.6E-11 6.0E-11 3.4E-11
-
Te-125m Te-127 Te-127m Te-129 Te-129m Te-131 Te-131m Te-132 Te-133 Te-133m Te-134
2.8E-10 3.9 E-10 1.5E-10 4.4E-10 3.1E-10 4.9E-10 3.9E-10 2.1E-10 4.1E-10 4.4E-10 4.6E-10
3.1E-12 8.9E-14 1.0E-12 1.5E-12 1.3E-12 7.6E-12 2.6E-11 5.0E-12 1.7E-11 4.2E-11 1.7E-11
-
I-129 I-130 I-131 I-132 I-133 I-134 I-135
9.2E-11 4.1E-10 3.7E-10 4.2E-10 4.2E-10 4.2E-10 4.0E-10
1.7E-12 4.0E-11 6.9E-12 4.2E-11 1.1E-11 4.8E-11 2.8E-11
-
Cs-134 Cs-134m Cs-135 Cs-136 Cs-137 Cs-138
2.8E-10 3.3E-10 1.5E-10 3.3E-10 3.7E-10 4.1E-10
2.9E-11 1.4E-12 4.0E-11 4.0E-11
-
Ba-137m Ba-139 Ba-140
4.5E-11 4.4E-10 4.1E-10
1.1E-11 7.4E-13 3.9E-12
-
La-140 La-141 La-142
4.3E-10 4.2E-10 4.1E-10
4.0E-11 7.5E-13 4.4E-11
-
Landolt-Börnstein New Series VIII/4
Ref. p. 8-34]
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Radionuclide
Radiation Gamma-Radiation
Alpha-Particles
Ce-141 Ce-143 Ce-144
Electrons/BetaParticles 4.1E-10 4.3E-10 2.1E-10
1.4E-12 5.7E-11 5.1E-13
-
Pr-143 Pr-144m Pr-145
6.6E-13 1.1E-11 4.2E-10
6.7E-12 9.2E-13 2.5E-13
-
Nd-147
4.0E-10
3.2E-12
-
Pm-147 Pm-148 Pm-148m Pm-149 Pm-151
1.3E-10 4.2E-10 3.3E-10 4.2E-10 4.1E-10
1.0E-16 1.0E-11 3.7E-11 1.9E-13 5.7E-12
-
Eu-152 Eu-152m Eu-154 Eu-155 Eu-156
2.0E-10 3.3E-10 4.5E-10 8.4E-11 3.7E-10
2.1E-11 5.7E-12 2.3E-11 1.4E-12 2.3E-11
-
At-211
1.4E-11
1.6E-12
-
Ra-226
1.1E-11
1.4E-13
-
U-234 U-235 U-238
8.7E-12 6.1E-11 -
4.6E-13 3.5E-12 3.8E-13
-
Np-237 Np-238 Np-239
6.2E-11 2.9E-10 6.2E-10
3.1E-12 1.2E-11 4.8E-12
-
Pu-236 Pu-238 Pu-239 Pu-240 Pu-241 Pu-242
1.0E-11 8.0E-12 1.9E-13 8.4E-12 1.0E-09 7.0E-12
5.3E-13 4.6E-13 2.5E-13 4.4E-13 2.8E-16 3.7E-13
6.9E-20 5.5E-20 1.2E-20 1.3E-20 -
Am-241 Am-242 Am-242m Am-243
4.7E-11 3.1E-10 2.6E-13 2.1E-11
3.1E-12 1.4E-12 1.0E-12 1.7E-12
5.7E-20 8.1E-20 2.1E-20
Landolt-Börnstein New Series VIII/4
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Radionuclide
Radiation
Cm-242 Cm-243 Cm-244 Cm-245 Cm-246 Cm-247 Cm-248
Electrons/BetaParticles 4.7E-12 3.3E-10 1.2E-10 3.5E-11 -
[Ref. p. 8-34
Gamma-Radiation
Alpha-Particles
4.3E-13 4.2E-12 4.0E-13 3.7E-12 3.5E-13 6.0E-12 2.7E-13
7.7E-15 3.0E-16 8.6E-20 3.0E-20 3.0E-20 4.0E-21 5.4E-21
8.2.4 Decontamination measures 8.2.4.1 Organisational and preliminary measures
When handling radioactive substances contamination should always be anticipated. Therefore, organisational measures and practical procedures for personal decontamination have to be provided. There is need for developing special instructions of decontamination for the respective operation and also for individual workplaces. These measures include, among others, to take off contaminated clothing before decontamination measures are started. Care shall be taken that no additional parts of skin will be polluted and no contaminated dust will be emitted into the air. The emergency staff should wear protective gloves or protective clothing. 8.2.4.2 First aid measures of skin decontamination
Simple decontamination appliances which can be used immediately after contamination and at any place should be available for decontamination. If need be, immediate decontamination measures will be given priority over assessing the value of skin contamination by activity measurement. Based on the experience that both specific activity concentration of a contaminated liquid and the time of influence or action may be the decisive parameters for skin permeation, skin decontamination should be started immediately after contamination, if possible. However, it should always be considered that incorporation of radionuclides due to permeation might be effectively reduced by simple and quickly performed washing measures. In general washing with lukewarm water, special soaps or wash lotions using soft hand-brushes, if required, are first and rapid decontamination methods. Also secondary contamination of the surrounding skin from washing procedures with lukewarm water is widely negligible in practice, because if enough water is immediately used, the radionuclide concentration is significantly decreased and the time of influence is short. Minor contaminations can usually be removed already in a first washing course. During this stage only the contaminated skin parts should possibly be cleaned with lukewarm water, e.g. only the palm of the hand. Washing should be finished after 2 minutes and the skin should be dried with absorbent material. In case of remaining contamination the procedure according to Section 8.2.5 should be applied.
Landolt-Börnstein New Series VIII/4
Ref. p. 8-34]
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8.2.4.3 Specific decontamination procedures Decontamination by removing contaminated corneal cells
Small-surface contamination can be removed by taking off corneal cells with adhesive film: after 5 strippings 97 % of the radioactive substance was removed from the skin at the forearm [58Bor]. However, this stripping method is only suitable with dry skin and fails at the palms of the hand. Cleaning by sorption agents
To remove substances penetrated deeper into the corneal layer, cleaning methods are required where the procedure of penetration shall be inverse, i.e. from internal to external penetration. In order to avoid transfer of radioactive substances into the corneal layer from capillary forces, it is suitable to apply for the decontamination sucking sorption agents, e.g. silicon dioxide, titanium dioxide or silica gel. Cleaning by detergents
Although cleaning with customary detergents is a natural decontamination method, it should be considered that without penetration of a decontamination agent into the corneal layer, the substance deposited there cannot be reached. Therefore when cleaning a substance situated in the corneal layer it shall be dissolved by the detergent and rinsed to the skin surface. Therefore the corneal layer must not be decontaminated by alkaline or strong acidic cleaning agents, because the bond capacity of keratin to ions increases with each deviation from the pH value 4.2, as it has been proved by experimental results with 22 Na and 131I ions in corneocytes [71EI, 84Pra]. Specific decontamination
Specific decontamination methods have been described by Wijker [66Wij]. However, they should only be performed by specially trained experts. Warning is particularly issued about chemicals for nuclidespecific decontamination, if there is no knowledge about the chemical composition of radionuclides involved. Decontamination agents for skin and hair
For the decontamination of skin and hair the following decontamination solutions are normally used: Decontamination lotions for skin and hair • • • •
Titanium oxide paste (general skin decontamination procedure) Wiping paste (general skin decontamination procedure) Citric acid 3 % (e.g. decontamination of hair and external auditory canal) Complexing solutions (e.g. in case of contamination of the eyes, general skin decontamination procedure) [97Ger] • Potassium permanganate solution (general skin decontamination procedure), removing brown skin colouring with sodium disulfite solution • Physiological sodium chloride solution (decontamination measure also in case of contamination of the eyes).
Landolt-Börnstein New Series VIII/4
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[Ref. p. 8-34
8.2.4.4 Decontamination of specific body regions and organs Hair
Contaminated hair should be washed with a wash lotion (see Section 8.2.4.3), assisted by a helper wearing protective gloves, in an adequate hair washbasin with the head bent backwards. Then the hair should be rinsed with plenty of water. Special care must be taken that no contaminated water runs into the face, eyes or ears. Before hair drying, control measurement by contamination monitor is required. Eyes
In case of contamination of the eyes, these should be properly rinsed with plenty of water; cleaning lotions (physiological sodium chloride solutions and integration solutions (see Section 8.2.4.3) should be used, if necessary. This method of eye decontamination must be performed under medical supervision. Mouth, nose, ears
If mouth, nasopharynx and auditory canal are contaminated, a physician must always be contacted. The mouth should be rinsed with plenty of water for decontamination. Contamination of the nasal cavities can be decreased by blowing the nose. Rinsing (wash bottle) with physiological sodium chloride solution or citric acid must only be done on the instructions or assistance of a physician, if possible, because there is hazard of radionuclide incorporation. In any case secondary contamination has to be avoided. Skinfolds, groove of the nail bed, and fingernails
If contamination is detected in skinfolds, in the groove of the nail bed or under the fingernails, this should be specifically removed. Simple instruments can be used such as nail cleaner, soft brush or adhesive strips.
8.2.5 Procedure at residual contamination and fixing a reference value 8.2.5.1 Frequency of decontamination steps
If the first decontamination procedure (see Sections 8.2.4.2, 8.2.4.3 and 8.2.4.4) is not successful, the decontamination method can be repeated up to two times while measuring each individual decontamination effect. If the decontamination effect is lower than 10 %, and the surface-related residual activity is lower than the reference value of 10 Bq/cm2 − averaged over 100 cm2 with contamination predominantly dispersed over the whole surface (see Section 8.2.5.2) − the additional decontamination step can be waived. As far as the decontamination effect is higher than 10 % and the skin condition is good, further wash procedures may be reasonable. 8.2.5.2 Derivation of the reference value for residual contamination
The reference value of 10 Bq/cm2 of a remaining skin contamination after several decontamination steps leads to the fact that for more than 90 % of the radionuclides listed in Table 8.5 which are essential for practical radiological protection purposes the remainder equivalent dose is significantly less than 1 % of Landolt-Börnstein New Series VIII/4
Ref. p. 8-34]
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the annual dose limit to skin of 500 mSv [91ICR] in a 1-week-dwell time. All dose values lie below 5 % of this limit, except for 254Cf. This stipulation ensures that even in case of several contamination events in the calendar year which were not successfully removed, the distance to the limit value is sufficient. When fixing a reference value to decide on further decontamination measures, it can also be assumed that activity concentration decreases exponentially with depth in the corneal layer. The complete scaling off of this layer within two weeks leads to a very quick exponential decrease with time of the residual activity in the skin. Consequently the 1-week-dwell time taken as a basis for dose calculations (see equation 8.2.1 in Sect. 8.2.3.1) overestimates significantly the actual equivalent dose. Table 8.6 gives an example of the activity per area values in some radionuclides relevant for radiological protection purposes. Considering the physical half-life at 1-week-dwell time they lead to a dose of about 1 % of the annual dose limit to skin of ICRP of 500 mSv [91ICR]. Table 8.6 Activity per area of some radionuclides leading after 1-week-dwell time to a skin equivalent dose of about 5 mSv (1 % of the annual dose limit for skin of 500 mSv [91ICR]). Radionuclide Activity per area [Bq/cm2] 14 C 170 60
Co
35
90
Sr
20
90
Y
40
131
I
35
137
Cs
20
141
Ce
20
The reference value of 10 Bq/cm2 is adequately conservative and can also be used for radionuclides which, due to the short range of radiation emitted (mainly Auger electrons), provide a main contribution to dose in the skin layer sensitive to radiation (stratum germinativum). Consequently, precaution against the hazard of transmitting radioactive substances from the restricted access area is ensured, whereby it should be taken into account that remaining residual contamination is a very rare event and that remaining activity clings tightly to the skin. Apart from exposure to skin, radionuclides situated in the corneal layer may principally lead to exposure in other body regions: • Radiation – mainly gamma radiation – may expose other body organs or tissues • Radioactive substances may reach body liquids by permeation and thus disperse in the body with final irradiation of organs or tissues. Usually, external exposure to other body regions by gamma radiation resulting from the corneal layer compared to skin exposure is negligible. In case of very high surface contaminations, however, very large affected body surfaces and long contamination periods, the permeation of radionuclides through the skin and hence internal dose to body organs and tissues can play a role that should not be underestimated.
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8 Decontamination
8.3 References 56USS 58Bor 60Ayr 64Ame 66Wij 70Ayr 71El 78All 78Rie 79Rem 79TMI 80Bar 80Des
80Man 81McV 81Nel 81Rem 82Com 82Gar 82Gar1 83Eic 83Int 84Ebe 84KWU
Fabrication of USS Stainless Steels, 2nd ed., p 88, Bulletin published by United States Steel Corporation, Pittsburgh, Pa, 1956. W. Born: Beseitigung radioaktiver Verunreinigungen von der Haut des Menschen; Strahlentherapie 106, 435 (1958) Ayres, J. A., Demmitt, T.F., Larrick, A.P., Neubow, G.E., Richman, R.B., Perrigo, L.D., Weed, R.D.: Decontamination studies for HAPO water-cooled reactor systems; USAEC HW-67937. Dec. 27, 1960. American Society for Metals: Abrasive blast cleaning. Metals Handbook, 8 ed. Vol. 2, 1964. Skin Contamination; Euratombericht 41-67 (1966) Ayres, J.A. (ed.): Decontamination of nuclear reactors and equipmentNew York: . The Ronald Press Co., 1970. R. El-Julani: Zur Adsorption von Na+ und I an Keratinzellen der menschlichen Haut; Med. Dissertation, Ludwig-Maximilians-Universität München (1971) Allen, R.P., Arrowsmith, H.W, Charlot, L.A., Hooper, J.L.: Electropolishing as a decontamination process: Progress and applications; PNL-SA-6858, April, 1978. Riess, R., Bertholdt, H.: Chemische Dekontamination von Reaktoranlagen, Reaktortagung 1978, 4.-7. April in Hannover, Tagungsbericht, S. 963-966. Remark, J.F., Miller, A.D.: Review of plant decontamination methods, Sun Valley (ID), USA: American Nuclear Society, September 17-19, 1979. TMI Reports: Evaluation of strippable decon coatings; TMIReports No.'s 1-22 (1979). Barbier, M.M., Chester, C.V.: Decontamination of large horizontal concrete surfaces outdoors; CONF-800542-2, ORNL, TN (USA), 1980. Desryoches, J., Koenig, J., Lebrun, J.C.: La décontamination en milieu gélifie ou colloidal; Workshop on waste washing; Organised by Radioactive Waste Management Committee and OCED Nuclear Energy Agency at Centre d'Etudes Nucléaires de Cadarache, France. November 19-21, 1980. Manion, W.J., Laguardia, T.S.: Decommissioning Handbook; DOE/EV/10128-1, November 1980. McVey, J.T., et al.: Tools and equipment: From nuclear waste to reusable items, Nucl. Chem. Waste Manag. 2-3 (1981). Nelson, J.L., Divine, J.R.: Decontamination processes for restorative operations and as a precursor to decommissioning: A literature review. PNL-3706, Battelle-Pacific Northwest Laboratory, May 1981. Remark, J.F.: Plant decontamination methods review; EPRI NP-1168; May 1981. Commissariat à l’Energie Atomique, Institut de Protection et de Sûreté Nucléaire: Décontamination radioactive du matériel', Publication PMDS, Mars 1982. Gardner, H.R., Allen, R.P., Polenz, L.M., Skiens, W.E., Wolf, G. A.: Evaluation of nonchemical decontamination techniques for use on reactor coolant systems; EPRI NP-2690, October 1982. Gardner, H.R., et al.: Comparison of decontamination techniques for reactor coolant system applications; EPRI NP-2777, December 1982. Eickelpasch, N., Lasch, M.: Electrochemical decontamination experience at Gundremmingen power plant; Water chemistry of nuclear reactor, Systems 3, Bournemouth (UK), 17-21 October, 1983. International Atomic Energy Agency: Decommissioning of nuclear facilities: Decontamination, disassembly and waste management; Vienna: Technical Report Series No. 230, 1983. Ebeling, W., Boedeker, B., Rose, K.: Dekontamination von Betonoberflächen durch Flammstrahlen; EUR 8969, 1984. KWU Service Report, No. 1, April 1984. Landolt-Börnstein New Series VIII/4
8 Decontamination 84Pra 85Duc 85Ebe 85EPR 85EPR1 85EPR2 85EPR3 85Hen 85Lör 85Pas 85Pfl 86EPR 86EPR1 86Gau 86Man 87Pav 88Int 89Brun 89Cos 89Cro 89Sta 89Wil 90Ber 90Sap
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H. G. Pratzel, K. Dirnagel, H. Drexel: Kontamination der menschlichen Haut durch Radionuklide; Nuklearmedizin 23, 197-200 (1984) Duce, S.W., Simpson, F.B., Mandler, J.W.: Observations of plant decons; EPRI Seminar on Chemical Decontamination of BWRs, Charlotte (USA), February 26-28, 1985. Ebeling, W., Boedeker, B., Rose, K., Schaller, R.H.: Decontamination of concrete, with particular reference to flame scarifying; Decommissioning of nuclear power plants; Luxembourg 22-24 May 1984; EUR 9474, 1985. EPRI: EPRI NP-4222, Vol.2, 1985. EPRI: EPRI NP-4222, Vol.3, 1985. EPRI: EPRI NP-4222, Vol.4, 1985. EPRI: EPRI NP-4222, Vol.5, 1985. K. Henrichs, C. Eiberweiser, H.G. Paretzke: Dosisfaktoren für die Kontamination der Haut und der Kleidung; GSF-Bericht 7/85, 5-1285 (1985) Lörcher, G., Chapuis, A.M., Essmann, J.: Factors to be considered in deciding whether to decontaminate for unrestricted release; Decommissioning of nuclear power plants, Luxembourg. 22-24 May 1984, EUR 9474, 1985. Pascali, R., Bregani, F., Ahlfänger, W., Lasch, M., Gauchon, J.P.: Chemical and electrochemical decontamination; Decommissioning of nuclear power plants; Luxembourg, 22-24 May 1984, EUR 9474, 1985. Pflugrad, K., et al.: Treatment of steel waste arising from decommissioning of nuclear installations by melting; Bethesda, (MD) USA: Nuclear Reactor Decommissioning Planning, 1985. EPRI: EPRI NP-4687, 1986. EPRI: EPRI NP-468, 1986. Gauchon, J.P., Mordenti, P., Bezia, C., Fuentes, P., Kervegant, Y., Munoz, C., Pierlas, C.: Decontamination par des methodes chimiques, electrochimiques et au jet d'eau; EUR 10043, 1986. Mang, M.: Issue of the second seminar on chemical decontamination, BWRs-Section 3, Corrosion Issues, 1986. Pavlik, O., Sipos, T., Vicsevne, M., Miko’, M.: Decontamination of Nuclear Facilities by Electrochemical methods; 1987 International Decommissioning Symposium, Pittsburgh, (PA) USA, October 4-8, 1987. International Atomic Energy Agency: Decontamination and demolition of concrete in the decommissioning of nuclear facilities; Vienna: Technical Reports Series No. 286, 1988. Brunel, G.: Decontamination using chemical gels, electrolytical SWAB, abrasives; Decommissioning of Nuclear Installations, Bruxelles (B), October 24-27, 1989, EUR 12690, 1990. Costes, J.R., Antoine, P., Gauchon, J.P.: Decontamination before dismantling a fast breeder reactor primary cooling system; Decommissioning of Nuclear Installations. Elsevier Science Publishers Ltd, EUR 12690, 1989, ISBN 1-85166-523-4, p. 554 Crossley, H., Wakefield, J.R.: Development of techniques to decontaminate the WAGR Heat Exchangers, UK, Windscale, Decommissioning of Nuclear Installations. Elsevier Science Publishers ltd, EUR 12690, 1989, ISBN 1-85166-523-4, p.567. Stang, W., Fischer, A., Rubischung, P.: Large-scale application of segmenting and decontamination techniques, Decommissioning of Nuclear Installations, Bruxelles (B), October 24-27, 1989, EUR 12690, 1990. Wille, H., Bertholdt, H.O.: Recent developments in component and system decontamination, conference on water chemistry of nuclear reactor systems 5. London: BNES, 1989, p. 163. Bertini, A.: Some remarks about decontamination; Decommissioning of Nuclear Installations, Bruxelles (B), October 24-27, 1989; EUR 12690, 1990. Sappok, M., Lukacs, G., Ettemeyer, A., Stang, W.: Melting of radioactive metal scrap from nuclear installations, Decommissioning of Nuclear Installations, Bruxelles, October 24-27, 1989, EUR 12690, 1990.
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8 Decontamination Brunel, G., Gauchon, J.-P., Kervegant, Y., Josso, F.: Nouvelles techniques de décontamination: Gels chimiques, electrolyse au tampon et abrasivfs; EUR 13497 FR, 1991. 1990 Recommendations of the International Commission on Radiological Protection ICRP Publ. 60; Annals of the ICRP, 1991, Vol. 21 No. 1-3; Pergamon Press Oxford, New York, Frankfurt, Seoul, Sydney, Tokyo (1991) Commission of the European Communities: Ispra, Joint Research Centre, Institute for Safety Technology, Private Communication, 1992. H.G. Pratzel: Dekontamination der Haut aus der Sicht experimenteller Ergebnisse; in SSK Veröffentlichungen Band 18; Gustav Fischer Verlag Stuttgart, Jena, New York, 71-95 (1992) Strahlenschutzkommission SSK: Massnahmen nach Kontamination der Haut mit radioaktiven Stoffen; Empfehlung der SSK vom 22. September 1989; in SSK Veröffentlichungen Band 18; Gustav Fischer Verlag Stuttgart, Jena, New York, 1-30 (1992) P. Gerasimo, D. Jourdain, A. Cazoulat, D. Schoulz, P. Laroche, R. Ducousso: Modeling of cutaneous radio-contamination: effects of washings by soap and by solutions of DTPA (in French); Ann Pharm. Fr. 55 (3), 116-124 (1997) Hirabayashi, T., Ishigure, K., et. al.: Handbook of reactor water chemistry, Corona Pub. Co., 2000, p. 276. Hirabayashi, T., Ishigure, K., et. al.: Handbook of reactor water chemistry, Corona Pub. Co., 2000, p. 275. Kato, S., Ishigure, K., et. al.: Handbook of reactor water chemistry, Corona Pub. Co., 2000, p. 284. Kato, S., Ishigure, K., et. al.: Handbook of reactor water chemistry, Corona Pub. Co., 2000, p. 284. Wille, H., Bertholdt, H., Lessons, H.O.: Learned in full system decontamination by application of the CORD family concept, BNES,VIII Int. Conference on Water Chemistry of Nuclear Reactor Systems, Bournemouth, UK -26.10.2000. Eickelpasch, N., Steiner, H.: Stilllegung von Kernkraftwerken, VGB Power Tech. 6, 2001, p. 142.
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This Chapter first provides an overview of the factors which influence the treatment of persons internally contaminated with radionuclides and of the available methods of treatment. However it is devoted mainly to the decorporation of tritium, strontium and iodine isotopes and the actinides plutonium, americium, thorium and uranium which continue to be a matter of concern. Important cases published in the scientific literature are summarised and progress made in research studies designed to optimise treatment for different chemical forms of the actinides reviewed. The Chapter concludes with priorities for future research and an extensive bibliography.
List of symbols and abbreviations a ALI Bq CED DTPA EDTA EHDP EU ICRP ID ILD i.p. i.v. log β N ORAU SD SE Sv TBP
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Year (annum) Annual Limit on Intake Becquerel Committed Effective Dose Diethylenetriaminepentaacetic acid Ethylenediaminetetraacetic acid Ethane-1-hydroxy-1,1,-biphosphonate European Union International Commission on Radiological Protection Injected Dose Initial Lung Deposit Intraperitoneal injection Intravenous injection The overall stability constant for a metal-ligand complex or chelate Number of observations Oak Ridge Associated Universities Standard Deviation Standard Error of the Mean Sievert Tri-butylphosphate Mean (average)
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9.1 Introduction Several useful handbooks and reviews on the decorporation of radionuclides from the body have been published over the years [78V1, 80N1, 84W1, 92B1, 00H1]. However, in the course of time, views and opinions change on the need for treatment, the radiation doses at which treatment should be considered or implemented, the most appropriate substance to be used and the optimum treatment regimen. The purpose of this Chapter is to review and update these issues with particular emphasis on hydrogen (tritium), strontium, iodine, caesium, plutonium, americium, thorium and uranium. These elements are amongst those of most concern as a consequence of accidents and incidents involving radioactive materials. Priority is given to uptakes resulting from inhalation and wound contamination. The Chapter commences with an overview of factors that affect the efficacy of treatment, treatment decisions, decision levels and the perception of risk (Section 9.2). This is followed by summaries of the various methods of treatment (Section 9.3), the efficacies of chelating agents for different chemical forms of the actinides (Section 9.4) and recent developments in this field (Section 9.5). Much of the Chapter is devoted to the most effective treatment regimens for different chemical forms of the elements considered here, as identified by both human experience and animal studies (Section 9.6). The Chapter concludes with suggestions for future research (Section 9.7) and a comprehensive bibliography (Section 9.8).
9.2 General considerations 9.2.1 Factors affecting the efficacy of treatment The efficacy of treatment using chelating agents can be affected by the mode of intake, mass and physicochemical form of the contaminant, the reactions of the radionuclide with biological ligands at the site of entry in the blood and at the sites of secondary deposition, the absorption kinetics of the radionuclide into the blood, the method and duration of treatment and the mole ratio of the radionuclide to chelating agent. In principle, the efficacies of clinically approved chelating agents are best evaluated after accidental human exposure. In practice this may be difficult for some radionuclides, notably the actinides, owing to uncertainties in the physico-chemical form, pattern of intake, and assessment of intake. Moreover, the chelating agent may not have been administered by the most appropriate route or the optimum protocol adopted. Animal studies, when properly executed, need not suffer these disadvantages and moreover are likely to be the only effective means for evaluating new substances and protocols. The method of administration favoured by most physicians is slow intravenous injection or infusion since it is considered that chelation will be most effective when the radionuclide is present in circulating blood. In general, this is not true. Many studies with laboratory animals have shown that chelating agents are most effective for biologically soluble forms of radionuclides when they are present at the site of deposition, for example in the lungs or at a wound site. In these circumstances, local administration of the chelate is almost certainly the best option. However, when absorption into the systemic circulation occurs over an extended period then continual intravenous infusion, either directly or as a consequence of oral administration may be the most effective regimen. For inhaled biologically insoluble materials, bronchopulmonary lavage may be the only viable method of treatment (see Sections 2.3.1, 2.4)
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9.2.2 Factors influencing treatment decisions The aim of treatment is to reduce the risk of deleterious effects to the patient, usually cancer, by reducing the radiation dose. However, for some compounds of uranium considerations of chemical toxicity may over-ride radiotoxicity. It is most important that the organisation or industry should have a clear policy concerning treatment and that all personnel are familiar with this policy. This should state that the final responsibility for treatment must always rest with the physician or appointed doctor. Nevertheless, in formulating policy several points should be addressed, such as: • • • • • • • •
Does the maximum credible dose warrant treatment? What are the uncertainties in the assessment of intake and dose? Can any intake of sufficient magnitude to require treatment be rapidly confirmed? Is the material amenable to treatment? What is the likely reduction in dose? Has the age and general health of the individual been taken into account? What is the psychological condition of the individual? Is the outcome of treatment likely to be beneficial when compared with the potential risks ?
9.2.3 Decision levels An important aspect of radiological protection for the organisation or industry concerned is that there should be a clear policy concerning treatment and that all personnel are familiar with this policy. The International Commission on Radiological Protection (ICRP) in Publication 60 [91I1] advises against the application of current dose limits for deciding on the need for, or scope of, treatment whilst recognising that at some level of dose treatment should occur. Hence a clear distinction must be drawn between the dose limit and decisions concerning treatment. Nevertheless, in practice, treatment decisions, other than for soluble compounds of uranium, will usually be related to the effective dose limits recommended by the ICRP, namely 20 mSv a−1 averaged over a defined period of 5 a with a further provision that the dose should not exceed 50 mSv in a single year [91I1, 96O1]. Since in many countries the annual dose limit is restricted to 20 mSv a−1, this value forms the basis of the decision levels suggested here. It is recognised that there are likely to be differing views on the magnitude of such decision levels. The recommendations given below are the same as those given in recently published EU reports on decorporation from the human body [92B1, 00W11]. 9.2.3.1 Inhalation For intakes of biologically soluble material, treatment should not be considered when the assessed dose is below 20 mSv. For assessed doses between 20 mSv and 200 mSv, treatment should be considered [92B1, 00W1]. Although clinical effects from the intake are unlikely, psychological factors will probably be important. Single or short-term administration will usually be sufficient. However, if the assessed dose is greater than 200 mSv, then extended or protracted treatments should be considered depending on the magnitude [92B1, 00W1]. For intakes of biologically insoluble material such as 239Pu dioxide, the treatment of choice is bilateral pulmonary lavage but should only be undertaken if there is a likelihood of deterministic effects [92B1, 00W1]. It has been suggested that lung lavage should be considered only when the estimated lung dose is Landolt-Börnstein New Series VIII/4
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likely to exceed 5 Sv within a few weeks. It should be noted that whilst this procedure is considered to be of low risk [95D1], attributable to mortality from general anaesthesia [00W1], the lung content will be reduced only by about two-fold [89N1]. Biologically soluble compounds of low enriched, depleted or natural uranium are potentially nephrotoxic. The basis for current limits on intake is a maximum kidney concentration of 3 µg g−1 [73S2, 96H]. It can be deduced, using the ICRP human respiratory tract model [94ICRP] and the systemic model for uranium [95ICRP], that this value will be attained after acute inhalation of 30 mg and 230 mg of a very soluble (Type F) or a moderately soluble (Type M) compound, respectively, of uranium [97S1, 98S2]. 9.2.3.2 Wound contamination After a serious accident involving injury and wound contamination, necessary life-saving procedures must take precedence over decontamination. For accidents involving biologically soluble forms of radionuclides, the first approach should be to reduce contamination by copious washing with water. However, if radioactivity has entered the systemic circulation, similar criteria to those described previously for inhalation should apply. For wounds contaminated with insoluble materials, washing with copious amounts of water should again be considered first. In many cases deposits at wound sites can be removed by surgical excision. Under these conditions it is considered inappropriate to recommend decision levels since many physicians would wish, provided there is little risk of functional impairment, to remove the radioactivity until it is below the limit of detection, perhaps a few tens of Bq or less. When, there is a risk of impairment, a balanced judgement must be made by the physician, preservation of normal function always being the primary objective. 9.2.3.3 Ingestion For radionuclides that are extensively absorbed into the bloodstream such as 3H and 137Cs, the criteria for treatment should again be the same as for inhaled soluble compounds. For ingested insoluble materials the dose to the lower intestine may be large, with the possibility of deterministic effects. In these circumstances the use of cathartic or binding agents to accelerate faecal excretion should be used.
9.2.4 Perception of risk and its implications Rather than have a series of risk coefficients for different individuals in different circumstances, the most conservative approach will be considered here, i.e. the risk for members of the public. The probabilities assumed by ICRP for the risk of radiation induced stochastic effects in members of the public are illustrated in Table 9.1, which shows that the overall risk of health detriment from stochastic effects will be 7.3 %/Sv [91I1]. Table 9.1. The nominal probabilities for radiation induced harmful effects (from ICRP Publication 60, 1991); [91I1]. Risk [%/Sievert] Fatal cancer Non-fatal cancer Serious hereditary Total effects General public 5.0 1.0 1.3 7.3
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Since decisions on treatment will be based on the net benefit to the patient, it is more appropriate to consider the overall risk rather than only the risk from fatal cancers. For the suggested decision levels of 20 mSv and 200 mSv, the overall risks from stochastic effects are 0.15 % (~1 in 700) and 1.5 % (~1 in 70). However, other risks to be addressed include the following. 9.2.4.1 The risk from the administration procedure The route of administration carrying the most significant risk is intravenous injection, in which an air embolism, leading to serious cardiovascular or neurological effects, or to death is likely to occur in 1 in 20,000 injections (0.005 %) [00W1]. Minor, and reversible, adverse reactions are known to occur in 1 in 40 injections [00W1]. Clearly the likelihood of adverse effects would be increased by repeated administration. For treatment of inhaled insoluble materials by whole lung lavage, the risks are considered to be essentially those of a general anaesthetic- between 1 in 50,000 (0.002 %) and 1 in 200,000 (0.0005 %) [00W1]. It is considered that this level of risk justifies the use of lavage to reduce potential deterministic effects, but only for lung doses in excess of 5 Sv [00W1]. 9.2.4.2 The risk from adverse effects of the therapeutic agent This is difficult to quantify, and reference to well defined case histories provides the best information. For example, other than for uranium, the agent of choice for most actinides is DTPA and the usual human dosage is 0.5 to 1 g of the calcium or zinc salt. In France, over 500 workers have been given a single dose by slow intravenous infusion and over 200 workers have received multiple doses of DTPA without adverse effects [87B1]. The Oak Ridge Associated University (ORAU) Registry reported that between 1958 and 1987, 485 patients received a total of 3,077 dosages of DTPA, about two-thirds of them as the calcium salt. Minor transient effects were observed in 12 patients, but no serious or long term effects were reported [87B1]. In the Hanford americium accident, 583 g of DTPA, primarily as the zinc salt were administered to an individual over a 4-year period without any observed toxic effects [89B1]. 9.2.4.3 The reduction in risk from treatment This is again difficult to quantify since it will depend amongst other things on the biokinetics of different chemical forms of the radionuclide, the method of administration of the chelating agent, and the frequency and duration of treatment. The spectrum may range from marginal to almost complete removal of the radionuclide from the body. In broad terms, the extent of removal will reflect the reduction in risk. 9.2.4.4 The risk to the patient in the absence of treatment Clearly, the risk coefficients for health detriment and the risks associated with the various treatment procedures referred to above will affect the decision making process. It should be noted that when the estimated doses are less than about 20 mSv, the risk of treatment may surpass the anticipated benefit.
9.2.5 Approaches to treatment In broad terms, there are two alternative approaches on which treatment decisions are based. For convenience these are referred to here as urgent and precautionary. Landolt-Börnstein New Series VIII/4
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The urgent approach is advisable when a potentially serious intake is suspected but which would take time to confirm. In this approach the chelating or complexing agent should be administered as a single dose as soon as possible. The desirability of further administrations would be decided when additional information on the physico-chemical form of the contaminant, individual monitoring data (whole-body monitoring or bioassay data as appropriate) or the psychological reaction of the patient becomes available. The advantage of this approach is that if a high uptake is confirmed, therapy, at least in most cases, will have commenced at the optimum time. The disadvantage is that if the uptake was not confirmed, or was trivial, or the material was not amenable to effective chelation therapy, then the patient might have been subject to an unnecessary, albeit small, risk. It should also be remembered that even the single administration of a chelating agent may substantially delay the accurate assessment of uptake, particularly if this is from excretion monitoring, and may increase stress to the patient. On the other hand, treatment often reduces stress to the patient. In the precautionary approach, treatment is withheld until uptake is confirmed. The decision to treat can then be based on the likely magnitude of the uptake and probable reduction in risks of late effects. Although confirmation of small uptakes can take some time, the advantage of the precautionary approach is that should the intake be unconfirmed, or assessed as low, then any risks associated with treatment will have been avoided. The disadvantage of this approach is that should the estimated uptake be above the decision level, then the efficacy of treatment is likely to be reduced appreciably. Other than for lavage, which can be delayed for a few weeks without reduced effectiveness, the authors do not favour one approach over another. They leave any decisions to the professional judgement of the physician and radiological health-team who will have considered all the options based on local knowledge.
9.3 Methods of treatment This Section considers in some detail the different treatment regimens suitable for removing radionuclides from the body, but with emphasis on tritium, iodine, strontium, caesium, and the actinides plutonium, americium, thorium and uranium.
9.3.1 Non-specific procedures These procedures can be applied to any radionuclide and any radioactive compound. They include gastric lavage to remove material from the stomach; copious washing of a wound; the administration of laxatives for cleansing the gastrointestinal tract; surgical incision for removing material from a wound; and pulmonary lavage for removing insoluble material from the lungs. 9.3.1.1 Removal from the gastrointestinal tract Non-specific procedures can be effective when used immediately after ingestion of radioactive material which is rapidly absorbed from the gastrointestinal tract or which may result in a high dose to the intestine. Orally administered antacids or adsorbents are useful for reducing the uptake of soluble forms of radionuclides from the gastrointestinal tract. The substances recommended have been used frequently in clinical practice and they represent virtually no risk to the patient after short- term administration. Suitable laxatives, such as sodium or magnesium sulphate, will be desirable for reducing irradiation of the lower large intestine irrespective of the chemical form. Enemas or colonic irrigation may also be used for the same purpose. Landolt-Börnstein New Series VIII/4
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Specifically, aluminium phosphate or aluminium hydroxide is suitable for strontium (and barium and radium) [69S2, 92B1], whilst Prussian Blue (ferric ferrocyanide) [Fe4(Fe(CN6)3] will bind caesium (and rubidium and thallium) in the gut by ion exchange. Prussian Blue thus renders caesium insoluble in the intestinal lumen and prevents initial absorption from the gut. By breaking the secretion–reabsorption cycle, its continuous administration will reduce appreciably the systemic content of the element. At the recommended human dosage, usually 3 g d−1, Prussian Blue has no known toxicity [92B1]. The oral administration of alginate has been investigated for strontium [64W1, 67H1 68C1], barium [72H2] and radium [72V1]. 9.3.1.2 Lung lavage Lung lavage is used to remove alveolar macrophages from the lungs in which particulate material is entrained. During a lavage procedure, both lungs will be treated alternately under a single general anaesthesia with multiple washes of warm isotonic saline whilst oxygen is administered to the other lung. The procedure can be repeated if necessary after 3-4 days. The technique should not be performed before the particles in the upper airways have been cleared naturally, but it remains a viable option up to several weeks after exposure. However the total amount of material which can be removed from the lungs does not generally exceed 50 % [89N1]. The risk associated with lavage is mainly that of general anaesthesia. However, it has been suggested that it should only be used on healthy people and where the radiation dose over a period of a few weeks is likely to exceed 5 Sv [00W1].
9.3.2 Procedures to enhance systemic radionuclide excretion As indicated previously the procedure of choice will be determined by the biokinetic behaviour of the contaminant and the different mechanisms by which excretion can be enhanced e.g. by the use of diluting, immobilising or chelating agents. 9.3.2.1 Diluting and immobilising agents An important example of a diluting agent is the enhancement of tritiated water excretion by means of forced fluid intake, often in combination with a diuretic, under medical supervision. In these circumstances, the biological half-time of 3H in the body, usually about 10 d, can be reduced by about two-fold during the period of treatment; the reduction in the committed effective dose is somewhat less due to the short period of treatment [71L1, 72H1, 86L1]. Another important example is the reduced deposition of iodine in the thyroid by the administration of stable iodide or iodate immediately after intake of the radio-isotope. The most effective treatment regimen for systemic radio-strontium appears to be the prompt intravenous or oral administration of stable alkaline earth metal salts, usually as their gluconates [80N1, 84W1, 92B1]. Similarly, attempts can be made to dilute, and hence reduce the systemic deposition of cobalt and radium isotopes by the administration of stable isotopes, or analogous elements. At present the best available treatment for systemic caesium is the oral administration of Prussian Blue, which immobilises the element in the gastrointestinal tract (see 9.3.1.1). However, even with extended administration the reduction in dose is likely to be only about 2 to 3 fold [94M1, 98I1].
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9.3.2.2 Chelating agents The formation of radionuclide complexes in the body that lead to their excretion via the kidneys and urine, and/or liver and faeces is the most appropriate procedure for many radioactive heavy metals, in particular the lanthanides and actinides. The chelates most widely used for enhancing the excretion of plutonium, americium and thorium isotopes are the trisodium calcium or zinc salts of diethylaminetriaminepenta-acetic acid, referred to hereafter as CaDTPA and ZnDTPA. The former is normally used for initial and single administration, but since it can remove the essential biometals iron, manganese and zinc from the body, the zinc salt is preferred for extended or protracted administration. The mode of action of DTPA is the formation of chemically stable complexes of radionuclides in the extracellular fluids, most potently lung fluid and blood, that are rapidly excreted in the urine and, to a lesser extent, the faeces without being reabsorbed. DTPA is normally administered by slow intravenous infusion or injection at dosages of 15-30 µmol kg−1 body mass (0.5-1 g for a body mass of 70 kg). Alternatively, it can be administered as an aerosol or orally, usually at a similar dosage. For wound contamination, local infiltration of the substance is likely to be most effective, but because severe pain is likely to be associated with the intramuscular injection of the DTPA, a local anaesthetic, e.g procaine should be added to the solution. No serious side effects have been observed in humans treated with DTPA [87B1, 89B1, 98 G1] (see Section 9.2.4). There is no evidence that DTPA is effective for significantly enhancing the removal of uranium from the body. In some guidebooks, the recommended agent is sodium bicarbonate. However, this is not supported by controlled studies with laboratory animals under realistic conditions e.g. with delays between exposure and treatment of 30 min or more. Chelation therapy is not an option for the alkaline earth elements strontium, barium or radium since EDTA (ethylenediaminetetraacetic acid), DTPA, and most other chelators, form stronger complexes with calcium, than with strontium, barium or radium. Thus calcium will be complexed preferentially, and no useful enhancement of the excretion of the other alkaline earth metals will be achieved. This point is illustrated by the stability constants given in Table 9.2. The stabilities are expressed as the overall constant β which is the product of the formation constants for each of the individual metal-ligand reactions involved in the formation of the chelate of interest. For convenience the values are given as log β, the negative logarithm of the constant. Table 9.2. Stability constants log β (see text) for the complexes between the alkaline earth metals and EDTA and DTPA [74M1] Element log β EDTA DTPA Ca 10.69 10.83 Sr 8.73 9.77 Ba 7.86 8.78 Ra 7.1 [7.9]* * Estimated value.
9.4 General comments on the efficiacy of chelating agents for the actinides In general, the greatest problems posed in decorporation of radionuclides from the human body involve the actinide elements. Most of the research conducted in the last decade or so has also concentrated on these elements. This Section concerns the authors’ responses to some of the most frequently asked questions on the efficacy of chelating agents for the actinides. Landolt-Börnstein New Series VIII/4
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9.4.1 What are the factors that govern the efficacy of chelating agents ? The efficacy of treatment can be influenced by the mode of intake, mass and physico-chemical form of the contaminant, the reactions of the radionuclide with biological ligands at the site of entry, the absorption kinetics of the radionuclide into the blood, the method and duration of treatment, the formation constant of the metal-ligand complex and the ligand-metal mole ratio. The mass of the material deposited in the respiratory tract or at a wound site is an important consideration for predicting the likely efficacy of treatment in human beings and for designing animal studies. This is particularly relevant for plutonium, americium and thorium which hydrolyse readily at physiological pH, but will also be important for uranium which can precipitate as phosphate in the lungs. In animal studies, 238Pu is used frequently for providing mass concentrations of Pu in the respiratory tract which simulate human exposures to 239Pu (see Section 9.6) more realistically. Clearly the physico-chemical form will also dictate the availability of the radionuclide to react with the chelator and thus to enhance excretion. For different materials, the efficacy may be influenced by the ultrafine component, the rate of dissolution of the particles in-vivo, the reaction of the radionuclide with biological ligands at the site of deposition and in systemic tissues, its rate of absorption into the blood and the tissue distribution. The influence of some of these factors are described in more detail in Section 9.6. The overall efficacy of treatment will also be influenced by the mode of intake of the radionuclide and the chelating or complexing agent. Invariably, the most likely routes of internal contamination result from inhalation and wound contamination. After inhalation, chelating or complexing agents could in principle be administered as an aerosol, by intravenous injection or infusion, or orally. The choice between the methods will depend on the biokinetics of the contaminant, and whether the substance will cross the airblood barrier or gut wall in sufficient amounts. After wound contamination, chelating agents could be administered by intravenous injection or infusion, or by local injection. The data obtained from animal studies suggests that local injection is the preferred method (see Section 9.6)
9.4.2 Can the efficacy of treatment be predicted from animal studies ? Yes, provided the aerosol characteristics, the mass concentrations of the appropriate chemical forms at the site of deposition and the mode of uptake represent a realistic accident scenario. In many cases this may involve the use in animals of a higher specific activity isotope e.g. 238Pu rather than 239Pu. Any conclusions reached purely on the basis of intravenous injection experiments should be treated with caution. The differences in the distribution pattern of the radionuclide between species and in their absorption rates to blood should also be recognised. For example, after the inhalation of a moderately absorbed compound such as plutonium nitrate, the retention half-time of plutonium would be greater in the human lungs than in the rat lungs and the fraction of absorbed plutonium deposited in the human skeleton would be less than in the rat skeleton. On both counts, the overall efficacy of say DTPA would be expected to be higher in the human than in the rat (see Section 9.6).
9.4.3 Are chelating agents always most effective when the radionuclides are present in circulating blood ? No! Ultimately the efficacy of the chelate will be influenced by the biokinetics of the contaminant, and this should be taken account of in designing the treatment regimen. Often, for radionuclides that are biologically soluble such as plutonium and americium nitrate, chelating agents will be most effective when they are deposited at the same site as the contaminant, e.g. in the lungs or at a wound site. However, for radionuclides which are absorbed into the blood at a moderate rate over a period of time, such as with 238 PuO2 and 241AmO2 after inhalation, then it may be more productive to complex the radionuclide in the blood so as to prevent its deposition in systemic tissues such as liver and bone. This may require the continual infusion of the chelate, or its oral administration in drinking water over weeks or months. Landolt-Börnstein New Series VIII/4
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9.4.4 Is DTPA effective for all actinides ? No! The major successes in animal studies have been with biologically soluble or moderately soluble forms of plutonium and americium referred to above. However at present the administration of DTPA cannot be considered an effective method of treatment for soluble thorium, uranium and neptunium compounds after inhalation or wound contamination under realistic conditions [00S1, 00S2].
9.4.5 Will the administration of chelating agents result in enhanced tissue deposition ? There appears to be no evidence from either human or animal studies that this is an important consideration when DTPA is used for the decorporation of plutonium and other actinides. However this may not be true for other chelators, particularly when they are unstable at physiological pH. For example, research studies with the siderophore analogue 3,4,3-LICAM(C) indicated enhancement of plutonium deposition in the kidneys [89S2, 89D2], whilst some phosphonates increase substantially the deposition of uranium in the liver [98H1].
9.4.6 Is the administration of sodium carbonate effective for uranium ? The administration of sodium bicarbonate has been recommended in various guidebooks and handbooks for the decorporation of uranium [80N1, 84W1, 92B1]. The evidence available suggests it is not effective, and in view of the possible side effects such as hypokalaemia and respiratory acidosis its use should be re-considered for human treatment. Whilst alternative substances such as tiron and some polyphosphonic acids and the siderophore analogue 3,4,3-LI (1,2-HOPO) have been suggested, the experimental data show that apart from instantaneous administration they are only partially effective, and usually high dosages are required [98H1]. The effective decorporation of uranium remains an important problem in radiological protection.
9.4.7 Must chelating agents be administered promptly to be effective ? In most cases, yes! It is particularly effective for soluble compounds of plutonium and americium deposited in the lungs or at wound sites [00S2, Section 9.6]. Prompt administration will minimise deposition in systemic tissues such as liver and bone from where appreciable removal is exceedingly difficult. However, the lung content can still be reduced appreciably should treatment be delayed for several days, although the evidence available suggests that this is not true for wounds [00S2, Section 9.6]. For other chemical forms of plutonium and americium such as 238PuO2 and 241AmO2, prompt administration will have little effect, and extended treatment as described above, and in Section 9.6 will be more appropriate. If effective chelators were available then prompt treatment is essential for soluble uranium compounds after inhalation and wound contamination in order to mimimise the nephrotoxic effect.
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9.4.8 Is intravenous injection the best mode of administration ? This method, favoured by many clinicians, is the means by which most chelating agents can be administered rapidly. Whilst, in principle, an even more rapid response and greater efficacy (Section 9.6), could probably be achieved with an aerosol form self-administered with a spinhaler, there is some doubt about the extent of aerosol deposition in the lungs, and the procedure may be in contravention of the standing instructions of the employing organisation which would require that a medical officer administers the substance. However in accidents involving wound contamination, or after intakes requiring the extended administration of chelates, then intravenous injection would not be the most appropriate method, and in such cases local and oral administration respectively would need to be considered.
9.4.9 How can judgements on efficacy be made ? In the emergency planning stage by consulting the scientific literature on the biokinetics of the same or similar material to ascertain the absorption kinetics, and whether human or animal data are available to indicate the likely efficacy of treatment and the optimum treatment regimen. After the accident from assessments of intake, and retention and excretion data using the most appropriate methods e.g. chest monitoring, wound monitoring, bioassay.
9.4.10 When should treatment start ? This will depend on the biokinetics of the contaminant. In most cases, whatever the radionuclide and the route of contamination, treatment should begin as soon as possible. Delays of a few days would be appropriate if lung lavage is considered an option (Section 9.2), and might not be critical for compounds such as 241AmO2 which dissolve fairly slowly in the lungs (see Section 9.6).
9.4.11 When should treatment stop ? Several criteria are possible. One, when it is evident that the excretion of the contaminant is low compared with that expected from the estimated internal deposit. This judgement may not be straightforward. For example, a substantial increase in the urinary excretion rate above background may still represent a very small fraction of the uptake. On the other hand, an apparent lack of early success does not preclude the effectiveness of extended therapy, as described for 241AmO2 (Section 9.6). Two, when the dose or risk has been reduced to an acceptable level (Section 9.2), taking account of the psychological needs of the patient.
9.4.12 For which materials are chelating agents likely to be effective ? In the context of this Section, effective treatment implies that the reduction in the committed effective dose is likely to be at least two-fold, and hopefully much greater. Judgements on the efficacy of treatment can be based on the published biokinetic behaviour of known chemical forms in animals, and knowledge of the treatment regimens which can be effective for known soluble or moderately soluble forms; these Landolt-Börnstein New Series VIII/4
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are described in more detail in Section 9.6. Whilst ligands currently approved for human use must be of prime consideration, account should be taken of experimental studies with new substances which may appear to be appreciably superior (Section 9.6) • Provided the criteria affecting decision levels are met (Section 9.2.3), treatment is expected to be effective for 238Pu and 239Pu inhaled or deposited in wounds as a pure chemical form, e.g. nitrate or tributylphosphate, provided that treatment commences early and continues for a few weeks (see Section 9.6) • 239Pu inhaled as an oxide with a large ultrafine component (ca 50 % by activity), such as Pu-Na mixed oxide aerosols [78S1, 79S1 80M1] • 238PuO2, provided treatment is extended over many months in order to chelate the Pu arising from the dissolution of fragmented particles [80 S1, 83M1] • 241Am inhaled or deposited in wounds as a pure chemical form of the nitrate, provided treatment commences early and continues for a few weeks ( see Section 9.6) • 241Am present in residues resulting from the refining of Pu metal, provided treatment continues for a few weeks [87S1] • 228Th inhaled or deposited in wound sites as the nitrate, provided treatment commences promptly and continues for a few weeks (see Section 9.6) • In principle inhalation and wound contamination of 237Np and 239Np. However no effective clinically approved substance is currently available [00S1] • In principle, uranium inhaled or deposited in wounds as ammonium diuranate, trioxide, nitrate, tributylphosphate, hexafluoride and tetrafluoride, octoxide and dioxide inhaled as ultrafine particles. However no effective clinically approved substance is currently available [00S1]
9.4.13 For which materials are chelating agents unlikely to be effective ? Based on extensive biokinetic studies in laboratory animals to which the appropriate literature references are given, it is considered that treatment is unlikely to be effective for • • • • • • • •
238
Pu and 239Pu nitrate intermixed with corrosion products or building dust [87S1, 94M2] PuO2+ 241AmO2 formed by calcination at high temperatures where 241Am is present as a decay product of 241Pu [87S1, 95S1] 239 Pu present in residues resulting from the refining of plutonium metal [87S1] 239 Pu present in residues arising from the corrosion of magnox fuel [89S1] 232 Th nitrate , fluoride, hydroxide or dioxide [00S2, 93M1] 237 Np or 239Np dioxide [96I1] uranium octoxide or dioxide, unless there is a substantial ultrafine component [94A1, 95S1, 96I1, 98A1] For other materials of potential concern it is recommended that further research on the absorption kinetics in laboratory animals is undertaken in order to make judgements of the likely efficacy of treatment. 239
9.4.14 Is lung lavage more effective than chelation treatment for inhaled materials ? Only if they are essentially insoluble, or dissolve slowly in the lungs over a long period of time. However, it should be remembered that at best, lung lavage will remove about one half of the radioactivity. The method would be inappropriate for biologically soluble forms of plutonium and americium. However for high uptakes of thorium nitrate and 241AmO2, the choice between chelation treatment and lavage is more difficult to determine. Landolt-Börnstein New Series VIII/4
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9.5 Recent developments No significant developments in enhancing the excretion of tritium, radiostrontium, radioiodine or caesium from the body have taken place for many years. For 3H and radioiodine, it is difficult to envisage how those procedures in current use can be improved. Recent developments in decorporation therapy have focussed primarily on the actinides. This is because the efficacy of DTPA for a variety of chemical forms of plutonium, americium and thorium taken into the body by various routes has not been fully examined, or treatment using the usual route of intake, intravenous infusion, has not been completely effective. In addition, no effective agent for uranium appeared to be available. All this work has been reviewed in more detail elsewhere [ 94S1, 98S1, 98 S2, 00S1,00S2]
9.5.1 Plutonium and americium It has been recognised for many years that anologues of siderophores were likely to be more effective than DTPA. Siderophores are sequestering agents produced by microorganisms in order to obtain Fe(III) from their environment. The basis for this approach was that since the biokinetics of the actinides in mammals are associated with the Fe(III) transport and storage systems, then the formation constants of the actinide complexes with siderophore derivatives would be much higher than with DTPA. This was subsequently found to be the case. Many siderophore analogues such as a linear catechoyl amide codenamed 3,4,3-LICAM(C), a dihydroxamic derivative of DTPA, DTPA-DX, a hydroxypyridinone derivative of desferrioxamine, DFO-HOPO, a hydroxypyridonate code-named 3,4,3-LI(1,2-HOPO) and ligands containing the isomer 3,2-HOPO have been synthesised and tested for the decorporation of plutonium and americium, usually after their intravenous injection as citrates [98D1]. However, animal experiments involving inhalation and simulated wound contamination using different chemical forms of these elements, and administration of the ligand by different routes, showed repeatedly that 3,4,3-LI(1,2HOPO) was substantially superior to DTPA (see Section 9.6). Whilst in the early stages the synthesis of 3,4,3-LI(1,2-HOPO) was difficult and expensive, this difficulty has now been largely overcome [98B1]. However it has not yet been approved for human use and studies on optimising treatment for different chemical forms of plutonium and americium with DTPA remain an important aspect of decorporation therapy.
9.5.2 Thorium Studies with rats have shown that DTPA is ineffective for 232Th when deposited as nitrate in the lungs in amounts that correspond to the annual dose limit for workers. At low masses of thorium, as 234Th, DTPA is moderately effective after simulating wound contamination by subcutaneous or intramuscular injection, provided it is administered within minutes. However the most effective ligand developed so far for the decorporation of thorium after inhalation and wound contamination is 3,4,3-LI(1,2-HOPO). More information on the most effective treatment protocols is given in Section 9.6.
9.5.3 Uranium Soluble compounds of uranium are nephrotoxic [89D1, 89L1]. Hence it is important that treatment should be prompt and effective. Several substances have been investigated in animals. These have included phenolic compounds such as Tiron, polyaminophosphonic acids, bisphosphonates, and phosphoalkylphosphinates, and more recently the siderophore analogues 3,4,3-LI(1,2-HOPO) and Landolt-Börnstein New Series VIII/4
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4-LI(Me-3,2-HOPO). Some of these compounds can reduce the kidney content by about an order of magnitude compared with untreated animals when the uranium and ligand are administered almost simultaneously, or within a few minutes. However the efficacy falls sharply with any delay in administration and they are ineffective beyond about 30 minutes post exposure.
9.6 Optimum treatment protocols This Section reviews human data where sufficient good quality information is available. However important data are often obtained from controlled studies with laboratory animals. This procedure is particularly useful since more than one regimen can be compared for the same exposure scenario, and new agents can be compared directly with the current clinically approved substance. The emphasis is placed on inhalation, wound contamination and ingestion. Whilst intravenous injection of both radionuclides and ligands are used widely in the testing of new substances in animals, this route of contamination is not important from the standpoint of accidental exposure. However, it is considered here when data on the other modes of intake are unavailable. In addition it should be borne in mind that the high efficacy of a ligand observed after intravenous injection of a radionuclide does not necessarily mean that this will be the case after inhalation or wound contamination. Conversely, the efficacy may be higher after these routes of intake than after intravenous injection of the radionuclide.
9.6.1 Tritium 9.6.1.1 Human data Incidents involving the uptake of substantial amounts of HTO are rare. The following case is included because it resulted in a substantial intake, about 35 GBq, and the treatment regimen used represents the optimum that can be achieved in practice [86L1]. The individual was encouraged to increase fluid intake soon after the accident and under medical supervision in hospital, forced diuresis was commenced 100 h after the accident and continued for 4 d. Diuresis was induced by an intravenous infusion of 7 litres per day, alternating 1 litre of isotonic saline with 1 litre of 5 % glucose, both being supplemented by 20 mM of potassium chloride, and further enhanced by giving 40 mg furosimide intravenously each day. In the 4 d of diuresis, which was considered the maximum medically justifiable, 15 GBq of 3H was excreted in the urine. It was calculated that the above treatment regimen reduced the radiation dose from 800 mSv to 470 mSv. Had forced diuresis commenced immediately after the accident, the dose would be about 410 mSv. Thus for treatment of incorporated 3H, the maximum reduction in dose is unlikely to be more than two-fold [86L1]. The authors are unaware of any definitive human or animal data on other chemical forms of tritium.
9.6.2 The alkaline earth elements, strontium, barium and radium The substances recommended for minimising uptake from the human gastrointestinal tract, and hence systemic deposition, are strontium gluconate (isotope dilution), barium sulphate (insoluble sulphate by ion exchange), magnesium sulphate (laxative), colloidal aluminium phosphate (antacid) [80N1, 84W1, 92B1].
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A number of animal studies showed that sodium alginate when given simultaneously or immediately after oral intake of strontium (or baium or radium) was able to reduce intestinal absorption and thus retention of the element [72H2, 72H3, 71V1, 77V1, 78V1, 80K1] with little effect on calcium absorption. A similar effect in humans was reported by Hesp and Ramsbottom [65H1]. Much attention has been paid to methods for enhancing the natural excretion of 90Sr from the body, these include the administration of diuretics, hormones and the administration of a variety of complexing agents [68V1, 68W1, 68S1]. However, none of these have suggested a clinically useful procedure and it is difficult to conceive how the efficacy of removing radiostrontium from the body can be improved with currently available agents and approaches. Some of the substances tested are given in Table 9.3. Table 9.3. Compounds which have been animals or humans Substance Reference Calcium gluconate 68S2, 68V1 Strontium gluconate 68S2, 68V1 Ammonium chloride 68S1, 68S2 Citrate 68S1 Polyphosphates 68S1 Fluoride 68S1 Salicylate 68S1 Phytate 68S1 Alginate 68S1
investigated for the ability to mobilize radiostrontium in Substance Pilocarpine Parathyroid hormone L-Triiodothyronine Oestradiol Hydrocortisone Alginate Chlorothiazid (Saluric®) Mercurihydrin
Reference 68S1, 68S2 68S1 68W1 68W1 68W1 65H1, 68H1, 71V1 68S2 68S2
9.6.2.1 Human data In human volunteer studies Spencer et al [67S1, 67S2, 69S2] found that aluminium phosphate effectively inhibited the absorption of radiostrontium from the human gastrointestinal tract. Aluminium phosphate is used clinically for the treatment of colitis, however, the recommended method of administration is by enema and the only listed pharmaceutical preparation is an aqueous solution containing 6.5 % AlPO4 (Phosphalugel-Klys® 00R1). For the immediate treatment of an accidental oral intake of radiostrontium or radium an appropriate volume of this solution could, with caution, be administered orally. In one study [69S2], a single oral dose of aluminium phosphate gel ranging from 300 ml to 100 ml (100 ml contained 886 mg aluminium and 1016 mg phosphate) was administered to 12 healthy adults immediately before an oral administration of 85Sr half-way through breakfast. The amount absorbed, 3.6±0.5 % (mean ± se) was substantially less than in untreated controls, 28.8±1.9 %. In the same study using 9 volunteers, the amount of 45Ca absorbed in treated and control volunteers was much higher and more variable, 27.0±3.0 % and 45.0±4.2 % respectively. Increasing the dosage of the gel from 100 ml to 300 ml had little effect on absorption in either case. Vanderborght et al [72V1] administered 15 g sodium alginate per day in bread to a woman who had been contaminated accidentally with 226RaSO4. The faecal excretion appeared to be enhanced for several days, suggesting that absorption of the isotope from the intestine would have been reduced. However, this was a single case and any alginate-induced increase in faecal excretion could only be deduced by comparison with published data from persons contaminated under different circumstances. Sodium alginate is a polysaccharide isolated from seaweed and containing guluronic and mannuronic acid residues; it forms very viscous solutions and is not easy to administer in the required amounts. For many of the studies reported the alginate was incorporated into bread, [71V1] at a level of 5 % alginate. However, today a number of alginate-containing pharmaceutical preparations are licensed for human use; for example Gaviscon®, is available as tablets containing 500 mg alginate, 100 mg aluminium hydroxide, 25 mg magnesium trisilicate and 170 mg sodium bicarbonate. Such preparations could be used safely for the immediate treatment of an oral intake of a strontium, barium or radium. Landolt-Börnstein New Series VIII/4
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9.6.2.2 Animal data Humphreys et al [72H1] studied the effect of a single feed of bread containing 5 % of different alginates on the whole-body retention of orally administered 47Ca, 85Sr, 133Ba and 226Ra in mice. Sodium alginate had no significant effect on the retention of 47Ca, but that of 85Sr, 133Ba and 226Ra were reduced by factors of 5, 9 and 9, respectively. Harrison [68H1] also reported that the addition of 10 % of sodium alginate to the diet decreased the absorption of orally administered 85Sr in rats by factors of 4-5 without influencing the absorption of 45Ca. Kestens et al [80K1] fed sodium alginate containing bread daily to mice for 3 months following intraperitoneal injection of 226Ra and studied the retention of the radionuclide in the femur. The 226Ra activity of the femurs was slightly reduced in the alginate-treated animals but the amount of radium removed was independent of the injected dose that varied by a factor of 4. The reduction in femur content presumably reflects decreased uptake due to reduced reabsorption of strontium excreted into the gut, this was reflected in an increased faecal excretion in the treated animals. A reduction in the absorption of 226Ra and 85Sr in mice following administration of aluminium phosphate was reported by Kesley et al. [72K1].
9.6.3 Iodine The substances recommended for human use are potassium iodide or iodate. These can be given orally in the form of a suspension, or as Lugol’s solution which contains 50 mg iodine and 100 mg potassium iodide per ml [80N1, 84W1, 92B1]. One blocking dose of 300 mg potassium iodide, if given within 30 min, will prevent further uptake by the thyroid. However, it may be advisable to administer 100 mg for a further few days to prevent recycling of the radio-iodine. Potassium iodate, at similar doses, can be given as an alternative to iodide. For current guidelines on iodine prophylaxis, and reviews of treatment efficacy, the reader is referred to three recent publications [99H1, 99W2, 00G1]. 9.6.3.1 Human data One of the best examples of the efficacy of prompt and delayed treatment has used human volunteers [67R1]. This work showed that if iodide administration is delayed by 6 h, the thyroid uptake is blocked by only about 50 %, is whilst after 12 h, the uptake of iodine by the thyroid is scarcely affected by the treatment. If stable iodide is given after the first 24 h, there may be a prolonged retention of radio-iodine by the thyroid due to the suppression of thyroid hormone release. Further, besides diluting the radio-iodine, treatment with stable iodide and the massive increase in the iodine pool in the body also inhibits thyroid metabolism. Under treatment with 300 mg sodium iodide followed a few daily doses of 100 mg, toxic reactions are rare, although a few individuals may be over sensitive to iodide and develop angioedema. If a reaction occurs, symptoms should disappear within a few days after cessation of treatment. Iodide should also be administered with caution to persons with goitre or being treated for hyperthyroidosis because the condition may exacerbate to thyrotoxicosis. This condition may also result if individuals have a low dietary intake of iodine. Some people are allergic to large doses of iodide and such cases should be treated with perchlorate.
9.6.4 Caesium The substance recommended for the decorporation of caesium isotopes is Prussian Blue [80N11, 84W1, 92B1]. Landolt-Börnstein New Series VIII/4
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9.6.4.1 Human data One of the major and most comprehensively investigated accidents involving internal contamination with 137 Cs occurred in Goiania in 1987 [94M1, 98I1]. Prussian Blue was administered orally to 46 individuals. The dosages administered ranged from 1 to 3 g d−1 for children, and from 3 to 10 g d−1 for adolescents and adults. In general treatment commenced about 10 d after exposure and continued over a period of about 3 weeks for children and over periods ranging from 3 weeks to 3 months for adults. During the administration of the chelate, the mean retention half-times of 137Cs in the body were, on average, 43 %, 45 % and 69 % respectively of the values after termination of treatment. The committed effective doses were reduced by between 1.7 fold and 6.2 fold, with a median value of 2.1 fold [94 M1, 98I1]. These results are summarised in Table 9.4. The reduction in dose appeared to be independent of the dosage of Prussian Blue, and the age of the patient [94M1, 98I1]. Reductions in doses of 2 to 3 fold have also been found after other accidental intakes involving caesium isotopes [96M1, 85M1, 88T1]. Table 9.4. Committed whole body doses for individuals treated with Prussian Blue in the Goiania accident [94M1, 98I1, 00S2] Dose(1) Dose(2) [mSv] Subject Sex Age [y] Weight [kg] [mSv] Ratio(3) 2.0 360 180 17 5 F 1 1.8 220 120 20 6 F 2 1.8 210 120 26 7 M 3 2.0 90 46 23 8 M 4 1.7 240 140 25 8 M 5 1.7 250 140 27 10 M 6 2.0 350 180 31 13 M 7 1.8 1200 700 38 13 M 8 3.3 670 200 55 13 M 9 1.7 39 22 55 13 M 10 1.9 68 35 58 14 M 11 2.0 290 140 50 19 M 12 5.5 5000 910 66 23 M 13 4.0 3800 970 69 28 M 14 3.6 3100 850 66 29 F 15 3.7 1400 370 61 32 M 16 2.7 140 49 80 33 M 17 2.4 390 160 58 36 F 18 6.2 1900 300 63 41 M 19 4.0 800 200 73 43 M 20 4.8 220 46 64 46 M 21 1
)With Prussian Blue treatment, 2)Without Prussian Blue treatment, 3)Reduction in dose with treatment.
9.6.4.2 Animal data Many studies on the efficacy of Prussian Blue have been undertaken in laboratory animals [94M1, 96M1, 98I1], however in view of the detailed human studies that have been published they are not described here. Landolt-Börnstein New Series VIII/4
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In the last decade or so, work on improving the decorporation of caesium has involved investigations of other hexacyanoferrates [90D1, 93D1, 91N1]. However, at present they do not appear to be significantly more advantageous than Prussian Blue and it is difficult to see how substantial improvements can be made.
9.6.5 Plutonium and americium The currently recommended substances for enhancing biologically soluble forms of plutonium and americium are CaDTPA or ZnDTPA (see Section 9.3). 9.6.5.1 Human data Well-documented studies on the treatment of plutonium and americium after inhalation and wound contamination have been published in the scientific literature. Perhaps the most notable example on the efficacy of treatment is that often referred to as the Hanford americium accident, in which an individual sustained an intake of about 41 MBq 241Am by inhalation and wound contamination resulting from the explosion of an ion-exchange column [83B1, 89B1]. Briefly, in the absence of treatment, the bone and liver deposits were each predicted to be about 18,500 kBq, resulting in life-threatening doses of 0.07 Gy d−1 and 1 Gy d−1 respectively [89B1]. The minimum 241Am contents of these tissues after an initial intravenous adminstration of CaDTPA followed by the protracted administration of ZnDTPA by the same route, were about 220 and less than 4 kBq about 2 years after the accident. The values then increased to about 350 and 20 kBq by the 10th year as the frequency of treatment was reduced. In total 583 g of DTPA was administered. Importantly no toxic side effects were observed. The individual died 11 years after the accident from a medical condition unrelated to the accident. A summary of the tissue content and excretion of 241Am is given in Table 9.5. Table 9.5. Summary of tissue content and excretion of 241Am [89B1] Organ content [kBq] Time Skin Lungs Bone Liver Day 0 185,000 Day 3 26,000 960 480 1,400 Day 10 14,000 290 320 590 Day 60 5,500 74 250 150 1 year 1,300 74 230 150 2 years 740 55 1) 220 ND 5 years 196 ND 280 9.6 2) 7 years 190 ND NM 17 10 years 110 ND 350 19 11 years NM AS NM 23 (AS)
Cumulative excretion [kBq] Urine Faeces 4,800 0 5,000 4,700 22,000 6,800 31,000 7,000 33,000 7,000 34,000 7,000 34,000 7,000 5.4 3) 1.4 3) 3) 4.8 0.036 3) NM NM
1
) not detected at 3 years ) increase in liver content due to reduction in DTPA treatment 3 ) values per year based on 1-2 assays per year ND not detectable, NM not measured, AS measured in autopsy sample 2
A good example of the comparative efficacies of surgical treatment and the efficacy of DTPA after the intake of 239Pu oxalate through a puncture wound in the hand has also been reported [74S1]. It was estimated that the wound was contaminated with 525 kBq of 239Pu. Cleaning the wound immediately after the accident removed 144 kBq Pu. Continual monitoring of the wound with a probe showed that by 15 d, about 285 Bq remained. During this time 17 kBq had been removed with surgical dressings and 16 kBq Landolt-Börnstein New Series VIII/4
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was excreted in urine as a consequence of repeated DTPA treatment. The monitoring data implied that 37-74 kBq had been transferred to systemic tissues. A wide excision of the wound area performed in hospital 15 d after the accident removed a further 255 kBq of 239Pu from the hand. Subsequent measurements with a wound probe indicated that about 67 kBq remained at the wound site. In all there were five courses of chelation therapy which commenced 40 minutes after the accident and were spread over a period of 163 d. It was estimated that this treatment caused the elimination of 21 kBq of Pu in addition to the 0.7 kBq that would be expected in its absence. Whilst the overall efficacy of treatment with DTPA cannot be quantified due to the uncertainty in the estimate of systemic content, a reduction of 21 kBq would, based on the current ICRP model for 239Pu [93I1], imply an averted dose of 10.5 Sv [00P1]. The published information on the effectiveness of oral treatment for inhaled plutonium and americium is sparse. In general, it appears to have been useful for treating accidental intakes by workers [60N1, 67L1, 77S1]. However, uncertainties in the chemical forms of these actinides, the amounts inhaled and the delays between exposure and treatment make assessments of its potential efficacy difficult. Importantly however, the DTPA appeared to be of low toxicity; in one case 249 g of the free acid were administered over a period of 16 weeks without any apparent side effects [67L1]. There are of course many other examples of the treatment of humans after accidental exposure to plutonium and americium, published in the scientific literature. In many of these cases the administration was only partially effective [60N1, 69S1, 72J1, 73S2, 76O1, 77S1, 80P1, 80V1, 89C1, 94W1]. In part this may have been due to uncertainty in the chemical form of intake or exposure pattern, the implementation of treatment regimens which may not have been optimised, or simply that the chemical form was not amenable to treatment. Animal studies when properly executed need not suffer these disadvantages. Moreover, such studies may also be used to evaluate alternative methods of treatment or new substances. 9.6.5.2 Animal data 9.6.5.2.1 Inhaled plutonium and americium nitrate Most animal experiments have been conducted with plutonium nitrate. In these circumstances it is important that the mass concentrations of plutonium in tissues at the site of entry simulate a realistic accident scenario, say intakes corresponding to doses up to two orders of magnitude greater than the annual limit. Otherwise, the experimental data may provide information which could prejudice the use of the ligand or the mode of administration. Animal studies have shown that the administration of DTPA as an aerosol, by injection or orally in drinking water can substantially reduce the lung deposit and hence systemic deposit of plutonium and americium. Information on the efficacy of inhaled DTPA in the rat after the inhalation of 238Pu and 241Am nitrate is given in Tables 9.6 and 9.7 respectively. The tables show that concentrations of the chelate well below the usual human dosage removed nearly all the contamination from the lungs. The small amounts retained in other body tissues probably resulted from absorption and deposition in systemic tissues before the commencement of treatment. It is noteworthy that the inhalation of DTPA was almost as effective as repeated injection of the substance. These results contrast sharply with those obtained in the rat after intakes in which the mass concentrations in the lungs were about 100 times higher. Under these conditions aerosol DTPA was completely ineffective [77Bal]. Other studies, with the hamster, have shown that plutonium and americium can also be nearquantitatively removed from the lungs (i.e 1-3 % of controls) when either aerosol DTPA (2µmol kg−1) or the combined administration of aerosol (2µmol kg−1) and injected (30 µmol kg−1) was delayed for up to 11 days after exposure. However in these cases, the total body contents of plutonium and americium could be up to 30 % and 54 % of controls respectively, reflecting the difficulty in removing systemic deposits that had accumulated before the commencement of treatment [00S2].
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After inhalation of plutonium as nitrate by rats and treatment by intravenous injection, the siderophore analogue 3,4,3-LI(1,2-HOPO) is appreciably more effective than DTPA, particularly after repeated administration (Table 9.8). Under similar conditions of exposure and treatment, the ligands were considered similarly effective for americium (Table 9.9). The oral, and intraperitoneal, administration of DTPA has also been shown to be an effective method of treatment in rats after the inhalation of plutonium and americium as their nitrates (Tables 9.10 and 9.11). Importantly, a ZnDTPA concentration an order of magnitude higher than that reported here did not result in any observed histopathological changes to the liver, kidneys or gastrointestinal tract. It is also noteworthy that the higher ZnDTPA concentration did not improve its efficacy. The low toxicity of DTPA after oral administration is consistent with the data obtained from human [67L1] and other animal [80T1, 90T1] studies. The intravenous infusion and repeated injection of DTPA after the inhalation of plutonium nitrate by dogs [92G1] have also been shown to be an effective method of treatment (Table 9.12). Tables 9.6 to 9.12, see pages 9-21 and 9-22 9.6.5.2.2 Inhaled tributyl phosphate Studies on the efficacy of injected DTPA and 3,4,3-LI(1,2-HOPO) have been undertaken after inhalation by the rat. The repeated intraperitoneal injection of CaDTPA proved to be an effective treatment regimen (Table 9.13). In another study involving a higher lung deposit and a shorter period of treatment (Table 9.14), the reduction in the plutonium contents of the lungs and systemic tissues were less, but the study emphasised the higher efficacy of 3,4,3-LI(1,2-HOPO). Table 9.13. Efficacy of injected DTPA on retention of 238Pu in rats after inhalation as TBP [85S2, 00S2] % controls at 28d (ξ ± SE, N=5)(1) Treatment Lungs Total body CaDTPA(2) 4.3 ± 0.7 16 ± 2 ZnDTPA(2) 2.5 ± 0.6 15 ± 2 4.2 ± 0.8 26 ± 2 ZnDTPA(3) Initial lung deposit 384 Bq 238Pu, 0.59 ng Pu. Equivalent intake of Pu-239 by workers 0.86×104 ng or 1.98×104 Bq, i.e. 32 ALIs (CED 640 mSv). 1 )% inhaled activity in controls at 1 d, lungs 23.1 ± 1.8, total body 62.0 ± 3.0. % inhaled activity in controls at 28 d, lungs 9.5 ± 1.0, total body 43.1 ± 2.7. 2 )30 µmol kg−1 CaDTPA or 200 µmol kg−1 ZnDTPA injected i.p. at 30 min, 6 h, 1 d, 2 d, 5 d, and then every 3-4 d to 26 d. 3 )Treatment regimen as for (a) but delayed for 1 d.
Table 9.14. Efficacy of injected 3,4,3-LI(1,2HOPO) and CaDTPA on retention of 238Pu in rats after inhalation as TBP [93P1, 00S2] % controls at 7d (ξ ± SE , N=4) Treatment Lungs Skeleton iv injection(1) LIHOPO 27 ± 2 12 ± 3 CaDTPA 30 ± 1 22 ± 2 im + iv injection(2) LIHOPO 28 ± 2 2.9 ± 0.9 CaDTPA 45 ± 2 14 ± 2 1
) Initial lung deposit 5200 ± 400 Bq, 8.1 ng Pu. Equivalent intake of Pu-239 by workers 1.18×105 ng or 2.72×105 Bq, i.e. 435 ALIs (CED 8.7 Sv) Administration of 30 µmol kg−1 after 1 h Liver contents were 6.0 ± 0.7 % and 17.5 ± 3.4 % of those in controls after administration of LIHOPO and DTPA respectively. 2 ) Initial lung deposit 34000 ± 3000 Bq, 53 ng Pu. Equivalent intake of Pu-239 by workers 7.70×105 ng or 1.78×106 Bq, i.e. 2840 ALIs (56.8 Sv) Administration of 30 µmol kg−1 after 1h (iv) and 1 d and 2 d (im). Liver contents were 1.7 ± 0.2 % and 8.1 ± 2.2 % respectively of those in controls after administration of LIHOPO and DTPA respectively.
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Table 9.6. Efficacy of aerosol and injected DTPA on retention of 238Pu in rats after inhalation as nitrate [85S1, 00S2]. ξ = arithmetic mean, N = number of observations. % controls at 28 d (ξ ± SE, N=5)(1)
Table 9.7. Efficacy of aerosol and injected DTPA on retention of 241Am in rats after inhalation as nitrate [85S1, 00S2] % controls at 28d (ξ ± SE, N=5)(1) (2) Treatment Lungs Total body
Treatment(2) Aerosol(3) Aerosol plus injection(4) Injection only(5)
Aerosol(3) 2.3 ± 0.5 (4) Aerosol plus injection 1.6 ± 0.3 5.0 ± 2.1 Injection only(5)
Lungs 2.1 ± 1.1
Total body 7.6 ± 1.2
1.1 ± 0.1 4.4 ± 2.4
4.2 ± 0.7 5.7 ± 1.4
Initial lung deposit, 505 ± 37 Bq, 0.78 ng Pu. Equivalent intake of 239Pu by workers 1.13×104 ng or 2.61×104 Bq, i.e. 42 ALIs (CED 840 mSv). 1 ) % ILD in controls at 28 d; lungs, 29.3 ± 3.8, total body, 45.1 ± 5.8. 2 ) DTPA administration at 30 min, 6 h, 1 d, 2 d, 3 d, 5 d and then twice weekly to 27 d; first administration CaDTPA, then ZnDTPA. 3 ) Inhalation, 2 µmol kg–1. 4 ) Inhalation, 2 µmol kg−1and intraperitoneally (i.p.) injection, 30 µmol kg−1 5 ) i.p. injection, 30 µmol kg−1
Table 9.8. Efficacy of injected 3,4,3-LI(1,2LIHOPO) and DTPA on retention of 238Pu in rats after inhalation as nitrate [92S1, 00S2] % controls at 7d (ξ ± SE, N=4)(1) (2) Treatment Lungs Total body LIHOPO(3) DTPA
(3)
LIHOPO DTPA
(4)
(4)
LIHOPO(5)
11 ± 1
11 ± 1
16 ± 2
18 ± 2
1.8 ± 0.3
11 ± 1
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Initial lung deposit, 350 ± 25 Bq, 2.8 ng Am. Equivalent intake of 241Am by workers 4.05×104 ng or 5.09×106 Bq, i.e. 6870 ALIs (CED 137 Sv). 1 ) % ILD in controls at 28 d; lungs, 14.3 ± 1.8, total body 30.9 ± 4.1. 2-5 ) Treatment regimens as given in Table 9.7
Table 9.9. Efficacy of injected 3,4,3-LI(1.2HOPO) and DTPA on retention of 241Am in rats after inhalation as nitrate [92S1, 00S2] % controls at 7d (ξ ± SE, N=5)(1) (2) Treatment Lungs Total body LIHOPO(3) 41 ± 4 31 ± 3 DTPA(3)
21 ± 2
15 ± 1
LIHOPO (4)
13 ± 2
11 ± 2
13 ± 2
9±1
81 ± 8
64 ± 4
12 ± 1
4.5 ± 0.4
DTPA (4)
24 ± 2
27 ± 2
LIHOPO
Initial lung deposit, 600 ± 25 Bq, 0.92 ng Pu. Equivalent intake of 239Pu by workers 1.34×104 ng or 3.08×104 Bq i.e. 49 ALIs (CED 980 mSv). 1 ) % ILD in controls at 7 d: lungs, 64.6 ± 4.4, total body 86.3 ± 4.8. 2 ) DTPA, 30 µmol kg−1 administered by intraperitoneally (i.p.) injection; first administrations CaDTPA then ZnDTPA . 3 ) 30 min only. 4 ) 30 min, 6 h, 1 d ,2 d ,3 d. 5 ) 1d only.
3.7 ± 0.6 2.9 ± 0.5 3.6 ± 0.9
(5)
Initial lung deposit, 623 ± 25 Bq, 4.97 ng. Am. Equivalent intake of 241Am by workers 7.22×104 ng or 9.05×106 Bq, i.e. 12,230 ALIs (CED 244 Sv). 1 ) % ILD in controls at 7 d : lungs, 40.4 ± 3.7, total body, 71.5 ± 4.4. 2-5 ) Treatment regimen as in Table 9.8
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[Ref. p. 9-31
Table 9.10. Efficacy of oral and intraperitoneally (i.p.) injected ZnDTPA on retention of 238Pu and 241 Am in rats after inhalation as nitrate: prompt administration [95G1, 00S2] % controls at 21 d (ξ ± SE, N=4)(1) (2) Treatment Lungs Total body Plutonium Oral(3) 2.2 ± 0.4 8.8 ± 1.5 Oral(4) 2.2 0.3 7.8 ± 0.8 1.7 ± 0.3 5.2 ± 0.7 i.p(5) Americium Oral(3) 3.6 ± 0.6 6.0 ± 0.6 3.2 ± 0.3 4.8 ± 0.5 Oral(4) i.p(5) 1.7 ± 0.6 2.5 ± 0.4
Table 9.11. Efficacy of oral and intraperitoneally (i.p.) injected ZnDTPA on retention of 238Pu and 241 Am in rats after inhalation as nitrate: delayed administration [93S1, 00S2] % controls at 28 d (ξ ± SE , N=4)(1) (2) Treatment Lungs Total body Plutonium Oral(3) 6.1 ± 0.4 19 ± 3 Oral(4) 6.2 ± 0.3 17 ± 2 11 ± 1 25 ± 4 i.p(5) Americium Oral(3) 3.6 ± 0.6 23 ± 3 3.2 ± 0.3 20 ± 3 Oral(4) i.p(5) 1.7 ± 0.6 29 ± 4
Initial lung deposit, Pu 676 ± 96 Bq, 1.04 ng Pu , Am, 354 ±49 Bq , 2.82 ng Am. Equivalent intake by workers: 239Pu 1.51×104 ng or 3.49×104 Bq, i.e. 56 ALIs (CED 1120 mSv); 241Am 4.10×104 ng or 5.14×106 Bq, i.e. 6950 ALIs (CED 139 Sv). 1 ) % ILD in controls at 21 d: Pu, lungs, 41.0 ± 3.6 , total body 63.9 ± 4.7; Am, lungs, 20.1 ± 1.6, total body 51.7 ± 4. 2 ) Treatment commenced 1 h after exposure. 3 ) 950 µmol kg−1 d−1 for 21d. 4 ) 95 µmol kg−1 d−1 for 21d. 5 ) 30 µmol kg−1 twice weekly for 21 d.
Initial lung deposit, Pu 676 ± 96 Bq, 1.04 ng Pu , Am, 354 ± 49 Bq , 2.82 ng Am. Equivalent intake by workers: 239Pu 1.51×104 ng or 3.49×104 Bq, i.e. 56 ALIs (CED 1120 mSv); 241Am 4.10×104 ng or 5.14×106 Bq, i.e. 6950 ALIs (CED 139 Sv) 1 ) % ILD in controls at 28 d: Pu, lungs, 30.7 ± 2.7, total body 60.0 ± 4.9; Am, lungs, 14.8 ± 1.0, total body 51.3 ± 3.8. 2 ) Treatment commenced 7 d after exposure. 3-5 ) Treatment as given in Table 9.10.
Table 9.12. Efficacy of injected CaDTPA on retention of 238Pu in dogs after inhalation as nitrate [93G1, 00S2] % controls at 64d (ξ ± SD, N=2)(1) Treatment Lungs Total body DTPA injections(2,3) 20 ± 6 22 ± 4 DTPA infusions(4,5) 22 ± 6 17 ± 2
2
) CaDTPA i.v. (30µ mol. kg.-1 ) after 1 h, 1 d , 2 d, 3 d, 4 d, and ZnDTPA twice weekly thereafter. 3 ) Liver and bone content reduced to 8.3 ± 4.2 % and 36 ± 5 % of controls. 4 ) CaDTPA i.v. after 1 h ,then subcutaneous infusion with ZnDTPA (30 µmol kg−1 d−1 ) from 1 d. 5 ) Liver and bone content reduced to 3.8 ± 1.4 % and 28 ± 6 % of controls.
1
) Initial lung deposit, 16-26 kBq. % initial deposit in controls after 64 d , lungs 10.8 ± 2.9, liver 31.3 ± 4.6 , bone 30.5 ± 3.3 total body 76 ± 4.
9.6.5.2.3 Inhaled plutonium dioxide Under normal conditions the soluble or ultrafine component of 239Pu dioxide aerosols would be expected to be appreciably less than 1 % [72ICRP]. Hence the administration of DTPA by whatever route would have little impact on reducing the committed effective dose. However the presence of other metals during the formation of the aerosol, particularly those of low atomic weight such as sodium, can substantially increase the ultrafine component, and in such circumstances the administration of the ligand will be much more effective [80S1]. The experimental data are summarised in Table 9.15. Landolt-Börnstein New Series VIII/4
Ref. p. 9-31]
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The absorption of 238Pu into blood after the inhalation of 238PuO2 is governed by the formation of particles about 1 nm in diameter by radiolytic fragmentation. The transportable fraction arising from this process is retained in part in the lungs from where it can be mobilised by DTPA [82S1]. The experimental data summarised in Table 9.16 illustrate that DTPA was effective for removing Pu from the lungs as judged by the appreciable increase in urinary excretion. On the other hand the reduction in the body content after treatment was only about 20 % as a consequence of the competing action of mucociliary clearance. It is concluded that the protracted treatment required for a small reduction in the lung content would be unlikely to be used in humans, where for large intakes lung lavage would be more beneficial. Table 9.15. Efficacy of injected CaDTPA on retention of Pu and Am in hamsters after inhalation of a mixed aerosol of PuO2 (+AmO2) and Na2O, Pu:Na atomic ratio 1:30 [80S1] % controls at 30 d (ξ ± SE, N= 8) (1) Treatment (2)
Lungs
Total body
Pu
33 ± 6
22 ± 3 (c)
Am
40 ± 7
20 ± 3 (d)
) Initial deposit, 3.6 kBq kg−1 body mass; % ILD in controls at 30 d; Pu: lungs 27.4 ± 4.6, total body 81.2 ± 8.2, Am: lungs 21.0 ± 3.0, total body 83.0 ± 8.5 2 ) DTPA, 30 µmol kg−1 andimistered i.p. at 3 h, 1 d, 2 d, 4 d. 3 ) % controls in liver and bone, 13 ± 3 and 17 ± 4 respectively 4 ) % contols in liver and bone, 11 ± 2 and 16 ± 3 respectively 1
Table 9.16. Efficacy of DTPA on retention and excretion of 238Pu in the hamster after inhalation as 238 PuO2 [82S1] Treatment % body deposit at 7 d (ξ ± SE, N=4 - 6) (1) Lungs Systemic Urine Faeces % ILD at 154 d aerosol DTPA (2) 31.5 ± 2.7 4.0 ± 0.2 23.5 ± 1.0 41.0 ± 3.6 controls 39.2 ± 1.1 3.1 ± 0.4 2.8 ± 0.2 54.9 ± 1.4 % ILD at 147 d aerosol (2)+ i.p (3) DTPA 30.8 ± 1.4 4.1 ± 0.1 29.6 ± 0.7 35.5 ± 2.4 controls 37.7 ± 1.1 7.6 ± 0.6 5.2 ± 0.5 49.5 ± 0.8 1
) Initial body content of 238PuO2 at 7 d, 510 ± 40 Bq, of which 98.8 % was in the lungs. ) DTPA administration, 2 µmol kg−1 commenced 7 d after exposure and continued at weekly intervals to 147 d 3 ) DTPA injections, 26 µmol kg−1, commenced 10 d after exposure and continued at weekly intervals until 143 d 2
9.6.5.2.4 Inhaled americium dioxide The efficacy of DTPA for inhaled 241AmO2 has been investigated after administration as an aerosol and intraperitoneal injection in the hamster [84S1] and by intravenous injection and infusion using implanted osmotic pumps in the dog [88G1]. All methods of administration were moderately effective (Tables 9.17, 9.18). The latter treatment was particularly impressive since it virtually prevented the deposition of 241Am in systemic tissues. Implanted osmotic pumps in humans for long periods may be impracticable. However the data obtained for 241Am nitrate after the oral administration of DTPA referred to above suggest that this could be an alternative mode of administration for inhibiting systemic deposition.
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9 Decorporation of radionuclides
Table 9.17. Efficacy of injected ZnDTPA for 241 Am nitrate and 241AmO2 inhaled by the hamster [84S1, 00S2] % controls at 74 d (ξ ± SE, N=5) Treatment Lungs Total body Am nitrate(1) Injection(3) 3±1 13 ± 1 Inhalation(4) 3±1 54 ± 4 Am dioxide(2) Injection(3) 14 ± 2 16 ± 2 Inhalation(4) 27 ± 4 56 ± 4 ILD 150 Bq or 80 Bq at the commencement of treatment. 1 ) % ILD in controls at 74 d, lungs, 19.8 ± 2.6, total body, 54.4 ± 2.6. 2 ) % ILD in controls at 74 d, lungs, 25.2 ± 2.4, total body, 58.6 ± 3.1. 3 ) Zn DTPA injected intaperitoneally at weekly intervals from 4 d to 67 d at a dosage of 200 µmol kg−1. 4 ) Zn DTPA inhaled at weekly intervals from 4 d to 67 d at a dosage of 2 µmol kg−1.
[Ref. p. 9-31
Table 9.18. Efficacy of CaDTPA on retention of Am in dogs after inhalation of AmO2 [88G1, 00S2] % controls at 64 d (ξ ± SD, N=2)1 Treatment Lungs Total body DTPA injections(2,3,5) 63 ± 25 29 ± 10 DTPA 30 ± 9 11 ± 3 infusions(4,5)
241
1
) Initial lung deposit, 17-39 kBq. % initial deposit in controls after 64 d, lungs 24.5 ± 2.5, liver 25.1 ± 3.4, bone 21.8 ± 4.2, total body 76.7 ± 7.8. 2 ) CaDTPA i.v. (30 µmol kg−1 ) after 1 h, 1 d, 2 d, 3 d, 4 d, and ZnDTPA twice weekly thereafter. 3 ) Liver and bone content reduced to 4.7 ± 3.2 % and 18 ± 4 % of controls. 4 ) CaDTPA i.v. after 1 h ,then subcutaneous infusion with ZnDTPA (30 µmol kg−1 d−1 ) from 1 d. 5 ) Liver and bone content reduced to 0.43 ± 0.28 % and 1.7 ± 0.7 % of controls.
9.6.5.2.5 Wound contamination with plutonium and americium nitrate For the purpose of investigating new or alternative treatment protocols, human wounds are usually simulated in animals by subcutaneous or intramuscular injection of radionuclides. The ligands have been administered as either a single and repeated local administration or combinations of local and intraperitoneal injections, intravenous injection and oral administration. In general, local administration has proved to be most effective and intravenous injection ineffective. The comparative efficacies of 3,4,3-LI(1,2-HOPO) and DTPA in rats after the intramuscular injection of plutonium and americium nitrate are shown in Tables 9.19 and 9.20. The data show that virtually all of the plutonium and americium were removed from the body by a single local injection of 30 µmol kg−1 of 3,4,3-LI(1,2-HOPO). The retention of plutonium and americium in the body using a similar treatment protocol with DTPA were about 30 and 20 times more respectively. The efficacy of both ligands falls appreciably with delayed administration. However the retention of plutonium and americium at the wound site and in the total body is still about 3 to 4 times less with 3,4,3-LI(1,2-HOPO) than with DTPA (Tables 9.21, 9.22). The high efficacy of 3,4,3-LI(1,2-HOPO) after wound contamination has also been reported elsewhere [96V1].
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9 Decorporation of radionuclides
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Table 9.19. Efficacy of injected 3,4,3-LI(1,2HOPO) and DTPA on retention of 238Pu after intramuscular injection as nitrate: prompt treatment [93S1, 00S2] % controls at 7 d (ξ ± SE, N=4)(1) Treatment Wound site Total body LIHOPO(2) 4.8 ± 0.4 5.9 ± 0.5 LIHOPO(3) 0.9 ± 0.1 0.9 ± 0.1 0.6 ± 0.1 1.0 ± 0.1 LIHOPO(4) DTPA(5) 33 ± 2 32 ± 1 LIHOPO (6) 33 ± 3 33 ± 2
Table 9.20. Efficacy of injected 3,4,3-LI(1,2HOPO) and DTPA on retention of 241Am after intramuscular injection: prompt treatment [93S1, 00S2] % controls at 7 d (ξ ± SE, N=4)(1) Treatment Wound site Total body LIHOPO(2) 8.5 ± 0.1 8.8 ± 0.6 LIHOPO(3) 0.6 ± 0.1 0.8 ± 0.1 0.4 ± 0.1 1.2 ± 0.1 LIHOPO (4) DTPA(5) 27 ± 3 22 ± 2 LIHOPO(6) 43 ± 4 39 ± 2
1
) Injected activity 190 ± 5 Bq , 0.3 ng Pu. % injected activity in controls at 7 d, wound site 70.2 ± 1.7 total body 95.7 ± 1.0. 2 ) 3 µmol kg−1 locally at 30 min. 3 ) 30 µmol kg−1 locally at 30 min. 4 ) 30 µmol kg−1 at 30 min, plus i.p. at 6 h, 1 d, 2 d and 3 d. 5 ) as (4) with CaDTPA for local injection and ZnDTPA for i.p. 6 ) 30 µmol kg−1 i.v. at 30 min.
1
Table 9.21. Efficacy of injected 3,4,3-LI(1,2HOPO) and DTPA on retention of 238Pu after intramuscular injection: delayed treatment [94G1, 00S2] % controls at 7 d (ξ ± SE, N=4)(1) Treatment Wound site Total body LIHOPO (2) 1.0 ± 0.1 1.2 ± 0.1 CaDTPA(2) 39 ± 2 31 ± 2 17 ± 1 15 ± 1 LIHOPO (3) Ca DTPA (3) 71 ± 2 76 ± 2 LIHOPO (4) 24 ± 1 23 ± 1 75 ± 2 81 ± 2 CaDTPA(4)
Table 9.22. Efficacy of injected 3,4,3-LI(1,2HOPO) and DTPA on retention of 241Am after intramuscular injection: prompt and delayed treatment [94G1, 00S2] % controls at 7 d (ξ ± SE, N=4)(1) Treatment Wound site Total body LIHOPO (2) 0.7 ± 0.1 1.0 ± 0.1 CaDTPA(2) 28 ± 2 25 ± 2 15 ± 1 13 ± 1 LIHOPO(3) CaDTPA(3) 69 ± 2 68 ± 2 LIHOPO(4) 23 ± 1 22 ± 1 72 ± 2 73 ± 2 CaDTPA(4)
1
1
) Injected activity 200 Bq , 0.3 ng Pu. % injected activity in controls at 7 d, wound site 68.4 ± 0.9, total body 97.2 ± 1.3. 2 ) 30 µmol kg−1 locally at 30 min. 3) 30 µmol kg−1 locally at 6 h 4) 30 µmol kg−1 at 1 d.
) Injected activity 200 ± 5 Bq , 1.6 ng Am. % injected activity in controls at 7 d, wound site 70.5 ± 2.6 total body 96.8 ± 2.0. 2-6 ) Treatment protocols as given in Table 9.19
) Injected activity 200 Bq, 1.6 ng Am. % injected activity in controls at 7 d, wound site 71.4 ± 1.1, total body 97.0 ± 1.0. 2-4 ) Treatment protocols as given in Table 9.21.
9.6.5.2.6 Wound contamination with plutonium tributylphosphate (TBP) The protocols used have in broad terms been similar to those used for plutonium nitrate. A summary of the data obtained after the intramuscular injection of plutonium-TBP is given in Table 9.23. They show that the removal of plutonium from the wound site and systemic tissues is considerably less than for plutonium nitrate. This is attributed in part to the greater mass of plutonium deposited, but clearly the influence of chemical form is also important. It is noteworthy that the skeletal and liver contents after the administration of 3,4,3-LI(1,2-HOPO) are 3 to 4 times less than with DTPA.
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9 Decorporation of radionuclides
Table 9.23 Efficacy of injected 3,4,3-LI(1,2HOPO) and DTPA on retention of 238Pu after intramuscular injection as TBP [95P1, 00S2] % controls at 7 d (ξ ± SD, N=6)(1) Treatment Wound site Total body Local injection LIHOPO(2) 73 ± 24 60 ± 20 CaDTPA(2) 71 ± 15 72 ± 15 Iv injection 86 ± 13 70 ± 13 LIHOPO(2) CaDTPA(2) 100 ± 22 87 ± 15
[Ref. p. 9-31
1
) % injected activity in controls at 7 d, wound site 71.1 ± 9.8 , total body 97.6 ± 8.9. 2 ) 30 µmol kg−1 CaDTPA after 30 min.
9.6.6 Thorium The recommended substance for thorium isotopes is DTPA (see Section 9.3). The authors are unaware of any definitive human data which demonstrates the efficacy of the substance. 9.6.6.1 Animal data 9.6.6.1.1 Inhalation of thorium nitrate Studies in rats [91S1] have shown that DTPA is poorly effective when the amounts of 232Th simulated acute intakes equivalent to the dose limit, even when the substance was injected at dosages of 300 and 1000 µmol kg−1 (Table 9.24). It is noteworthy that treatment is unlikely to be implemented for intakes of less than 10 times the dose limit (see Section 9.2). Compared with the human equivalent dosage of DTPA, the increase in efficacy for 232Th using 3,4,3-LI(1,2-HOPO) could only be considered marginal (Table 9.25). The efficacy of DTPA increased to a moderate extent when substantially lower mass concentrations of thorium were used, as with 228Th. (Table 9.26). However the most effective treatment to date has involved the repeated injection of 3,4,3-LI(1,2-HOPO) whereby the thorium content of the lungs was reduced to 17 % of that in controls (Table 9.24). The results obtained with 3,4,3-LI(1,2-HOPO) suggest that the ineffectiveness of DTPA is unlikely to be due to the formation of hydrolysis products in the lungs. 9.6.6.1.2 Wound contamination with thorium nitrate Simulated wound studies in rats have been undertaken mainly with the high specific activity 228Th isotope. The data given in Table 9.27 show that 7 to 8-fold reductions in the body content occur after the prompt local injection of 3,4,3-LI(1,2-HOPO) followed by repeated administration at the same dosage. Under the same conditions, the reduction after DTPA administration was about 2-fold. The table also shows that the efficacy of both ligands is reduced appreciably when treatment is delayed only by 1 day. Based on other data with plutonium and americium [00S2], it is unlikely that intravenous or intraperitonal injections alone would have any beneficial effect.
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Ref. p. 9-31]
9 Decorporation of radionuclides
Table 9.24 Efficacy of single and repeated DTPA administration on retention of 232Th in rats after inhalation as nitrate (1) [91S1] % controls at 7 d (ξ ± SE, N=4)(2) Treatment (3) CaDTPA, (4) CaDTPA(5) Ca+ ZnDTPA(6) CaDTPA (7)
Lungs 75 ± 8 66 ± 7 66 ± 9 98 ± 10
Total body 74 ± 8 65 ± 8 65 ± 8 97 ± 10
1
) initial lung deposit of 230+232Th, 586 ± 16 Bq; 6.46 ± 0.18 µg Th 2 ) % of ILD in control animals at 7 d: lungs 76.2 ± 5.2, total body 82.0 ± 5.4 3 ) chelates administered intraperitoneally 4 ) 300 µmol kg−1 administered at 30 min only 5 ) 1000 µmol kg−1 administered at 30 min only 6 ) 1000 µmol kg−1 CaDTPA administered at 30 min and 300 µmol kg−1 ZnDTPA at 1 d, 2 d and 3 d 7 ) 1000 µmol kg−1 administered at 1 d only
Table 9.26 Efficacy of injected 3,4,3-LI(1,2HOPO) and DTPA on retention of 228Th in rats after inhalation as nitrate [98S1, 00S2] % controls at 7d (ξ ± SD, N=4)(1) Treatment Lungs Total body LIHOPO(2) 36 ± 3 29 ± 2 LIHOPO(3) 17 ± 2 17 ± 2 73 ± 4 78 ± 3 DTPA(3) 1
) Initial lung deposit, 4 ng Th Equivalent intake of Th by workers, 58 mg or 4.18×104 Bq 230Th and 1.74×109 Bq 228Th i.e. 59 ALIs 230Th (CED 100 mSv) and 2×106 ALIs 228Th (2000 Sv) % ILD in controls at 7 d: lungs 50.7 ± 1.9, total body 69.9 ± 3.5. 2 ) 30 µmol kg−1 i.p. after 30 min 3 ) 30 µmol kg−1 i.p. after 30 min, 6 h, 1 d, 2 d, 3 d
9-27
Table 9.25 Efficacy of injected 3,4,3-LI(1,2HOPO) and DTPA on retention of 232Th in rats after inhalation as nitrate(1) [91S1, 98S2] % controls at 7 d Treatment (ξ ± SD, N=4)(1) Lungs Total body LIHOPO (2) 93 ± 7 87 ± 5 LIHOPO (3) 73 ± 6 69 ± 5 ZnDTPA (3) 93 ± 6 91 ± 5 1
) Initial lung deposit, 4.2 µg Th Equivalent intake of Th by workers 61 mg or 247 Bq 232 Th, i.e. 0.36 ALI for 232Th (CED 7.2 mSv) % ILD in controls at 7 d: lungs 69.9 ± 4.5, total body 78.1 ± 4.6 2 ) 30 µmol kg−1 i.p. after 30 min 3 ) 30 µmol kg−1 i.p. after 30 min, 6 h, 1 d, 2 d and 3 d
Table 9.27 Efficacy of injected 3,4,3-LI(1,2HOPO) and DTPA on retention of 228Th after intramuscular injection: prompt and delayed treatment [95S1, 00S2] % controls at 7 d (ξ ± SE, N=4)(1) Treatment Wound site Total body LIHOPO (2) 14 ± 1 20 ± 1 CaDTPA(2) 60 ± 3 65 ± 3 LIHOPO(3) 12 ± 1 15 ± 1 Ca DTPA(3) 50 ± 3 55 ± 2 LIHOPO(4) 38 ± 2 40 ± 2 CaDTPA(4) 79 ± 3 79 ± 2 1 ) Injected activity 600 Bq 228Th, 0.1 ng Th. % injected activity in controls at 7 d, wound site 64.8 ± 1.5, total body 91.5 ± 1.7. 2 ) 30 µmol kg−1 locally at 30 min 3 ) 30 µmol kg−1 locally at 30 min and then by ip injection at 6 h, 1 d, 2d and 3d after exposure. 4 ) 30 µmol kg−1 at 1 d and then by i.p. injection at 6 h, 1 d, 2 d and 3 d later.
9.6.7 Uranium At the present time no agent can be recommended for the removal of uranium from the systemic circulation. Uranium complexation by sodium bicarbonate has been proposed for reducing the systemic deposit [80N1, 84W1, 92B1]. However, this is not supported by controlled studies with laboratory animals under realistic conditions e.g. with delays between exposure and treatment of 30 min or more. Landolt-Börnstein New Series VIII/4
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The recommended dose is 250 cm3 of 1.4 % sodium bicarbonate (42 mmol) administered by slow intravenous infusion, with further infusions on subsequent days if necessary [92B1]. Since the plasma bicarbonate concentration is held more or less constant at ~25 mmol dm−3, and its turnover time is fairly rapid, it appears unlikely that the slow infusion of a further 42 mmol would lead to a sufficiently large and sustained increase in plasma bicarbonate concentration to significantly enhance the excretion of uranium. It should also be noted that alkalosis, respiratory acidosis and hypokalaemia may result from such treatment. The authors are unaware of any definitive human data which demonstrates the efficacy of bicarbonate, or indeed any other substance. 9.6.7.1 Animal data 9.6.7.1.1 Injection of uranium Other than the octoxide and dioxide, uranium compounds formed in the nuclear fuel cycle are readily absorbed into the blood after inhalation e.g. ammonium diuranate, uranyl nitrate, uranium tetrafluoride. Since the role of treatment will be to minimise nephrotoxocity in the early lung clearance phase, the administration of uranium by intravenous injection would be appropriate for evaluating the likely efficacy of various treatment regimens. Since there appears to be no substantive evidence that the administration of sodium bicarbonate is an effective method of treatment, several alternative substances have been investigated in animals. Some of these such as tiron, certain phosphonates and hydroxypyridonate derivatives have caused large reductions in the kidney and skeletal contents of uranium when administered simultaneously or within minutes [00S2, 00S3]. However under more realistic conditions when delays between exposure and treatment may be 30 minutes, or probably longer, they are poorly effective. The efficacies after immediate and delayed treatment for some selected substances are shown in Tables 9.28 and 9.29. Notably, Table 9.29 also demonstrates the poor efficacy achieved with sodium bicarbonate. 9.6.7.1.2 Wound contamination The ligand 3,4,3-LI(1,2-HOPO) has been shown to be the most effective yet tested for plutonium, americium and thorium after inhalation and wound contamination. However the data in Table 9.29 show that it is only moderately effective even when administered immediately after the uranium. Under more realistic conditions of exposure and treatment it is poorly effective. The immediate administration of the bisphosphonate EHBP, (ethane-1-hydroxy-1,1-bisphosphonate) has been shown to prevent death in animals [94U1, 00M1]. However, the prevention of death in an acutely poisoned animal is not a reliable indicator of the effect of a chelator at the much lower level of human contamination that would be expected in any likely industrial accident, further the doses of EHBP used were more than 100 times those used in the treatment of human disease. The data given in Table 9.30 demonstrates that the efficacy of EHBP again falls rapidly with time. The main interest in EHBP is that some preparations have been licensed for other medical purposes, and hence its toxicity and metabolism are well known. It is concluded that at present no effective substance is available for the treatment of internal contamination by uranium.
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Table 9.28 Efficacy of injected polyphosphonic acids on retention of uranium [92G1, 98H1, 00S2] % controls at 4 d (ξ ± SE, N=4)(1) Treatment Kidneys Total body Immediate HMDTMP(2) 8±1 31 ± 3 DTPMP(2) 8±1 32 ± 3 Delayed 30 min HMDTMP(2) 60 ± 8 81 ± 6 DTPMP(2) 70 ± 9 86 ± 6 1 ) ID 300 Bq 233U, Ligand:uranium mol ratio 1.5×104 % ID in controls, kidneys, 9.9±0.9, total body, 27.3±1.3 2 ) 300 µmol kg−1 of hexamethylenediaminetetrameyhylene-phosphonic acid or diethylenetriaminepentamethylene-phosphonic acid.
9-29
Table 9.29 Efficacy of injected 3,4,3-LI(1,2HOPO) and sodium bicarbonate on retention of uranium [95H1, 98 H1, 00S2] % controls at 24 h (ξ ± SE, N=5)(1) Femora Treatment Kidneys (Bone) Uranium im Chelate im(2) LIHOPO 23 ± 3 46 ± 5 NaHCO3 87 ± 5 67 ± 22 Uranium im, chelate ip(3) LIHOPO 54 ± 9 82 ± 13 NaHCO3 64 ± 13 114 ± 33 Uranium iv, chelate iv(4) LIHOPO 21 ± 4 61 ± 8 NaHCO3 76 ± 20 102 ± 25 ) Initial deposit, 0.84 µmol kg−1. ) Treatment immediate. % ID in controls, kidneys 18.2, femora 1.4. 3 ) Treatment after 30 min. % ID in controls, kidneys 15.5, femora 1.25. 4 ) Treatment immediate. % ID in controls, 15.3, femora 2.2. LIHOPO, 30 µmol kg–1, NaHCO3, 640 µmol kg−1 in all cases. 1 2
Table 9.30 Removal of intramuscular 233U from rats by intramuscular injection of EHBP [98H1, 00S2] % controls at 24 h(1) EHBP Kidneys Total body 5 min 24 ± 7 70 ± 9 30 min 55 ± 21 89 ± 12
1
) Mean ± standard error, n=5 U injected 0.02 µmol kg−1, 1.5 kBq kg−1, ligand to U molar ratio, 5000. % initial deposit in controls: kidneys, 14.1 ± 2.9, total body 54.6 ± 6.0 EHBP = ethane-hydroxy-1,1-bisphosphonate
9.7 Future research needs For insoluble substances deposited in the lungs or in wounds, decorporation procedures involving lung lavage and surgical excision respectively are the most appropriate. Increasing the expertise in the treatment of obstructive lung disease should increase the availability of the lavage procedure after major accidents. However, it is difficult to perceive what improvements could be made to increase the efficacy of lavage, and surgical excision, other than through good medical practice and medical-team training. Likewise, it is difficult to envisage what improvements can be made to increase the elimination of tritium and iodine isotopes from the body beyond those procedures described earlier.
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The decorporation of caesium becomes an important issue in major radiation accidents e.g. contamination of members of the public from leaking radiotherapy sources, fission product release in a nuclear accident. The commonly used decorporating agent, Prussian Blue, is of only limited efficacy and it is recommended therefore that more effective agents are developed and tested. Within the last decade in particular, considerable progress has been made in evaluating treatment regimens and developing new chelating agents for the decorporation of plutonium, americium, and high specific activity forms of thorium. Animal studies have shown that some of these ligands are much superior to DTPA, comprehensive toxicity testing has not yet been undertaken, and for practical purposes DTPA must at present remain the agent of choice for these elements. It is essential, however, that patients who receive DTPA therapy are followed up in order to establish the efficacy of the treatment regimen for the particular physico-chemical form. Little progress has been made on the decorporation of uranium. In view of the large scale and widespread use of the element, this should be viewed with some concern. It is recommended that research to find new methods for the decorporation of uranium should be expedited.
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9.8 References 60N1 64W1 65H1 67H1 67L1 67R1 67S1 67S2 68C1 68H1 69N1 68S1 68S2 68V1 68W1 69S1 69S2 71L1 71V1 71V2 72H1 72H2 72H3 72I1 72J1 72K1 72V1 73S1 74S2 73S3 76O1 77B1 77S1 77V1
Norward, W.D., in: Proceedings of the 13th International Congress on Occupational Health, New York, 1960, p. 335. Waldron-Edward, D., Paul, T.M., Skoryna, S.C.: Can. Med. Assn. J. 91 (1964) 1006. Hesp, R., Ramsbottom, B.: PG Report 686 UKAEA, 1965. Hesp, R., Ramsbottom, B.: Strontium Metabolism (Leniham, J.M.A., Loutit, J.M., Martin, J.H. eds.) Academic Press, London (1967) 311. Lagerquist, C.R., Putzier, C.A., Piltingsgrud, C.W.: Health Phys. 13 (1967) 965. Ramsden, D.,Passant, F.H.,Peabody, C.O.,Speight, R.G.: Health Phys. 13 (1967) 633. Spencer, H., Lewin, I., Samachson, J.: Int. J. Appl. Radiat. Isot. 18 (1967) 779. Spencer, H., Lewin, I., Samachson, J.: Lancet 2 (1967) 156. Carr, T.E.F., Harrison, G.E.,Humphreys, E.R., Sutton, A.: Int. J. Radiat. Biol. 14 (1968) 225. Harrison, G.E., in: Diagnosis and treatment of deposited radionuclides, Kornberg, H.A., Norwood, W.D. (eds.): Excerpta Med. (1968) 333. National Academy of Sciences: The halothane study, Washington DC: National Research Council, 1969. Smith, H., in: Diagnosis and treatment of deposited radionuclides, Kornberg, H.A., Norwood, W.D. (eds.): Excerpta Med. (1968) 372. Spencer, H., in: Diagnosis and treatment of deposited radionuclides, Kornberg, H.A., Norwood, W.D. (eds.), Excerpta Med. (1968) 489. Volf, V., in: Diagnosis and treatment of deposited radionuclides, Kornberg, H.A., Norwood, W.D. (eds.), Excerpta Med. (1968) 355. Woodard, H.Q., in: Diagnosis and treatment of deposited radionuclides, Kornberg, H.A., Norwood, W. . (eds.), Excerpta Med. (1968)361. Schofield, G.B., in: Proceedings of Symposium on Handling of Radiation Accidents, Vienna: IAEA, 1969, p. 163. Spencer, H., Lewin. I., Belcher, M.J., Samachson, J.: Int. J. Appl. Radiat. Isot. 20 (1969) 507. Lambert, B.E., Sharpe, H.B.A., Dawson, K.B.: Am. Indust. Hyg. Assoc. J. 32 (1971) 682. Vanderborght, O., Keslev, D., Van Puymbroeck, S.: Environ. Physiol. (1971) 119. Vanderborght, O., Van Puymbroeck, S., Colard, J.: Health Phys. 21 (1971) 181. Henry, P., in: Assessment of radioactive contamination in man, STI/PUB/290. Vienna: IAEA, 1972, p. 641. Humphreys, E.R., Van Puymbroeck, S., Vanderborght, O.: 2nd Int. Conf. Strontium Metabolism, Glasgow and Strontian, USAEC Conf. 72 0818, 1972, p. 309. Humphreys, E.R., Howells, G.R.: 2nd Int. Conf. Strontium Metabolism, Glasgow and Strontian, USAEC Conf. 72 0818, 1972, p. 315. International Commission on Radiological Protection: Publication 19. Oxford: Pergamon Press, 1972. Jolly, J., McClearen, H.A., Poda., G.A., Walke,W.P.: Health Phys. 23 (1972) 333. Kesley, D., Van Puymbroeck, S., Vanderborght, O.: Experientia 28 (1972) 524. Vanderborght, O.L., Colard, J., Boulenger, R.: Health Phys. 23 (1972) 240. Schofield, G.B., Lynn, J.C.: Health Phys. 24 (1973) 317. Schofield, G.B., Howells, H., Ward, F.A., Lynn, J.C., Dolphin, G.W.: Health Phys. 26 (1974) 541. Spoor, N.L.,Hursh, J.B., in: Uranium, Plutonium, Transplutonium elements, Berlin: SpringerVerlag, 1973, p. 241. Ohlenschlager, L., Schieferdecker, H., in: Diagnosis and treatment of incorporated radionuclides, Vienna: IAEA, 1976. Ballou, J.E., Dagle, G.E., McDonald, K.E., Buschbom, R.L.: Health Phys. 32 (1977) 479. Spoor, N.L.: Report R-59. Chilton: National Radiological Protection Board, 1977. Van Barneveld, A.A., Van Puymbroeck, S., Vanderborght, O.: Health Phys. 33 (1977) 533.
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9 Decorporation of radionuclides Stradling, G.N., Loveless, B.W., Ham, G.J., Smith, H.: Health Phys. 35 (1978) 229. Volf, V.: Technical Report Series No 184, Vienna: IAEA, 1978. Stather, J.W., James, A.C., Brightwell, J., Strong, J.C., Rodwell, P., in: Biological implications of radionuclides released from nuclear industries, Vienna: IAEA, 1978, p. 37. Kestens, L., Schoeters, G., Van Puymbroeck, S., Vanderborght, O.: Health Phys. 39 (1980) 805. Metivier, H., Masse, R., Rateau, G., Lafuma, J.: Health Phys. 38 (1980) 769. National Council on Radiation Protection and Measurements. Management of persons accidentally contaminated with radionuclides. Report 65. Bethesda, MD: NCRP Publications, 1980. Poda, G.A., in: The medical basis for radiation accident preparedness, Hubner, K.F., Fry, S.A. (eds.), New York: Elsevier North Holland, 1980, p. 327. Smith, H., James, A.C., Stradling, G.N., in: Pulmonary toxicology of respirable particles, Springfield, VA: National Technical Information Service, 1980, p. 558. Taylor, D.M., Volf, V.: Health Phys. 38 (1980) 147. Voelz, G.L., in: The medical basis for radiation accident preparedness, Hubner, K.F., Fry, S.A. (eds.), New York: Elsevier North Holland, 1980, p. 311. Stather, J.W., Stradling, G.N., Smith, H., Payne, S., James, A.C., Strong, J.C., Ham, S., Sumner, S., Bulman, R.A., Hodgson, A., Towndrow, C., Ellender, M.: Health Phys. 42 (1982) 520. Breitenstein, B.D.: Health Phys. 45 (1983) 855. Mewhinney, J.A., Diel, J.H.: Health Phys. 45 (1983) 39. World Health Organisation: Radiation accidents; management of overexposure. Collection No 84-03, Geneva:WHO, 1984. Stradling, G.N., Stather, J.W., Sumner, S.A., Moody, J.C., Strong, J.C.: Health Phys. 46 (1984) 1296. Ma, R., Jin, Y., Zhou, Y., in: Assessment of radioactive contamination in man, Vienna: IAEA, 1985, p. 499. Stather, J.W., Stradling, G.N., Gray, S.A., Moody, J.C., Hodgson, A.: Hum. Toxicol. 4 (1985) 573. Stradling, G.N., Stather, J.W., Sumner, S.A., Moody, J.C., Hodgson, A.: Health Phys. 49 (1985) 499. Lloyd, D.C., Edwards, A.A., Prosser, J.S., Auf der Maur, A., Etzweiler, A., Weickhardt, U., Gössi, U., Geiger, L., Noelpp, U., Rösler, H.: Radiat. Prot. Dosim. 15 (1986) 191. Breitenstein, B.D., Fry, S.A., Lushbaugh, C.C., in: The medical basis for radiation accident preparedness. New York: Elsevier Science, 1987, p. 397. Stradling, G.N., Stather, J.W., Gray, S.A., Moody, J.C., Bailey, M.R., Hodgson, A., Collier, C.G.: Hum. Toxicol. 6 (1987) 365. Guilmette, R.A., Muggenburg, B.A.: Int. J. Radiat. Biol. 53 (1988) 251. Tang Minh-hua, Gong, Yi-fen, Shen Cheng-yao, Ye Chang-quing, Wu De-chang: J. Radiol. Prot. 8 (1988) 25. Breitenstein, B.D., Palmer, H.E.: Radiat. Prot. Dosim. 26 (1989) 317. Carbaugh, E.H., Decker, W.A., Swint, M.J.: Radiat. Prot. Dosim. 26 (1989) 345. Diamond, G.L.: Radiat. Prot. Dosim. 26 (1989) 26. Durbin, P.W., White, D.L., Jeung N., Weitl, F.L., Uhlir, L.C., Jones, E.S., Bruenger, F.W., Raymond, K.N.: Health Phys. 56 (1989) 839. Leggett, R.W.: Health Phys. 57 (1989) 365. Nolibe, D., Metivier, H., Masse, R., Chretien, J.: Radiat. Prot. Dosim. 26 (1989) 337. Stradling, G.N., Stather, J.W., Gray, S.A, Moody, J.C., Hodgson, A., Collier, C.G.: Report M-162, Chilton: National Radiological Protection Board, 1989. Stradling, G.N., Stather, J.W., Gray, S.A., Moody, J.C., Ellender, M., Hodgson, A., Volf, V, Taylor, D.M., Wirth, P., Gaskin, P.W.: Int. J. Radiat. Biol. 56 (1989) 503. Dresow, B., Neilson, P., Heinrich, H.C.: Z. Naturforsch. 45 (1990) 676. Landolt-Börnstein New Series VIII/4
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Mettler, F.A., Kelsey, C.A., Ricks, R.C.: Medical management of radiation accidents. Boca Raton: CRC Press, 1990. Taylor, D.M., Volf, V.: Plzensky Lekarsky Sbornik (Pilsen Medical Report) Suppl. 62 (1990) 101. International Commission on Radiological Protection: Publication 60, Oxford: Pergamon Press, 1991. Neilson, P., Dresow, B., Fischer, R., Heinrich, J.C.: Int. J. Appl. Instrum. 18 (1991) 821. Stradling, G.N., Moody, J.C., Gray, S.A., Ellender, M., Hodgson, A.: Human Toxicol. 10 (1991) 15. Stradling, G.N., Gray, S.A., Ellender, M., Moody, J.C.,Hodgson, A., Pearce, M., Wilson, I., Burgada, R., Bailly, T., Leroux, Y., El Manouni, D., Raymond, K.N., Durbin, P.W.: Int. J. Radiat. Biol. 62 (1992) 487. Bhattacharyya, M.H., Breitenstein, B.D., Metivier, H., Muggenburg, B.A., Stradling, G.N., Volf, V.: Radiat. Prot.Dosim. 41 (1) (1992) 1. Gray, S.A., Stradling, G.N., Pearce, M., Moody, J.C., Ebetino, F.: Report M-339, Chilton: National Radiological Protection Board, 1992. Dresow, B., Neilson, P., Fischer, R., Pfau, A.A., Heinrich, H.C.: J. Toxicol. Clin. Toxicol. 31 (1993) 56. Guilmette, R.A, Muggenburg, B.A.: Int. J. Radiat. Biol. 53 (1988) 251. International Commission on Radiological Protection: Publication 67, Oxford: Elsevier Science Ltd, 1993. Moody, J.C., Davies, C.P., Stradling, G.N.: Report M-435, Chilton: National Radiological Protection Board, 1993. Poncy, J.L., Rateau, G., Burgada, R., Bailly, T., Leroux, Y., Raymond, K.N., Durbin, P.W., Masse, R.: Int. J. Radiat. Biol. 64 (1993) 431. Stradling, G.N., Gray, S.A., Ellender, M., Pearce, M., Wilson, I., Moody, J.C., Hodgson, A.: Human Toxicol 12 (1993) 233. Stradling, G.N., Gray, S.A., Moody, J.C., Pearce, M., Wilson, I., Burgada, R., Leroux, Y., Raymond, K.N., Durbin, P.W.: Int. J. Radiat. Biol. 64 (1993) 134. Ansoborlo, E., Henge-Napoli, M.-H., Donnadieu-Claraz, M., Roy, M., Pihet, P.: Radiat. Prot. Dosim. 53 (1994) 163. Gray, S.A., Stradling, G.N., Wilson, I., Moody, J.C., Burgada, R., Durbin, P.W., Raymond, K.N.: Radiat. Prot. Dosim. 53 (1994) 319. International Commission on Radiological Protection: Publication 66, Oxford: Elsevier Science Ltd, 1994. Melo, D.R.,Lipzstein, J.L., Oliveira, C.A., Bertelli, L.: Health Phys. 66 (1994) 245. Moody, J.C., Stradling, G.N., Britcher, A.R.: Radiat. Prot. Dosim. 53 (1994) 169. Stradling, G.N.: Radiat. Prot. Dosim. 53 (1994) 297. Ubios, A.M., Braun, E.M., Cabrini, R.L.: Health Phys. 66 (1994) 540. Wood, R., Britcher, A.R, McGinn, J., in: Proceedings of Regional Conference of IRPA, Portsmouth, UK, Ashford: Nuclear Technology Publishing, 1994, p. 165. Dean, M.R.: An evaluation of the use of broncopulmonary lavage in the treatment of plutonium oxide, Report NRPC 54, Greenwich: Royal Naval College, 1995. Gray, S.A., Pearce, M.J., Stradling, G.N., Wilson, I., Hodgson, A., Isaacs, K.R.: Human Toxicol. 14 (1995) 902. Henge-Napoli, M.-H., Archimbaud, M., Ansoborlo, E., Metivier, H., Gourmelon, P.: Int. J. Radiat. Biol. 68 (1995) 389. International Commission on Radiological Protection: Publication 69, Oxford: Elsevier Science, 1995. Paquet, F., Poncy, J.L., Metivier, H., Grillon, G., Fritsch, P., Burgada, R., Bailly, T., Raymond, K.N., Durbin, P.W.: Int. J. Radiat. Biol. 68 (1995) 663. Stradling, G.N., Moody, J.C.: Radioanalyt. Nucl. Chem. Articles 197 (1995) 309. Stradling, G.N., Gray, S.A., Pearce, M., Wilson, I., Moody, J.C., Burgada, R., Durbin, P.W., Raymond, K.N.: Human Toxicol. 14 (1995) 165.
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9 Decorporation of radionuclides An American National Standard - Bioassay Programs for Uranium. Report HPSN.22-1995. McLean VA: Health Phys. Soc. 1996. International Commission on Radiological Protection: Publication 71, Oxford: Elsevier Science Ltd, 1996. Official Journal of the European Communities: L159, Vol. 39, 29 June 1996, Luxembourg: Office for Official Publications of the European Community, 1996. Volf, V., Burgada, R., Raymond, K.N., Durbin, P.W.: Int. J. Radiat. Biol. 70 (1996) 109. Stradling, G.N., Hodgson, A., Fell, T.P., Rance, E.R.: Report M-801, Chilton: National Radiological Protection Board, 1997. Ansoborlo, E., Hodgson, A., Stradling, G.N., Hodgson, S., Metivier, H., Henge-Napoli, M.H., Jarvis, N.S., Birchall, A.: Radiat. Prot. Dosim. 79 (1998) 23. Bailly, T., Burgada, R.: Comptes Rend. Acad. Sci. Paris t.1, Ser. IIc (1998) 241. Durbin, P.W., Kullgren, B., Xu, J., Raymond, K.N.: Radiat. Prot. Dosim. 79 (1998) 433. Gourmelon, P.: Unpublished data. Henge-Napoli, M.-H., Ansoborlo, E., Houpert, P., Mirto, H., Paquet, F., Burgada, R., Hodgson, S.,Stradling, G.N.: Radiat. Prot. Dosim. 79 (1998) 449. International Atomic Energy Agency: Dosimetric and medical aspects of the radiological accident in Goiania in 1987. IAEA-TECDOC-1009, Vienna: IAEA, 1998. Madshus, K., Stromme, A., Bohne, F., Nigrovic, V.: Int. J. Radiat. Biol. 10 (1966) 519 Stradling, N.,Hodgson, S.A., Pearce, M.: J. Radiat. Prot. Dosim. 79 (1998) 445. Stradling, G.N.: J. Alloys Compounds. 271-273 (1998) 72. Harrison, J.R., Paile, W., Baverstock, K.: Radiation and cancer , Singapore: World Scientific, EUR 18552, 1999, p. 455. Guidelines for iodine prophylaxis following nuclear accidents, update 1999, Geneva, World Health Organization, 1999. Rote Liste: Bundesverband der Pharmazeutischen Industrie eV, Frankfurt, 2000. Geoffroy, B.,Verger, P., Le Guen, B.: Radioprotection 35 (2) (2000) 151. Henge-Napoli, M.-H., Stradling, G.N., Taylor, D.M. (eds): Radiat. Prot. Dosim. 87 (2000) 1. Martinez, A.B., Cabrini, R.L., Ubios, A.M.: Health Phys. 78 (2000) 668. Phipps, A.: NRPB, personal communication, 2000. Stradling,G.N., Henge-Napoli, M.-H.,Paquet, F., Poncy, J.-L.,Fritsch, P.,Taylor, D.M.: Radiat. Prot. Dosim. 87 (2000) 19. Stradling, G.N., Henge-Napoli, M.-H., Paquet, F., Poncy, J.-L.,Fritsch, P.,Taylor, D.M. Radiat. Prot. Dosim. 87 (2000) 29. Stradling, G.N., Taylor, D.M., Henge-Napoli, M.-H., Wood, R., Silk, T.J.: Radiat. Prot. Dosim. 87 (2000) 41. Wood, R., Sharp, C., Gourmelon, P., Le Guen, B., Stradling, G.N., Taylor, D.M, HengeNapoli, M.-H.: Radiat. Prot. Dosim. 87 (2000) 51.
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10 Measuring techniques
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10 Measuring techniques
Since human beings do not possess any sense for the detection of ionizing radiation, they must entirely rely on special instruments in order to prevent or control any harmful radiation exposure or intake of radioactivity from the outset. Therefore, reliable instruments and methods for radiation detection and measurement form the precondition for the safe handling of radiation and radioactivity in medicine, scientific research, industry, and nuclear energy production. As summarized in Table 10.1, the primary tasks of radiation protection measurements can roughly be divided into dose and activity measurements employed to prevent and control hazard to man from ionizing radiation (external exposure) or the incorporation of radioactivity (internal exposure), respectively. The purpose of this Chapter is to give a short overview of radiation detectors frequently utilized for these tasks (10.1) and of their application in practice to monitor and quantify external (10.2) and internal exposures (10.3).
10.1 Detectors for radiation protection 10.1.1 Overview and general characteristics of radiation detectors The function of a detector designed of measuring ionizing radiation is to generate a measurable response that is related to the energy deposited in the detector material or the number of particles entering into it – such as the charge produced in a gas, the intensity of visible light emitted by some solid or liquid matter, the degree of blackening of a photographic film, or the number of chromosome aberrations in a biological sample. The most common radiation effects used for radiation protection measurements are summarized in Table 10.2. In this Chapter (10.1) the emphasis is on how these detectors – and the instruments based on them – respond to radiation and how this response can be interpreted to determine the desired quantity such as dose, exposure, or activity. Clearly, radiation protection measurements are performed by means of instruments common to other fields of radiation detection and measurement (e.g. nuclear and particle physics, radiation therapy, diagnostic radiology, and nuclear medicine). Therefore, much has been published about the design and operation of these devices. An enormous amount of information is provided, among others, in the excellent textbooks by Attix [86Att], Kiefer and Maushart [72Kie], Knoll [00Kno], Leo [94Leo], Lutz [99Lut], Shani [00Sha], Tsoulfanides [95Tso], and in the comprehensive three-volume treatise edited by Attix, Roesch, and Tochilin [68Att]. These texts form the basis of the following brief review and the reader is referred to them for supplementary information and an exhaustive bibliography.
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Table 10.1. Major objectives and categorization of radiation protection measurements. Objective of External exposure Internal exposure Measurement Prevention of exposure
Control of Exposure
Area monitoring Assessment of protective measures
Personal monitoring
Determination and identification of radioactivity in air, water, and food that may be incorporated Environmental monitoring Detection of surface or skin contaminations Determination and identification of incorporated radioactivity by in vivo and excretition measurements
Table 10.2. Radiation effects frequently used for radiation protection measurements (adapted from [96Hoh]). Radiation effect Type of instrument Detector material Determina- Determina- Identification of dose tion of tion of or dose rate activity or radioparticle nuclides counting Ionization chamber Gas ++ + + Electrical Proportional counter Gas + ++ ++ Geiger-Müller counter Gas + ++ Semiconductor Crystalline ++ ++ ++ detector semiconductor Optical Scintillation counter Crystal, plastic, or liquid ++ ++ ++ Thermoluminescence Crystal ++ dosimeter Radiophotolumines- Glas + cence dosimeter Thermal Superheated drop Liquid drops + detector Chemical Film Photographic emulsion ++ Etched-track detector Plastic foil + Biological Scoring of chromoChromosomes + somal aberrations ++ : frequently employed, recommended procedure + : in specific cases usefully applicable Before the different radiation detectors mentioned in Table 10.2 are described in more detail in the sub-sequent Sections, two fundamental aspects which are relevant for all of them are outlined [68Att, 86Att]: Absolute measurement and calibration. A device is regarded as ‘absolute’ if it can be constructed and subsequently used to measure radiation without the necessity of calibrating its response in a known radiation field. Such instruments are primarily installed and operated in national standards laboratories. For radiation protection measurements it is sufficient to use devices that are properly calibrated by accredited laboratories that guarantee traceability to an official standards laboratory. Calibration can be accomplished, for example, by exposing a device (in some cases in or in front of an appropriate phantom) in a reference radiation field, and then determining a calibration factor N by which the detector reading R has Landolt-Börnstein New Series VIII/4
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to be multiplied to achieve the true value of the measurand. If the quantity to be measured is not linearly related with detector response, a calibration curve has to be established by varying the considered parameter of the radiation field over the relevant range. It is important to note, that detector response generally depends on the type and energy of the radiation (photons or particles) to be detected. Due to this reason, calibration of a radiation measuring instrument is restricted to a specified range of both the quantity to be determined and the energy of the radiation considered. Repeatability, accuracy, and uncertainty. The repeatability of an instrument is its ability to yield the same result for repeated measurements in a constant radiation field. Consequently, it can be stated in terms of the standard deviation estimated from the readings of repeated measurements. On the other hand, the accuracy of an instrument expresses the closeness of the reading to the true value of the quantity being measured and, thus, mainly depends on the correctness of its calibration. In other words: repeatability characterizes random errors due to instabilities of the instrument and the stochastic character of ionizing radiation, whereas accuracy quantifies systematic errors of the measurement process. In general, the result of a measurement is only an approximation of the true value of the measurand, and thus is complete only when accompanied by a quantitative statement of its uncertainty, i.e. an interval within which the true value lies with given probability [95ISO]. Possible sources of error in radiation dosimetry and procedures for estimating the resulting magnitude of the uncertainties in the measurement results can be found in [02ISO].
10.1.2 Gas-filled ionization detectors 10.1.2.1 Ionization and gas amplification Ionization detectors were the first electrical devices developed for radiation detection. They are still widely used for radiation protection measurements. Radiation detection is based on the collection of electrons and ions produced by ionizing particles in the detector material – preferential a gas because of the high mobility of electrons and ions in this medium. An atom can be ionized in a variety of ways, but collisions of charged particles with the atomic electrons via Coulomb interaction are the most important mechnisms. For most gases of practical interest, the average energy W spent by a charged particle to produce an electron-ion pair is between 20 and 40 eV [79ICR]. In air, for example, the value is about 36 and 34 eV for α and β particles, respectively. Consequently, a charged particle will produce a large number of electron-ion pairs if it deposits its energy completely in the gas. As shown in Fig. 10.1, a typical detector configuration consists of an anode wire inside of a cylindrical cathode. When an electric potential difference (voltage) is applied between the electrodes, electrons and ions are attracted to the electrodes and generate an electric output signal, which is passed through a series of electronic circuits for amplification, processing, and storage. There are two different ways to process the signal. In the pulse mode, the electrical signal from each event is processed individually, whereas in the current mode, the signals from individual interactions are integrated, yielding a net current signal. Depending on the voltage between the electrodes, there are three basic types of gas-filled ionization detectors taking advantage of different physical effects: the ionization chamber, the proportional chamber, and the Geiger-Müller counter. To understand the differences between these devices, consider − under the assumption of a constant flux of radiation − the behavior of a gas-filled ionization detector as the electrical voltage is increased. As illustrated in Fig. 10.2 for two different types of particles, six different regions of operation can be distinguished [00Kno, 94Leo]: At very low voltages, the electric field is insufficient to avoid recombination of the electron-ion pairs. However, recombination is lessened as the voltage is raised, giving rise to a saturation curve (region I). Usually, the saturation curve reaches a more or less distinct plateau, where the number of ions collected is nearly independent of the applied voltage. A detector operating in this region (II) is denoted as ionization chamber.
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As the voltage across the device is raised further, the number of ions collected starts to increase again, because the electric field is strong enough to accelerate electrons from the primary ionization to kinetic energies that are sufficiently high to ionize further gas molecules by collisions (secondary ionization). In the case of a cylindrical detector arrangement, the electrical field increases steeply near the anode wire and thus the amplification process results in an electron ‘avalanche’ confined to a very small length of the anode wire. Due to this reason, electron avalanches initiated by primary ionizations at different points in the gas do not interfere and thus the output of the device is – for a given voltage – directly proportional to the number of primary ionizations or, equivalently, to the energy dissipated inside the detector volume. A detector operating in this region (III) is known as a proportional counter. Increasing the applied voltage still more results in a nonlinear amplification process (region IV) since the gain is so large that spacecharge effects become relevant distorting the electrical field within the detector. If the voltage between the electrodes exceeds the region of limited proportionality, the number of ions collected becomes independent of the type and energy of the incident particles. The reason is that instead of various independent avalanches confined to small Sections along the wire (as in regions III and IV), a chain reaction of many avalanches is triggered, which spread out along the entire wire. This transitory electrical breakdown is initiated by excited gas atoms undergoing radiative de-excitation. Ultraviolet photons emitted by these processes propagate to other parts of the gas volume and initiate secondary avalanches. Detectors working in this region (V) are called Geiger-Müller counters. As the voltage is further increased, a continuous breakdown occurs (region VI). E ( r ) ~1/r to electrometer or amplifier
i −
+
Fig. 10.1 Typical arrangement of a cylindrical ionization detector operated in the current mode. (r, radial distance from the anode wire; E, transaxial electric field strength).
V
Number of ions collected per time (logarithmic scale)
β particle α particle
I
II
III
IV
V
VI
Fig. 10.2 Schematic plot showing the relation between the voltage applied to a gas-filled ionization detector and the charge collected for two different types of particles depositing different amounts of energy within the ioncollecting gas volume. Six regions of operation can be distinguished: I, recombination region; II, ionization region; III, proportional region; IV, limited proportional region; V, Geiger-Müller region; VI, discharge region.
Applied voltage
10.1.2.2 Ionization chambers Ionization chambers are frequently used for radiation protection measurements and are available in a large variety of types. They are normally used in the current and not in the pulse mode. The electrical current caused by a device operating in the ionization region, however, is much too small to be measured using conventional galvanometer techniques. Therefore, a sensitive electrometer with a sufficiently high impedance indirectly measures the electrical current in the circuit by detecting the voltage drop across a high-load series resistor (cf. Fig. 10.1 and [00Kno]). Depending on whether the charge or the current is measured, the chamber will register the total ionization that has occurred in a given time (exposure, dose) or the rate of ionization at any instant (exposure rate, dose rate). Landolt-Börnstein New Series VIII/4
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To discuss the basic concepts, we first consider free-air ionization chambers, which are in operation primarily in national standards laboratories. They form the standard against which simpler instruments are compared to yield reliable calibration factors. A typical configuration of a parallel-plate ionization chamber is shown in Fig. 10.3. An important feature of this device is that the sensitive volume, i.e. the volume common to both the collecting volume defined by the electrical field between the plates and the collimated beam (cf. Fig. 10.3), is defined electronically and not by the walls of the chamber. Free-air chambers are specifically well-suited for the measurement of radiation exposure since this quantity − according to the operational definition given in Section 4.4.2.3 − is directly related to the ionization produced by collision interactions in air by charged particles resulting from interactions of photons (or neutrons). However, exposure measurements require the detection of all the ionization produced in the selecting volume. To this end, the lateral distance between each of the electrodes and the border of the sensitive volume must exceed the range of the secondary electrons which originate in the sensitive volume. This guarantees that electrons, the path of which remains within the collecting volume (like e1 in Fig. 10.3), can produce all their ionization in this region where it will be collected and measured. But most electrons released in the sensitive volume will leave the collecting volume (like e2) and thus part of the ionization they produce will not reach the collector. This loss, however, is of no relevance if as many electrons from photon interactions elsewhere in the beam enter the sensitive volume with the same energy (like e3) as leave it. A detailed discussion of this condition, which is known as charged particle equilibrium (or in the considered case of a photon beam as electronic equilibrium), can be found in [86Att]. Charged particle equilibrium in the sensitive volume is attained when the photon flux remains constant across the chamber and when the distance from the diaphragm of the chamber to the border of the sensitive volume is greater than the maximum electron range. Since the maximum range of electrons in air increases rapidly with the energy of incident X- or γ-rays, the size of a free-air ionization chamber that can be realized in practice limits the energy of photons to be measured to about 200 keV. to electrometer Collector
Guard electrodes Diaphragm Source
Wires
e1 e2
e3
Sensitive volume
Collecting volume
X or γ rays
Fig. 10.3 Typical configuration of a parallel-plate free-air ionization chamber. The grounded guard electrodes and the electrically biased wires are used to produce a uniform electrical field within the collecting volume. The ionization measured is that produced by electrons in the collecting volume.
To avoid the use of large and cumbersome free-air ionization chambers, practical cavity ionization chambers have been developed which basically consist of a solid wall surrounding a gas-filled cavity. In this cavity an electric field is established to collect the ions produced by radiation entering the chamber (cf. Fig. 10.1). The criteria that determine the dimension and the material of a chamber depend on both the desired dosimetric quantity (exposure or absorbed dose) and the type and energy of the radiation to be measured. Cavity ionization chambers can be designed and operated either as equilibrium or Bragg-Gray chambers in which the local photon (or neutron) field or the local secondary charged-particle field is sufficiently well characterized, respectively [86Att]. Dosimetric measurements in photon and neutron fields can be performed with cavity chambers operated under charged particle equilibrium conditions. The physical basis is described by the Fano theorem which states that in an infinite medium of given atomic composition exposed to a uniform field of indirectly ionizing radiation, the field of secondary radiation is also uniform and independent of the density of the medium as well as of density variations from point to point [86Att]. That means that the ionization Landolt-Börnstein New Series VIII/4
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collected in the gas is not influenced by the density of the wall, provided that the materials of the wall and the gas are matched with regard to their atomic composition – i.e. the effective atomic number Z – and that the thickness of the wall exceeds the range of the charged secondary particles released in the wall. The dimensions of the gas-filled cavity can be adapted to the flux of the radiation field to be measured in order to generate a sufficiently high ionization. For exposure measurements, the wall should be made of air-equivalent material with an effective atomic number of air, such as some plastics or graphite with silicon additives. Although there is no material that is exactly air-equivalent, the deviations are of minor importance for the purposes of radiation protection measurements. The use of cavity chambers operated under charged particle equilibrium conditions is limited to photons with an energy of less than about 3 MeV. At higher energies it is not possible to build a chamber that meets the two central conditions, namely a sufficiently thick wall to establish electronic equilibrium and negligible attenuation of the photon beam crossing the chamber. Chambers can also be made tissue-equivalent, which means that the atomic composition of both the wall and the gas must be similar to that of tissue. Since the absorbed dose in air or tissue (at a depth corresponding to the wall thickness) is directly proportional to the ionization per unit volume of gas measured under charged particle equilibrium conditions with an air- or tissueequivalent chamber, respectively, these devices can be calibrated to directly read the desired dosimetric quantity. Dosimetric measurements can also be performed with cavity ionization chambers taking advantage of the Bragg-Gray principle, which states that the absorbed Dose Dm in a given homogeneous medium can be calculated from the ionization produced in a small gas-filled cavity suspended into this material according to the equation Dm = WS m− g J , where W is the average energy dissipated in the gas per electronion pair formed, S m− g the ratio of the average mass stopping powers of the medium and the gas for the charged particles considered, and J the ionization per unit volume of gas. This principle holds under two conditions: Firstly, the cavity must be small compared to the range of the charged particles striking so that the flux and energy spectrum of the charged particles is not disturbed. Secondly the energy lost by the charged particles in crossing the gas-filled cavity is equal to the energy deposited in the gas volume [86Att]. In practice, the gas-filled cavity is part of a cavity chamber with a wall that separates the cavity from the surrounding material. Therefore, the chamber wall should be either extremely thin to leave the charged particle field unchanged or matched to the atomic composition of the surrounding medium. Bragg-Gray cavity chambers can be used for dosimetry of charged particles, entering from outside the chamber, and of high-energy photons (>3 MeV) liberating electrons in the chamber wall. In practice, neither the idealized conditions of charged particle equilibrium nor that of the Bragg-Gray principle can be fully realized and thus a variety of corrections must be applied in order to make absolute dosimetric measurements possible. For radiation protection measurements, however, this complex task can be avoided when chambers are used that are calibrated under conditions similar to those in which the instrument will be applied and in terms of the desired dosimetric quantity (e.g., exposure, absorbed dose, ambient dose equivalent, or the corresponding dose rates). As an example, Fig. 10.4 shows a portable survey instrument with an air-equivalent chamber that can be used for dose and dose rate measurements over a wide range of photon energies.
Fig. 10.4 Dose and dose rate survey meter with an air- equivalent ionization chamber (volume 600 cm3) for the measurement of X- and γ-rays in the energy range between 6 keV and 3 MeV. Using an additional plastic shielding, even photons up to an energy of 7.5 MeV can be measured. The probe is detachable for remote measurements. (Courtesy Step Sensortechnik). Landolt-Börnstein New Series VIII/4
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- - - -
Charging electrode
+ + + + + + + +
- - - -
10-7
C Insulator
Fig. 10.5. Schematic representation of a condenser-type ionization chamber.
Other types of portable ion chambers are based on the charge integration principle illustrated in Fig. 10.5. In this case, the chamber electrodes are connected in parallel with a capacitor, which is initially charged up to establish an electric field in the chamber. When the chamber is exposed to radiation, the ionization caused in the gas-filled chamber is collected by the electrodes and the charge stored in the capacitor is reduced. The resulting drop in chamber voltage can be used as a measure of the total integrated ionization charge. In the case of a self-reading ‘pocket dosimeter’ employed for personal radiation protection measurements, a simple rugged quartz-fiber is mounted inside the ionization volume which is deflected when the device is charged. The position of the fiber, which varies as the charge on the capacitor is reduced due to ionizing radiation, is observed through a small built-in microscope that has a scale in the eyepiece (Fig. 10.6). The position against the scale can be calibrated in terms of the total radiation recorded by the pocket dosimeter. Over a longer period of time, however, leakage currents across the insulator surface can not be avoided and this limits the accuracy of the dosimeter. Progress in integrated circuit technology resulted in the development of direct ion storage (DIS) dosimeters, which combine a gas-filled ionization chamber with a non-volatile electronic charge storage element [96Kah, 00Kno, 99Wer]. A schematic representation of a DIS dosimeter is shown in Fig. 10.7. The charge is initially placed on the floating gate of a standard EEPROM (electrically erasable and programmable read-only memory) cell by injecting electrons by a tunneling process through the silicon oxide. Electrons are trapped at this gate for many years since at normal operating temperatures they have no conductive discharge path when the silicon dioxide formation is made of high-purity material (cf. Section 10.1.4). When the chamber is exposed to ionizing radiation, the ions produced in the fill gas are collected by the charged floating gate, which results in a reduction of the charge stored. Assessment of this quantity can be performed without changing the charge distribution by measuring the channel conductivity of the field effect transistor (FET), which forms the basis of the EEPROM, by means of an electronic readout unit. The passive electronic dosimeter thus makes it possible to instantly and nondestructively readout the accumulated dose whenever required. The first commercial DIS personal dosimeter consist of a series of separate dosimeter elements housed inside a small hermetically sealed container. Three elements are used for the measurement of the personal dose equivalent Hp(10) in the range from 1 µSv to 40 Sv and two elements for the determination of Hp(0.07) in the range between 10 µSv and 40 Sv [99Wer] (the definition of Hp(d) is given in Section 4.5.3.4). The dose-rate linearity is flat up to 40 Sv/h thus guaranteeing accurate dose assessment in accident situations. Field lens
Eye lens
Scale
a
Fiber
Insulator
Ionization chamber
Electrode
Bellows
Charging pin
Fig. 10.6. Pen-size direct-reading ion chamber dosimeter for personal radiation protection measurements in a γ- or X-ray environment. (a) Simplified schematic representation. (b) Pocket dosimeters for different dose and energy ranges. The nominal voltage required to ‘set the dosimeter to zero’ is adjusted with the charging unit (potentiometer). A test source is used to check the correct function and calibration of the personal dosimeters. (Courtesy Thermo Eberline ESM). Landolt-Börnstein New Series VIII/4
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Silicon oxide
Source
Fill gas
Channel
Floating gate
Drain
Silicon
[Ref. p. 10-25
Tunneling process
Fig. 10.7. Schematic representation of a DIS dosimeter consisting of combination of a gas-filled ionization chamber and an EEPROM.
10.1.2.3 Proportional counters A serious limitation of ionization chambers is that they are not sensitive enough to detect individual ionizing particles and thus can not be used for particle counting. This limitation can be overcome by operating an ionization detector in the proportional region to take advantage of the gas amplification process described in Section 10.1.2.1. However, proportional counters are not simply ionization chambers operated at high voltages (predominantly) in the pulse mode, but are specially constructed devices designed to optimize the gas amplification effect [00Kno]. The most important difference is that proportional counters always contain a thin anode wire to create a high electric field. Important geometrical factors are – among others – the uniformity, smoothness, and centricity of the thin anode wire with a diameter of between 5 µm and 50 µm. Specific demands on the gases are: low working voltage, high gas amplification, good proportionality, and high rate capability. These conditions are met by using mixtures of a noble gas and a polyatomic organic gas, such as 90 % argon and 10 % methane (P10 gas) or 96 % helium and 4 % isobutane. The organic additives, denoted as ‘quenchers’, improve the stability and performance of the counter by absorbing secondary ultraviolet photons − that are emitted from excited gas atoms − in a mode that does not lead to further ionization and thus avoids a transitory electrical breakdown. In practice, gain factors between 102 and 106 can be achieved. For particle detection and counting, proportional counters are operated in the pulse mode. Fig. 10.8 shows the equivalent circuit of a proportional counter, which replaces the circuit plotted in Fig. 10.1 for a current-type chamber. As mentioned above, the electrical field increases steeply near the central wire of a ionization detector and thus the amplification process initiated by an ionizing event in the sensitive volume results in an electron ‘avalanche’ in a region extending only a fraction of a millimeter from the anode surface. However, at the moment when the electrons are collected at the anode wire (within about 1 µs), the positive ions are still so close to the center wire that there is almost no change of the electric voltage. The output signal – a voltage pulse V(t) – is thus predominantly determined by the slower drift of the positive ions outward towards the cathode. Most of this process develops while the ions are still relative close to the wire and thus a sharply defined fast-rising electrical pulse can be observed. The subsequent decrease of the voltage pulse depends on the relative time constant of the external load circuit, which is given by the product of the resistance R and the equivalent capacitance C of both the detector and the measuring circuit (usually a preamplifier). When the capacitance of the circuit is fixed, the height of the voltage pulse is directly proportional to the charge generated within the detector and thus to the amount of energy the incident particle deposited in the gas [00Kno]. Based on this feature, a discrimination between particles depositing different amounts of energy in the gas volume – such as α- and β-particles – can be achieved. This can be realized, for example, by means of two separate read-out channels with different discriminator levels (cf. Fig. 10.8) in order to simultaneously detect either α-particles (high level) or both α- and β-particles (low level). Alternatively, the proportional counter can be connected to a multichannel analyzer, which records and stores pulses according to their height and thus allows the direct discrimination between different particles. If the channel number is related to the energy loss of the incident particles in the cavity by means of a suitable calibration procedure, a proportional counter can also be used for particle identification (particle spectroscopy).
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R
V (t)
Voltage V ( t )
Counter
Ref. p. 10-25]
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VD
Time t
Fig. 10.8. Equivalent circuit of a counter operated in the pulse mode (cf. Fig. 10.1). VD is the discriminator level. Only pulses with a pulse height exceeding this level are counted.
In practice, it is not possible to record all types and energies of radiation with sufficient efficiency by one detector only and thus different detector designs are used for α-, β-, and γ-radiation. For the detection of α- and low-energy β-particles, it is of particular importance to use counters with thin window foils (such as metalized plastic) or windowless entrance cathodes made of a metal wire grid to reduce absorption or energy-loss of the particles transversing the entrance material as far as achievable. Since such systems can be sealed only heavily, stationary counters are normally operated as gas flow units to avoid gradual contaminations and loss of fill gas. To monitor for contaminations on floors and surfaces of objects in laboratories or on hands, shoes, and clothing of personnel working with radioactive material, large-area proportional counters are used such as those shown in Fig. 10.9. They consist of a cathode container filled with gas, which encloses either a meander-shaped anode wire or multiple anode wires. Proportional counters (PC) made of a tissue equivalent (TE) plastic wall (most often A-150 plastic) and filled with TE gas mixtures (propane or methane gases mixed with carbon dioxide and nitrogen) are standard instruments in microdosimetry [83ICR]. An important feature of TEPCs, which are based on the Bragg-Gray cavity theory described in Section 10.1.2.2, is that the pressure of the filling gas can be adjusted so that a charged particle crossing the cavity deposits an identical amount of energy as a charged particle crossing a real tissue volume of microscopic dimensions [95Wak]. Multichannel pulse-height measurements with a low-pressure TEPC in radiation fields with an intensity low enough to allow the detection of single events thus give the distribution of the energy deposited by individual primary charged particles in a microscopically small tissue volume. Since the height of the recorded pulses strongly depends on the ionization density along the tracks of the charged particles − which varies considerably between different types of charged particles such as electrons, protons and heavier ions − a low pressure TEPC not only acts as recorder of deposited energy, but also as a spectrometer able to distinguish charged particles with a different linear energy transfer (LET) and thus provides an estimate on radiation quality [95Wak, 02Wak].
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Fig. 10.9. Large-area proportional counters for contamination detection. a: Sealed handheld monitor for detection of surface contaminations with α-, β- and γ-isotopes. (Courtesy Thermo Eberline ESM). b: Continuous gas-flow contamination monitor for detection of α- and β-contaminations of hands, shoes, and clothing. (Courtesy Berthold Technologies).
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Low pressure TEPCs have been an important laboratory tool in experimental microdosimetry for many years. The recent availability of microelectronics and progress in digital electronics enabled the development of portable TEPCs for area monitoring and thus the application of this microdosimetric method in practical radiation protection [89Men, 95Sch, 00Sha]. Particular advantages of these devices are the possibility (1) to separate dose fractions due to photons and neutrons in mixed n-γ fields, which release weakly respectively densely ionizing secondaries in the TE wall of the counter, and (2) to directly measure the operational quantity ambient dose equivalent [89Die]. 10.1.2.4 Geiger-Müller counters Although the design and operation of Geiger-Müller (GM) counters are in many respects similar to those of proportional counters, there are three important differences: Firstly, GM counters are operated at substantially higher tube voltages (cf. Fig. 10.2) so that a particle entering the gas volume triggers an electron avalanche extending along the entire length of the anode wire. As a consequence, the size of the output voltage pulses is more or less independent on the number of original ion pairs that initiate the gas amplification process and thus on the type and energy of particle entering the sensitive gas volume. The voltage between the electrodes that is required to sustain an avalanche ionization can be determined by exposing the counter to a constant source of radiation and observing the counting rate as a function of the applied voltage. Secondly, the UV absorbing quench gas used in proportional counters is omitted in GM tubes since ultraviolet photons emitted from excited gas atoms are essential to the process of propagating the discharge throughout the tube. Instead, other ‘quenchers’ such as gaseous halogens (Cl or Br) or organic substances (ethyl alcohol or ethyl formate) are added with a typical concentration of 5-10 % to the primary fill gas to prevent repeated or continuous gas discharge through the mechanism of charge transfer collisions. A detailed description of the underlying complex mechanism can be found in [86Att, 00Kno]. Thirdly, immediately after a discharge, a dense cloud of positive ions exists near the central wire and reduces the electric field in the counter to a great extent. This space charge not only terminates the discharge of the tube but also prevents that a further avalanche can be generated before the positive ions have moved − at least part of the distance − towards the cathode. The time between the detection of the initial pulse and the time at which a succeeding pulse can be counted because its amplitude exceeds the discriminator level is denoted as resolving time. Typical values are between 100 to 300 µs. In contrast, the resolving time of a proportional counter is less than a few microseconds. During the resolving time, the GM counter is ‘dead’ and any particles entering the tube during that time are lost [94ICR1]. In practice, gain factors between 106 and 1010 can be realized with a GM counter. The resulting voltage pulses have a height between 1 and 10 V and can easily be detected with simple electronic circuits − often completely without external amplification. GM tubes are thus simple, rugged, and relatively inexpensive particle-counting instruments. As mentioned above, however, they suffer from extremely long resolving times and are thus seldom used when accurate measurements are required at count rates greater than a few hundred counts per second. In many cases, GM counters are provided with removable covers on the entrance window in order to differentiate between penetrating (γ- and high-energy β-particles) and low-penetrating (α- and low-energy β-particles) radiation by measuring the difference between the count rates with and without the cover in place. As in the case of proportional counters, the entrance window must be sufficiently thin to permit passage of α-particles. For the detection of γ-particles, on the other hand, the thickness of the entrance window or of the cover should approximate the maximum range of the secondary electrons produced in the window or cover to increase detection sensitivity.
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10.1.3 Scintillation detectors The major limitation of gas-filled counters, namely the low detection efficiency for X- or γ-rays, can be overcome by the use of solid or liquid detector materials, which have a much higher density than gases. Scintillation materials are frequently utilized for photon or neutron detection. In these materials a small fraction of the energy deposited by charged particles will be emitted as visible or ultraviolet light on a time scale of nanoseconds to milliseconds, whereby the intensity of the light flash is proportional to the energy deposited in the scintillator. As shown in Fig. 10.10, a scintillation detector consists basically of scintillator material that is optically coupled to a photomultiplier tube to convert the light photons released in the scintillator into an electrical pulse which can then be amplified and analyzed electronically. Scintillator
γ − particle Analyzer
a
Semitransparent photocathode
Light shield, reflector Photomultiplier
gas envelope
HV
Shield e−
Amplifier
Pre − amplifier
Light photon
Dynodes
Anode
b
Fig. 10.10. Schematic representation of (a) a pulse-mode scintillation detector and (b) a linear focussed photomultiplier tube showing the cascade effect due to amplification of electrons from the photocathode by increasing secondary emission when the electrons strike the dynodes.
Scintillators fall into two major categories: inorganic and organic materials, the choice of which depends strongly on the type of measurement to be performed. Physical properties of a few representative scintillator materials are given in Table 10.3; a more comprehensive list can be found in [00Kno, 94Leo]. Inorganic scintillators are crystals of alkali halides (such as NaI, CsI) or oxides (such as Bi4Ge3O12, ‘BGO’) grown at high temperatures. In these materials, scintillation is a property of the electronic band structure of the crystals: When an ionizing particle enters the scintillator, it can raise electrons from the valence into the conduction band. The electrons and holes formed by this excitation process recombine and emit a photon. In the pure scintillator material, however, de-excitation is an inefficient process due to self-absorption. Therefore, small amounts of an ‘activator’ (e.g., thallium in the case of NaI) are added. These impurities create energy states within the forbidden band gap of the scintillator over which electrons can alternatively de-excite from the conduction band into the valence band. A more detailed description of the scintillation process can be found in [00Kno, 95Tso]. Since energy spacing between activator energy states is less than that between the conduction and valence bands of the pure solid, the emitted photons do not have enough energy to raise other electrons from the valence band to the conduction band and thus cannot be effectively reabsorbed by the scintillator. Moreover, the change in energy of the emitted photons results in a shift of the wavelength of maximum emission from the ultraviolet into the visible region, where the sensitivity of most photomultiplier tubes is maximal. Inorganic scintillators tend to contain elements with a high atomic number and have a relatively high density (cf. Table 10.3). Consequently, the photoelectric effect is the main interaction mechanism for X- or γ-rays in the energy range between 10 keV to 1 MeV, making inorganic scintillators favorable for particle identification by means of spectroscopic measurements (see below). They also have a high light output, but are hampered by a relatively slow response. Organic scintillators, on the other hand, are aromatic hydrocarbon compounds which contain benzenoid rings. They are broadly classed into three types: crystalline, liquid, and plastic, all of which utilize the ionization produced by charged particles to generate optical photons, usually in the blue to green wavelength regions. Examples of pure organic crystals are anthracene (C14H10) and trans-stilbene (C14H12). Plastic scintillators are non-fluid solutions consisting of fluorescent organic compounds disLandolt-Börnstein New Series VIII/4
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solved in a solidified polymer matrix or fluid solutions with similar fluorescent organic compounds [02Sai]. In contrast to inorganic scintillators, the fluorescence process in organic scintillators is an inherent molecular property which is characterized by the excited states of the individual molecules. Therefore, fluorescence can be observed independent of the physical state of the material. Organic scintillators are generally faster in their response than inorganics and are more suitable for β-particle spectroscopy and especially for fast neutron detection due to the high hydrogen fraction in their composition (cf. Section 10.1.7). Moreover, some organic scintillator materials (such as BC 501 / NE 213, cf. Tab. 10.3) offer the possibility to discriminate between photons and neutrons due to differences in scintillator response. In these materials, electrons released by γ-quanta cause scintillations at a rate faster than that due to photons created by neutrons. Table 10.3. Physical properties of a few representative scintillator materials (data from [02Sai]). Material Type Density RefracDecay Light Wave- Main application tion outputa constantb lengthc index [g/cm3] [%] [ns] [nm] Inorganic scintillators NaI(Tl) Crystal 3.67 1.85 100 250 415 γ, X-rays CsI(Tl) Crystal 4.51 1.79 45 1005 550 γ, heavy particles CsI(Na) Crystal 4.51 1.84 85 630 420 γ, heavy particles BGO Crystal 7.13 2.15 20 300 480 γ, X-rays LSO Crystal 7.40 1.82 63 40 420 γ, X-rays Organic scintillators Anthracene Crystal 1.25 1.62 100 30 447 General purpose Trans-stilbene Crystal 1.16 1.63 50 4.5 410 γ, fast n BC 400 / NE 102 Plastic 1.03 1.58 65 2.4 423 General purpose BC 422 / NE 111 Plastic 1.03 1.58 55 1.6 370 Ultra-fast timing BC 501 / NE 213 Liquid 0.87 1.51 78 3.2 425 Fast n with γ discrimination BC 509 / NE 226 Liquid 1.61 1.38 20 3.1 435 γ, insensitive to n a
Given relative to NaI(Tl) for inorganic scintillators and relative to anthracene for organic scintillators. The light output of anthracene is 40-50 % of NaI(Tl). b Main component. c Maximum emission.
As mentioned above, photomultiplier tubes (PMT) have two different functions: conversion of ultraviolet and visible photons emitted by the scintillator into an electrical signal and signal amplification. Fig. 10.10 shows the essential parts of a PMT, which are mounted inside an evacuated glass envelope, namely a photocathode, typically 10 to 12 electrodes denoted as dynodes, and an anode. The photons from the scintillator strike the photocathode − usually made of a semiconductor material formed from antimony plus one or more alkali metals − and release photoelectrons with an efficiency of about 10-30 % [94Leo]. These photoelectrons are attracted to the first dynode, which is at a higher potential than the cathode, so that the electrons strike the dynode with a sufficiently high kinetic energy to eject three to four secondary electrons from the surface. Since each dynode has a more positive voltage than the preceding one, this amplification process is repeated with each successive dynode, so that a multiplication factor of 106 or more can be obtained with a twelve stage PMT. The average gain of the dynode chain is independent of how many electrons are simultaneously ejected from the photocathode. As a consequence, the size of the electrical output signal at the anode is proportional to the number of electrons leaving the photocathode. To achieve a good performance, it is important to match the emission spectrum of the scintillator to the quantum efficiency of the photocathode material. The current measured at the anode of the PMT is fed into an RC circuit as shown in Fig. 10.8 to produce an electrical voltage pulse.
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The previous discussion reveals, that the output of a scintillator-PMT combination is proportional to the amount of energy deposited by an incident particle in the scintillator and can thus be used for particle spectroscopy. To this end, the voltage pulse goes into an amplifier and is then fed into a multichannel analyzer, which records and stores pulses according to their voltage amplitude into different channels. By using γ-ray sources of known energy, the channel numbers can be related to energy. As an example, Fig. 10.11 shows the pulse-height spectrum of 60Co recorded from a large NaI(Tl) scintillator. 100
100 Photopeaks Photopeaks Backscatter
60
Compton region
40
Relative count rate [% ]
Relative count rate [% ]
80
80 Compton edges
Backscatter 3 2
1 3
20
0
0 0
200
a
400
600 800 1000 1200 1400 1600 Energy [ keV ]
0
b
200
400
600 800 1000 1200 1400 1600 Energy [ keV ]
Fig. 10.11. Pulse-height spectra of 60Co measured with (a) a large NaI(Tl) scintillation detector and (b) a highpurity germanium detector. The two characteristic photopeaks are at energies of 1.17 and 1.33 MeV. In (b) a single (1) and a double escape peak (2) related to the incident 1.33-MeV-photons is apparent, as is an annihilation peak at 511 keV (3) due to pair production interactions in surrounding materials. Note that incident 1.17-MeVphotons do not give rise to escape peaks, since the photon energy is only slightly above the threshold for pair production interactions where the cross section is still very low.
A pulse-height spectrum recorded from a radiation source depends not only on the characteristics of the radiation to be measured but also on the type of scintillator used and the mechanisms by which the incident particles transfer their energy to the detector material. Since the ranges of charged particles are very short in most solid and liquid materials, they deposit their energy almost completely in the detector material giving rise to a well-defined peak in the spectrum at the particle energy. In the case of γ-quanta, energy is deposited to the detector primarily by the photoelectric effect, Compton scattering, and pair production. An incident photon undergoing a photoelectric interaction in the scintillator transfers (nearly) all of its energy to an electron and thus contributes to the ‘photopeak’ in the pulse-height spectrum, which is located at the energy of the incident photon. In Compton scattering, however, only part of the energy is transferred to the detector, via the recoil electrons. The scattered photon may either be absorbed by a photoelectric interaction within the scintillator or may escape from the detector. In the first case, the total energy of the incident photon is absorbed and the event will contribute to the photopeak. In the second case, however, the energy deposited by the recoil electron depends on the scattering angle. The ‘Compton region’ in the spectrum thus ranges from near zero (small-angel scattering) up to a maximum energy (‘Compton edge’) for 180° Compton scattering. If the energy of the incident photon exceeds 1022 keV, pair production can occur. When the positron created by this process comes to rest, it combines with an electron to create a pair of 511 keV annihilation photons. If one or both of these photons escape, the energy deposited in the scintillator is reduced by 511 or 1022 keV, respectively. As a consequence, additional ‘photopeaks’ − denoted as single and double escape peaks − appear in the spectrum at energies of 511 keV and of 1022 keV below the corresponding full-energy photopeaks (cf. Fig. 10.11). Finally, lowenergy peaks may appear in the spectrum resulting from γ-quanta that are scattered in material outside of the scintillator, and enter the detector having lost most of their energy. However, these backscatter peaks Landolt-Börnstein New Series VIII/4
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area usually easy to identify due to their low energies (<250 keV). Since both the energy and relative emission probability of γ-rays released by a radioactive decay is a characteristic property of nuclides, γ-spectroscopy is widely used for the identification of nuclides in a sample or in the environment [94ICR2]. A discussion of the various aspects related to particle spectroscopy can be found in [00Kno, 95Tso]. Scintillation detectors not only have a markedly higher detection efficiency for γ- or X-rays than gasfilled ionization detectors due to the higher density of the detector material but also a much shorter resolving time which allows them to respond more linearly to higher count rates than Geiger-Müller counters. Therefore, scintillation counters are frequently applied for particle counting and identification. As an example, Fig. 10.12 shows a portable NaI spectrometer for the localization, identification, and measurement of γ-rays. A major limitation of scintillation detectors, on the other hand, is their poor energy resolution. The reason is that an average energy loss of 100 eV or more in the scintillator material is required to release one photoelectron from the photocathode of the PMT and that, consequently, the average number N of photoelectrons – or information carriers – produced by an incident ionizing particle is no more than a few thousands. Due to the random nature of the interaction processes, the standard deviation characterizing the statistical fluctuations in that number is proportional to N and the relative uncertainty proportional to 1 N . Therefore, there is a significant inherent limitation on the energy resolution of scintillation counters [00Kno]. For example, scintillation detectors used in γ-spectroscopy typically show an energy resolution in the range of 5 - 10 % (cf. Fig. 10.11a). The intrinsic statistical limit on energy resolution can only be reduced by increasing the number of information carriers created per unit energy lost by the incident radiation – as, for example, by the use of semiconductor materials.
Fig. 10.12. Portable NaI spectrometer with 496 channels for the localization, identification, and measurement of γ-rays with an energy between 25 keV and 2 MeV. (Courtesy Berthold Technologies).
10.1.4 Semiconductor detectors Semiconductor detectors have experienced a rapid development in the last decades. They are basically solid-state analogs of gas-filled ionization detectors, in which electron-hole pairs are created by incident ionizing radiation instead of electron-ion pairs. Semiconductors are made of crystalline materials whose electrical conductivities are midway between those of conductors and insulators. Their electrical properties are characterized by their crystal structure: According to quantum theory, the energy of an electron in the crystal must fall within well-defined bands − the valence or the conduction band, which are separated by a forbidden energy gap. The most common semiconductor materials used are silicon and germanium which crystallize in the diamond structure (Table 10.4). In this structure, the four valence electrons form covalent bonds with each of the four nearest neighbor atoms in the crystal and are thus – at least at very low temperatures – immobile. The Landolt-Börnstein New Series VIII/4
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energy gap between the valence and conduction band is, however, much smaller in semiconductors (about 1 eV) than in insulators (more than 5 eV) and thus ionizing radiation, light, or heat can easily break covalent bonds and raise electrons into the conduction band, leaving behind electron vacancies or ‘holes’ in the valence band. When an electric field is applied across a semiconductor crystal, ‘free’ electrons in the conduction band and positive ‘holes’ in the valence band, which act as positive charge carriers, move towards the positive and negative terminal, respectively, and establish a small current when the terminals are connected to a detection circuit. The problem is, however, that at non-zero temperatures electron-hole pairs will be thermally generated in the semiconductor material and that the random fluctuations that unavoidably occur in the resulting steady-state leakage current are too high, even in the highest resistivity materials available, to permit the detection of the minute current caused by electron-hole pairs initiated by an ionizing particle. This problem can be solved by using the favorable properties of a p-n semiconductor junction, which acts as ‘blocking’ contact with an extremely high resistance and thus reduces the leakage current through the bulk of the semiconductor material. Table 10.4. Physical properties of silicon and germanium (data from [95Tso] and [99Lut]). Property Silicon Germanium Atomic number 14 32 Atomic weight 28.1 72.6 3 2.33 5.32 Density at 300 K [g/cm ] 1.11 0.67 Energy gap at 300 K [eV] 500 Intrinsic resistivity at 77 K [Ω m] ∞ 0.47 Intrinsic resistivity at 300 K [Ω m] 2300 3.7 2.96 Average energy per electron-hole pair at 77 K [eV] 3.65 Average energy per electron-hole pair at 300 K [eV] − To this end, a pure semiconductor is doped on one side with pentavalent impurity atoms (for example, phosphorus, antimony, or lithium) and on the other side with trivalent impurity atoms (for example, boron, gallium, or indium). When present in small concentrations, the impurity atoms will take the place of a tetravalent normal silicon or germanium atom in the lattice and introduce only lightly bounded excess electrons or additional electron vacancies (holes) in the crystal lattice, respectively. A semiconductor material containing an electron-donor impurity is denoted as n-type material, material doped with a holeforming impurity as p-type material. As already mentioned, the essential part of a semiconductor detector or diode is the region in the vicinity of the interface between the p-type and n-type material, which is denoted as p-n junction. Due to the difference in the concentration of electrons and holes between the two materials, electrons diffuse into the p-region and holes into the n-region. As a consequence, the diffusing electrons fill up holes in the p-region while the diffusing holes capture electrons on the n-side, leaving a region completely depleted of mobile charge carriers, as schematically shown in Fig. 10.13. Since p- and n-type materials are originally electrically neutral, the diffusion process creates a net negative and positive space charge on the p-side and n-side of the junction, respectively. At equilibrium, the electric diffusion potential (about 0.5 - 1 V) across the p-n-junction results in the transport of charge carriers in the opposite direction which precisely balances the diffusion process of charge carriers and limits the extension of the depletion region to rather small depths (about 50 to 100 µm). However, the region where the electric field exist – and thus the depletion depth – can be increased by applying a strong reverse-bias voltage across the semiconductor, as illustrated in Fig. 10.13. It is important to note, that thermal generation of electron-hole pairs in the depletion region does not result in a considerable steady-state concentration of carriers because removal of these carriers is a much faster process than their creation. Therefore, the small concentration of carriers created by ionizing particles can easily be detected above the highly suppressed concentration of thermally generated carriers [00Kno].
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Pure semiconductor crystals, referred to as wafers, are grown from the melt of high-purity polycrystalline silicon or germanium material with the help of a seed crystal of defined orientation. A p-n junction detector is usually produced by depositing sufficient p-type impurities into one end of a uniformly n-doped wafer so as to change that end into a p-type material. This can be done either by diffusion or implantation of dopants, or by a combination of both processes. A more detailed description of the fabrication and design of semiconductor detectors and of their application is presented in [99Lut]. Depletion region
- -- -- -- -- - - p − material
+ + + + +
+ + +
+
+ + + n − material
+
-
electron hole negative ion
+ positive ion
Fig. 10.13. Schematic representation of a p-n detector or diode. At the junction, electrons diffuse into the p-region and fill up holes while holes diffuse into the n-region and capture electrons. This process results in a region completely depleted of mobile charge carriers (grey area). By applying a reverse-bias voltage, the depth of the depletion region is increased.
For particle spectroscopy, semiconductor detectors are operated in the pulse counting mode. As in the case of a scintillation detector, the charge created in the detector material is transformed by means of a charge sensitive preamplifier stage into a voltage pulse, the height of which is proportional to the energy deposited by an incident ionizing particle in the depletion region. The energy resolution of a semiconductor detector, however, is much better than that of a scintillation counter, since the average energy required to produce an electron-hole pair in a semiconductor material (about 3 eV, cf. Table 10.4) is much smaller than the average energy required to release a photoelectron from the photocathode of the PMT (more than 100 eV). The superior energy resolution is demonstrated by the comparative pulse-height spectra of 60Co plotted in Fig. 10.11. Whereas silicon is the most widely used semiconductor material for charged particle detection, germanium is the preferred material for γ-spectroscopy because of its much higher atomic number (cf. Table 10.4) and thus its greater photoelectric cross section. In contrast to silicon, however, germanium must be operated at cryostatic temperatures because of its relatively narrow energy gap between the valence and conduction band. A principal drawback associated with all semiconductor detectors is the degradation in performance which can be brought about by radiation damage. At low energies, the efficiency of semiconductors for γ-detection is a function of photon cross-section and window thickness, whereas at higher energies the total active detector volume becomes the most important factor. Germanium detectors can be fabricated in many different geometries thus offering devices that can be tailored to the specific needs of the measurement [cf. 95Tso]. With the exception of the well-type configuration, the efficiency of germanium detectors is low relative to Na(Tl) scintillators. This is, however, more than compensated for by the better energy resolution. As representative examples, Fig. 10.14 shows detector efficiency and energy resolution curves for various types of germanium detectors. To take full advantage of their intrinsic energy response, detectors with thin contacts − such as low-energy germanium (LEGe) and reverse electrode germanium (REGe) detectors − are usually equipped with a beryllium cryostat window. In the last decade, compound semiconductors have gained increasing interest as detector materials applicable to γ-spectroscopy [00Kno]. Particularly cadmium zinc telluride (CdZnTe) based detectors have been developed intensively and have recently seen significant improvements [01Tak]. These devices offer some major advantages: operation at room-temperature due to the wide band gap of the material, a resolution that is intermediate between that of scintillation detectors and germanium devices, and a high density of the crystal providing excellent stopping power over an energy range of a few keV to over 1 MeV. Based on these features, portable γ-spectrometers have been developed for radiation protection measurements.
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5
2
1
Efficiency [%] 10
−2
4
3
1
−1
1 2 3 4 5
10-17
3
Resolution (FWHM) [keV ]
10
10
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REGe, 15% rel.efficiency LEGe, 10 cm2 × 15 mm thick LEGe, 200 mm2 × 10 mm thick Coaxial, 10% rel.efficiency BEGe, 5000 mm2 × 30 mm thick
Well 1
Coaxial REGe large LEGe LEGe small LEGe
0.1
5
a
10
20
50 100 200 Energy [keV ]
500 1000 2000
5.9 10
b
100 122 Energy [keV ]
1322
Fig. 10.14. Typical performance parameters for various types of germanium detectors frequently used for γ-spectroscopy. (a) Absolute detector efficiency as a function of energy compared to that of a 37 × 37 mm2 Na(Tl) crystal at a detector to source distance of 25 cm. (b) Energy resolution (full-width at half-maximum, FWHM) as a function of photon energy. Well: well-type Ge detector, REGe: reverse electrode Ge detector, LEGe: low-energy Ge detector, BEGe: broad energy Ge detector, Coaxial: coaxial Ge detector. (Figures adapted with permission from Canberra Industries).
Fig. 10.15. Direct-reading electronic personal dose and dose rate meter for measurements of X- and γ-rays over the energy range from 60 keV to 3 MeV with an accuracy of the dose reading of ± 15 % between 1 µSv and 10 Sv. The detector system, which utilizes an energy compensated silicon diode, is calibrated to give directly the personal dose equivalent Hp(10). (Courtesy Rados Technology).
With recent improvements in their applicability and reliability, silicon (PIN) diodes have also become popular as radiation detectors in electronic pocket dosimeters (cf. Fig. 10.15). To measure the absorbed dose, it is more appropriate to operate these devices in the current mode, since the current measured with a silicon diode is nearly proportional to the absorbed dose rate in soft tissue for photon energies between 150 keV to well over 1 MeV. The reason is that the mass absorption coefficients for silicon is within 10 % of soft tissue values over the given range of photon energies and thus conversion of photons into energetic electrons is similar in both cases [00Kno]. At lower energies, however, photon absorption in silicon deviates considerably from that in tissue. It is thus necessary to compensate for this effect by employing metallic filters around the detector. There are different approaches to realize energy compensation: One approach, which is designed to measure both γ-quanta and β-particles, is to use three diodes in parallel with individual filters to produce an appropriate energy response that is approximately energy-independent down to about 17 keV [95Hir]. A more cost effective solution is to use a single diode with a simple filter, usually tin, to flatten the energy response – with the major disadvantage of giving up response to photons with an energy below about 60 keV. By using a composite filter of two or more filters together with several openings, however, it is not only possible to compensate the energy response of a silicon diode but also to maintain an extended low energy response [96Ols]. As compared to passive personal monitoring devices such as photographic films or thermoluminescence dosimeters, described in the subsequent sections, electronic personal dosimeters offer various advantages, such as real-time Landolt-Börnstein New Series VIII/4
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measurement and display of the dose rate and cumulative dose, audible warning of radiation fields at user settable alarm levels, and the possibility to transmit data for remote readout and evaluation by dosimetry management system.
10.1.5 Thermoluminescence and radiophotoluminescence detectors The use of thermoluminescence detectors (TLDs) is another common method of solid-state dosimetry based on the electrical properties of crystals. In these crystals, too, electrons are raised by ionizing radiation from the valence into the conduction band. In contrast to scintillation materials, however, the electrons and holes formed by the excitation process do not recombine immediately but are caught in ‘traps’ for long periods of time at room temperature, as shown in Fig. 10.16. Whereas in some natural materials traps are formed by lattice imperfections and impurities inherent to the crystal, small concentrations of impurity (e.g., Mn, Ti, Tm, Mg, Dy) must be added in others, which function as an activator. Some TL materials commercially available are CaSO4:Mn, CaF2, CaF2:Mn, CaF2:Dy; Li2B4O7:Mn, LiF, LiF:Mg, Ti, and LiF:Mg, Cu, P (activators are given after the colon). The choice among TL materials depends on different factors: (1) the energy depth of the traps, which determines the number of charge carriers trapped per unit of absorbed dose and thus the sensitivity of the TL detector, (2) the retention of the trapped carriers for longer periods of time at normal temperatures, and (3) linear response over a large dose range. 100
Electron trap
Hole trap
Incident particle
Heating
Recombination with light emission
Valence band
Fig. 10.16. Energy-level diagram of a TLD crystal. Left: Radiation induced formation of an electron-hole pair leading to the population of an electron and a hole trap. Right: Release of a trapped electron by heat and subsequent recombination with a hole resulting in the emission of a photon. It is assumed that the electron is liberated first since the depth of the electron trap is less than that of the hole trap. In the reverse case, the hole would be thermally released first. The dashed lines represent drift of charge carriers in the valance and conduction band.
Relative thermoluminescence intensity [%]
Excitation by radiation
Conduction band
LiF:Mg, Ti CaF2 :Dy 80
60
40
20
0
5
10 Time [s ]
15
20
Fig. 10.17. Representative thermoluminescence glow curves of LiF:Mg, Ti (TLD-100) and CaF2:Dy heated from 50 °C to 325 °C within 20 s. The curves are normalized to the same maximum intensity.
The charge carriers stored in a TLD after having been exposed to radiation can be released by heating the detector, e.g., by a stream of heated gas, by laser light, or by heating the support. As schematically shown in Fig. 10.16, this process provides sufficient thermal energy to electrons or holes so that they are raised back into the valence or conduction band, respectively. ‘Free’ electrons (or holes) can migrate towards the position of a trapped hole (or electron) and recombine with the emission of light. The photons emitted by the heating process are detected by a photomultiplier tube (PMT) as in the case of a scintillation detector (cf. Section 10.1.3). The PMT output signal, which is proportional to the number of photons released in the TL material, is detected and plotted as a function of temperature or time during heating Landolt-Börnstein New Series VIII/4
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with a constant rate. The resulting plot is denoted as ’glow’ curve, which may consist of various resolved or unresolved glow peaks if there are several types of traps in the TLD material – each of which is characterized by a specific depth or binding energy (Fig. 10.17). The area under either the entire curve or individual peaks is a measure of the number of trapped charge carriers and thus of the amount of energy absorbed in the crystal. Determination and analysis of glow curves are performed by dedicated devices denoted as TLD readers (Fig. 10.18). Since the intrinsic TL efficiency, i.e. the ratio between the light energy emitted per unit of detector mass and the absorbed dose, is in the order of 1 % only, TLDs must be used under reproducible conditions to obtain consistent results. One important aspect is to thoroughly deplete all traps in the detector before it is reused for dose measurements. To this end, the temperature of TLDs must be raised to relatively high values over longer periods of time in an annealing oven (cf. Fig. 10.18). TLD dosimetry is a relative measurement technique and thus TLDs have to be calibrated individually against absolute dosimetry systems such as a calibrated ion chamber. Details on TLD dosimetry and calibration are presented in [86Att, 00Sha]. TLDs are available in diverse forms − such as loose powder, chips, rods, rings, or small wafers − and are widely used as single element dosimeters or assemblies for a large variety of radiation protection measurements (Fig. 10.18). They have found, for example, an important place in personal monitoring and are rapidly replacing the use of photographic films discussed in the next section. In personal monitoring devices, one or more small TLD elements are usually assembled into rigid aluminum cards and mounted within shielded filter-holders. For this and many other purposes, LiF:Mg, Ti is the material of choice because it is approximately tissue-equivalent (effective atomic number of 8.2 compared to 7.4 for tissue) and almost energy-independent for photons with an energy between 0.1 and 3 MeV [86Att]. Moreover, it is to some extent sensitive to thermal neutrons, since natural lithium (TLD-100) contains to 7.4 % 6Li that has a high (n,α) capture cross section. The response to neutrons can be enhanced or reduced by using 6Li (TLD-600, with 96.5 % 6Li) or 7Li enriched material (TLD-700, with 99.99 % 7Li) (cf. Section 10.1.7). a
b
c
Fig. 10.18. (a) TL detectors for radiation protection measurements (chips, rods, and ring). (b) TLD reader for the evaluation of irradiated detectors. (c) Annealing oven for the preparation of the detectors. TL detectors are placed in a metallic container which comes into contact with two heating plates to ensure optimal thermal equilibrium. (Courtesy PTW Freiburg).
In some substances − such as BeO, LiF, Al2O3, CaSO4, and some alkali halides − electrons liberated during the heating procedure may also have a chance to leave the material when the trap sites are located in a thin (<10 nm) surface layer. This process, that is closely related with the conventional TL mechanism [99Sak], is referred to as thermally stimulated exoelectron emission (TSEE). The low energy exoelectrons can be detected, for example, by a windowless proportional or Geiger-Müller counter (cf. Section 10.1.2.3 and 10.1.2.4, respectively). The number of exoelectrons detected is proportional to the dose in the surface layer. TSEE dosimetry has gained interest since it can be used for the measurement of weakly penetrating radiation, such as low energy γ-quanta or α- and β-particles [86Sch]. A limitation of conventional TLD technology is the complete annealing of populated traps, when the material is heated during the readout process. Therefore, alternative luminescence techniques have been investigated that permit successive readouts and the construction of integrating dosimeters. An approach that has been applied more frequently in health physics is based on the optical phenomenon of radiophotoluminescence (RPL) [68Att, 87Per]. RPL is a property of certain substances to emit fluorescent light Landolt-Börnstein New Series VIII/4
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in the visible range upon irradiation with ultraviolet light, when the substance has previously been exposed to ionizing radiation. Although various materials exhibit this property, silver-activated metaphosphate glasses are most commonly used. In RPL glasses, ionizing radiation results in the formation of stable fluorescing centers that emit orange light (about 640 nm) under (pulsed laser) ultraviolet stimulation (365 nm) [93Pie]. The intensity of fluorescence light, which is determined with a spectrally matched PMT as in the case of traditional TLD readout, is proportional to the number of light centers and hence to the radiation dose. Since the light centers are not destroyed by ultraviolet excitation, the readout procedure can be repeated as often as necessary. For practical application as personal dosimeters, optimized RPL (‘Yokato’) glasses are covered with metal filters or capsules which provide largely energy-independent dose measurements for γ-quanta with an energy above 50 keV [72Kie, 96Hoh].
10.1.6 Photographic films Film dosimetry is an attractive technique for many applications – especially in medical physics – due to its high spatial resolution, wide accessibility, and the flexibility to place the film in humanoid phantoms [00Sha]. Furthermore, photographic films are still widely used for monitoring radiation exposure of personnel although they do by no means meet the requirements of an ideal personal dosimeter in all respects. Photographic films consist of an emulsion of microscopic grains of silver halides, usually silver bromide, dispersed in a gelatine layer on one or both sides of a transparent film base of cellulose acetate or polyester. The photographic process is very complicated. A thorough discussion of the complex experimental and theoretical aspects, which are considerably simplified in the following description, can be found in [02Bus] and [81Bar], respectively. When the silver bromide grains in the emulsion are exposed to visible light or ionizing radiation, bromide ions absorb energy and are oxidized. The electrons from this oxidation process reduces silver ions to silver atoms. Experimental evidence indicates that a minimum of three to five reduced silver atoms form a ‘sensitized’ grain, that can act as a catalyst for the chemical amplification process during film development. When the film is placed in a chemical developer solution, all silver ions in the grains will be reduced to silver atoms – independently of whether the grains were affected by ionizing radiation or not. Silver atoms in ‘sensitized’ grains, however, greatly enhance the rate of reduction of additional silver ions. Based on this fact, the chemical development process is terminated after a time at which the reduction process in ‘sensitized’ grains is completed by taking the film out of the developer solution and placing it in a fixing solution that neutralizes residual developer present on the film. The chemical development process greatly increases the number of silver atoms and makes the radiation effect measurable. To ensure reproducibility of the results, careful control of the developing procedures is essential. The degree of macroscopic blackening of the processed film depends on the number of silver atoms deposited and thus on the amount of energy absorbed. It is usually expressed in terms of the optical density OD which is defined in terms of the transmission of light through the film as OD = log 10 ( I 0 I ) , where I 0 and I are the intensities of a light beam measured by an optical densitometer in front and behind the film. The relation between the optical density of the film and the exposure is described by the ‘characteristic curve’ for the particular film – typically a sigmoid-shaped curve with three characteristic regions: At low exposures, the optical density is low and independent of exposure level. Next, there is a segment over which the optical density and the logarithm of the exposure are approximately linear related. This is the region of normal operation. In the third segment, which corresponds to large exposures, the film becomes saturated since all silver ions are converted to metallic silver [81Bar]. The response of photographic films is strongly dependent on radiation energy. For photon energies below 100 keV, for example, the relative sensitivity is between 10 and 50 times higher than at higher energies [86Att]. In order to flatten the energy response, film badge dosimeters contain a set of metallic filters of various materials and thicknesses over different regions of the film as shown in Fig. 10.19. By comparison of the optical density behind these filters, it is possible to get rough spectral information Landolt-Börnstein New Series VIII/4
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which can be utilized to take the energy dependence of the film’s sensitivity approximately into account. Disadvantages of film badge dosimeters are, firstly, the considerable angular dependence of the detection efficiency and, secondly, fading of ‘sensitized’ grains in the period between exposure and development [72Kie]. Both effects can result in a considerable underestimation of radiation exposure. Fig. 10.19. Film-badge dosimeter for personal monitoring. The interior view of the holder shows at the back and front an open window, three copper filters of different thickness, and a staggered lead / tungsten filter to identify the direction of the incident radiation. The light-tight wrapping contains two dosimetry films of different sensitivities.
10.1.7 Detectors for neutrons In the previous sections, devices for the detection of photons and charged particles have been described, which deposit their energy in matter predominantly via electromagnetic interactions − in particular by inelastic collisions with the atomic electrons. In contrast, the interaction of fast neutrons with matter occurs through processes with the nucleus, e.g. elastic or inelastic scattering or various nuclear reactions. The most important nuclear reactions for the detection of neutrons in the eV to keV region are (n,α) and (n,p) reactions, such as 6Li (n,α) 3H, 10B (n,α) 7Li, or 3He (n,p) 3H, in which a neutron is captured and a charged particle is released. These substances have a high sensitivity at low neutron energies, which is of particular advantage considering the low dose rate levels routinely encountered in radiation protection, but become more and more ineffective at higher neutron energies E because the capture cross sections vary as 1 / E . Instead, elastic scattering of neutrons at nuclei becomes the most relevant interaction mechanisms, giving rise to recoil nuclei. Whereas the energy transfer to a recoil nucleus is very ineffective in the case of heavy nuclei, an incident neutron can transfer up to its entire energy in a single collision to a hydrogen nuclei (proton) because neutrons and protons have nearly the same mass. Therefore, detectors with a high content of hydogen in the detector or cover material are used for neutron detection by measuring the recoil protons. In addition, hydrogen-containing materials, so-called moderators, can be utilized for slowing down of fast neutrons to thermal energies. A more quantitative discussion of the various interaction processes, by which neutrons transfer all or part of their energy to charged particles capable of exciting and ionizing, can be found in [72Kie, 00Kno, 94Leo]. A sensitive and simple counting device for slow neutrons is a proportional counter filled with BF3 gas, usually enriched to more than 90 % in 10B. In such a counter, BF3 gas not only serves as the target for slow neutron conversion into α-particles via the above mentioned 10B (n,α) 7Li reaction but also as filling gas. Since α-particles yield a much higher output signal than γ-particles, BF3 counters can be used to discriminate between the neutron and γ-component in a mixed n-γ field (cf. Section 10.1.2.3). By sourrounding the counter by a moderating material (e.g. paraffin or polyethylene) in which the neutrons are slowed down, it can also be utilized for the detection of fast neutrons. An alternative approach for the detection of slow neutrons is the use of proportional counter, the walls of which are coated with a layer of B2O3 or BC4. Such ‘boron-lined counters’ offer the flexibility to use more appropriate filling gases than BF3 [00Kno, 95Tso]. A particular problem in neutron dosimetry is the broad energy range that can occur and the considerable variation of detector response over this range. Nevertheless, dosimeters for routine area monitoring can be constructed in such a way that the shape of their fluence response as a function of energy approximate that of the fluence-to-dose equivalent conversion function and thus will give a reading approximately proportional to the ambient dose equivalent H*(10) over most of the energy range of interest Landolt-Börnstein New Series VIII/4
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(cf. Section 4.5.3.3 and [01ICR]). To this end, thermal-neutron detectors are surrounded by a moderator, whose material and geometry is chosen to optimize the desired response function. A widely used survey instrument is the ‘rem counter‘ shown in Fig. 10.20. It consists of a massive sphere with a diameter between 25 and 30 cm of hydrogenous material – usually polyethylene with additional layers of cadmium or boron – that serves as neutron moderator surrounding a thermal-neutron detector, e.g. a 3He proportional tube. An extension of this approach is the ‘Bonner sphere’ spectrometer, which uses different moderator spheres with diameters ranging from 5 to 30 cm. Since the response functions for the diverse moderator spheres differ in shape and position of the maxima, the energy distribution of a neutron field can be derived from count rate measurements performed separately for each of the spheres [00Kno, 00Sha]. Shielding Front window 6 7
Body
LiF TLD LiF TLD
Albedo window Shielding
Fig. 10.20. Dose rate monitor (‘rem counter’) for the measurement of neutron ambient dose equivalent H*(10) for neutrons up to an energy of 20 MeV. The detector system consists of a proportional counter tube centered in a moderator sphere with a diameter of 250 mm. (Courtesy Berthold Technologies).
Fig. 10.21. Schematic representation of an albedo TLD dosimeter consisting of two pairs of 6LiF - 7LiF detectors that are shielded by cadmium- or boron-loaded material either against incident thermal neutrons or albedo neutrons from the rear.
As described in Section 10.1.2.3, tissue-equivalent proportional counters (TEPCs) are capable of measuring absorbed dose to the sensitive gas volume and of determining an approximation of the dose equivalent by its spectroscopic properties. They thus provide additional information normally not available from other neutron-measuring instruments. Due to practical limitations, however, TEPCs are not yet readily usable in everyday routine monitoring [01ICR, 00Sha]. For personal monitoring in mixed n-γ fields, the most commonly used devices today are albedo TLD badges, containing TL detector chips with a high fraction of 6LiF (such as TLD-600), that has a large 6 Li (n,α) 3H cross section for thermal neutrons, as well as chips with a high fraction of 7LiF (such as TLD-700), that is insensitive to neutrons. The dosimetry of fast neutrons becomes possible when the batch is held closely to the body where it will be exposed to low energy neutrons that are backscattered from the body as the result of the moderation of fast neutrons within the body. These scattered neutrons are called albedo neutrons. As an example, Fig. 10.21 shows the schematic design of a TLD albedo dosimeter that consists of two pairs of 6LiF - 7LiF detectors. One pair is shielded from the rear with a thermal-neutron absorber (e.g. cadmium- or boron-loaded plastic) the second from the front. The readings obtained from the incident-neutron and albedo-neutron detectors are combined into an overall calculation of neutron dose equivalent. It should be mentioned, however that the resulting dose equivalent response of the dosimeter varies greatly with neutron energy at intermediate and high energies. Therefore, a single calibration factor cannot be used in different neutron fields with widely varying spectra if accurate dose results are to be obtained. In such cases, it is necessary to keep a record of the location in which the TLD batch was used, and to apply the appropriate calibration factor for the reading [01ICR]. Landolt-Börnstein New Series VIII/4
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Alternatively, etched-track detectors are used for neutron personal dosimetry. These passive devices utilize the ability of some materials to register the tracks of (neutron induced) charged particles as damage to their structure. The permanent damage tracks can be greatly enlarged by chemical or electrochemical etching procedures. The resulting conical etched pits in the material at the damage sites are visible to the naked eye and can be counted automatically by optical systems. Most frequently, plastic foils of ‘polyallyl diglycol carbonate’ (PADC, also known by the trade name CR-39) are used, which are uniquely sensitive to recoil protons released by neutrons passing through the material. PADC is sensitive to fast neutrons with an energy down to about 100 keV, and, by adding a converter layer, also to thermal and epithermal neutrons. PADS foils are insensitive to photons and β-particles and are thus suitable to measure the neutron component in a mixed n-γ field [01ICR, 97Bar]. (Due to the same reason, PADC detectors are used to estimate personal radon doses as well as integrated indoor radon concentrations by detecting the α-particles from the decays of radon and its daughter products.) Superheated-drop or bubble detectors can be employed for both area and personal monitoring. They consist of microscopic droplets of liquid halocarbons and hydrocarbons dispersed in a gel or polymer matrix. The liquid droplets are superheated and are thus in a metastable state. When a neutron interacts with a nucleus inside or near a droplet, the resulting secondary charged particles locally increase the temperature and thus cause local vaporization. By this process, small vapor bubbles are formed that begin to expand by vaporizing adjoining liquid. If a vapor bubble reaches a critical size, all of the liquid in the droplet will be vaporized resulting in a visible gas bubble with a diameter of up to a millimeter. If the neutron energy exceeds a specific detector threshold − which depends on the atomic composition and size of the superheated droplets and the temperature and pressure of the matrix − the overall absorbed energy is correlated to the number of bubbles in a vial [01ICR, 00Kno, 00Sha]. Typical sensitivities are in the range of a few bubbles per µSv. Bubble detectors can be constructed as passive or active devices. In the first case, the superheated droplets are dispersed in a viscous polymer gel so that the bubbles remain fixed and thus can be counted at the end of the measurement by eye or automatically using an optical scanner. Active devices can be realized, for example, by placing an electro-acoustic transducer in contact with the detector, so that each time a bubble is formed, the sound that is produced is converted into an electrical pulse. An important feature of all superheated drop detectors is their almost complete insensitivity to photons and electrons with energies up to about 6 MeV [01ICR]. Further types of neutron detectors for area and personal monitoring as well as guidance concerning the measurement of operational dose equivalent quantities for neutron radiation are given in [01ICR, 99Alb].
10.1.8 Biological dosimetry The human genome contained in the cell nucleus is physically carried by 46 chromosomes, each of which is composed of an extremely long, double-stranded helical DNA molecule in a closely packed form. Chromosomes are clearly visible through a light microscope during the metaphase – a certain stage of cell division, in which the condensed chromosomes become attached to spindle fibers and lie in a central plane of the nucleus. In this stage of cell cycle, chromosomes are already replicated (doubled). The newly formed twin chromosomes, which are called chromatides, remain temporarily attached to each other at a point, the centromere, located near the center of each of the chromatides (Fig. 10.22a). There is a large body of evidence that biological effects of ionizing radiation result principally from damage to DNA. The most important effects caused by ionizing radiation are breaks in the DNA double helix. Single-strand breaks are restored immediately by molecular repair mechanisms using the opposite DNA strand as a template. If the repair is incorrect, the genetic information is altered. Although such point mutations may lead to inherited effects of offspring or the induction of carcinogenesis, they do not result in chromosomal aberrations visible under the microscope. On the other hand, breaks at both strands on opposite sides of the double helix that are juxtaposed may lead to a double-strand break. When this happens, different types of chromosomal aberrations can be observed: If the radiation-induced damage occurs before the chromosomes have been replicated, the lesion is dublicated during DNA synthesis and thus both chromatides show the same aberration (chromosome aberration). If, on the other hand, the cell Landolt-Börnstein New Series VIII/4
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is irradiated after the DNA material has already been doubled, only one of the two twin chromatides is damaged (chromatide aberration). Gross distortions that are clearly visible in the metaphase are di- and tricentric chromosomes. They are formed when a break occurs in two or even three chromosomes and the centric fragments – carrying the centromere – incorrectly join with each other at their broken ends. The remaining acentric fragments are lacking a centromere and can thus easily be detected in the metaphase, too. A detailed description of the various types of radiation-induced chromosomal aberrations is presented in [00Hal]. Cytogenetic analysis of chromosomal aberrations is a valuable tool for retrospective dose assessment of individuals that are accidentally overexposed to ionizing radiation. It fills a gap in dosimetry, since the radiation effect on the human body is determined directly without the intermediary of dosimetric measurements using technical devices. Scoring of chromosome aberrations in peripheral blood lymphocytes, mainly the dicentric assay, is regarded as the most specific method and has become a routine component of accidental dose assessment [02Voi]. Figure 10.22a shows representative chromosome aberrations in a metaphase preparation of irradiated lymphocytes. Many studies in animals and humans have shown a good relation between the results obtained in vivo and in vitro and this provides evidence that in vitro established dose-effect relationships can be used as calibration curves [02Voi]. However, as shown in Fig. 10.22b, the formation of dicentrics strongly depends on radiation quality and dose rate so that information about these variables needs to be established for each investigation. As mentioned above, DNA breaks must be induced in two different unreplicated chromosomes in order to form a dicentric chromosome. Since this can be achieved either by a single particle breaking on occasion both chromosomes or two particles each of which damaging only one chromosome, the frequency N of dicentric chromosomes per cell can be well fitted as a function of dose D by a linear-quadratic model, N = α D + β D2. At low doses, one-particle events (described by the linear term, α D) are more frequent, whereas two-particle events (described by the quadratic term, β D2) dominate at high doses. Besides the calibration curve used, the precision of dose estimation depends mainly on the number of cells observed and the background level. In practice, detection of low LET radiation is possible for doses above 150 mGy [02Voi]. 0.8
b Dicentric chromosomes per cell
a
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γ-rays, 60Co, 0.1 Gy/min X-rays, 250 kVp, 1.0 Gy/min X-rays, 90 kVp, 0.1 Gy/min neutrons, 15 MeV neutrons, 0.7 MeV
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Fig. 10.22. Radiation-induced chromosomal aberrations in human lymphocytes. (a) Metaphase preparation with normal, dicentric (D), and tricentric chromosomes (T) as well as various fragments (F). (b) Frequency of dicentric chromosome aberrations per cell for several types of radiation. The curves give the result of linear (low-energy neutrons) and linear-quadratic fits to measured data. (Courtesy G. Stephan, BfS).
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10.1.9 References for 10.1 68Att 72Kie 79ICR 81Bar 83ICR 86Att 86Sch 87Per 89Die 89Men 93Pie 94ICR1 94ICR2 94Leo 95Hir 95ISO 95Sch 95Tso 95Wak 96Hoh 96Kah 96Ols 97Bar 99Alb
99Lut 99Sak 99Wer 00Hal 00Kno 00Sha
Attix, H., Roesch, W.C., Tochilin, E.: Radiation Dosimetry. 2nd edition, Vols. I-III. New York: Academic Press, 1968-1969. Kiefer, H., Maushardt, R.: Radiation Protection Measurement. Oxford: Pergamon Press, 1972. ICRU Report 31: Average energy required to produce an ion pair. Bethesda, MD: ICRU Publications, 1979. Barrett, H.H., Swindell, W.: Radiological Imaging. London: Academic Press, 1981. ICRU Report 36: Microdosimetry. Bethesda, MD: ICRU Publications, 1983. Attix, F.H.: Introduction to radiological physics and radiation dosimetry. New York: John Wiley & Sons Inc., 1986. Scharmann, A, Kriegseis, W.: Radiat. Prot. Dosim. 17 (1986) 359. Perry, J.A.: RPL Dosimetry: Radiophotoluminescence in health physics. Bristol: IOP Publishing, 1987. Dietze, G., Menzel, H.G., Schuhmacher, H.: Radiat. Prot. Dosim. 28 (1989) 77. Menzel, H.G., Paretzke, H.G., Booz, J. (eds.): Implementation of dose-equivalent meters based on microdosimetric techniques in radiation protection; Radiat. Prot. Dosim. 29 (1-2) (1989). Piesch, E., Burgkhardt, B., Vilgis, M.: Radiat. Prot. Dosim. 47 (1993) 409. ICRU Report 52: Particle counting in radioactivity measurements. Bethesda, MD: ICRU Publications, 1994. ICRU Report 53: Gamma-ray spectrometry in the environment. Bethesda, MD: ICRU Publications, 1994. Leo, W.R.: Techniques for nuclear and particle physics experiments, 2nd edition. Berlin: Springer-Verlag, 1994. Hirning, C.R.; Yuen, P.S.: Health Phys. 69 (1995) 46. International Organization for Standardization: Guide to the expression of uncertainty in measurement. Geneva, Switzerland: ISO, 1995 (corrected of first print in 1993). Schmitz, Th., Waker, A.J., Kliauga, P., Zoetelief, H. (eds.): Design, construction and use of tissue-equivalent proportional counters – EURADOS report; Radiat. Prot. Dosim. 61 (4) (1995) Tsoulfanidis, N.: Measurement and detection of radiation, 2nd edition. Washington: Taylor & Francis, 1995. Waker, A.J.: Radiat. Prot. Dosim. 61 (1995) 297. Hohlfeld. K.: Nachweismethoden für ionisierende Strahlung, in: Kose, V.; Wagner, S. (eds.), Praktische Physik, Band 2. Stuttgart: B.G. Teubner, 1996, Kapitel 7.4. Kahilainen, J.: Radiat. Prot. Dosim. 66 (1996) 459. Olsher, R.H., Eisen, Y.: Radiat. Prot. Dosim. 67 (1996) 271. Bartlett, D.T.; Steele, J.D., Tanner, R.J., Gilvin, P.J., Shaw, P.V., Lavelle, J.: Radiat. Prot. Dosim. 70 (1997) 161. Alberts, W.G., Arend, E., Barelaud, B., Curzio, G., Decossas, J.L., d´Èrrico, F., Fiechtner, A., Grillmaier, R., Meulders, J.-P., Menard, S., Roos, H., Schuhmacher, H., Thevenin, J.-C., Wernli, C., Wimmer, S.: Advanced methods of active neutron dosimetry for individual monitoring and radiation field analysis (ANDO), Report PTB-N-39, Braunschweig, 1999. Lutz, G.: Semiconductor radiation detectors. Device Physics. Berlin: Springer-Verlag, 1999. Sakurai, T.; Tomita, A., Fukuda, Y.: J. Phys. D 32 (1999) 2290. Wernli, C., Fiechtner, A., Kahilainen, J.: Radiat. Prot. Dosim. 84 (1999) 331. Hall, E.J.: Radiobiology for the radiologist, 5th edition. Philadelphia: Lippincott Williams & Wilkins, 2000. Knoll, G.F.: Radiation detection and measurement, 3rd edition. New York: John Wiley & Sons, Inc. 2000. Shani, G.: Radiation dosimetry: Instrumentation and methods, 2nd edition. Boca Raton: CRC Press 2000.
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10 Measuring techniques ICRU Report 66: Determination of operational dose equivalent quantities for neutrons. Journal of the ICRU, Volume 1, No 3, 2001. Takahashi, T; Watanabe, S.: IEEE Trans. Nucl. Sci. 48 (2001) 950. Bushberg, J.T.; Seibert, J.A., Leidholdt, E.M., Boone, J.M.: The Essential Physics of Medical Imaging, 2nd edition. Philadelphia: Lippincott Williams & Wilkins, 2002,. ISO/ASTM51707-2002(E): Guide for estimating uncertainties in dosimetry for radiation processing. Geneva, Switzerland: International Organisation for Standardization, 2002. Saint-Gobain Crystals & Detectors: Product Data Sheets, www.detectors.saint-gobain.com. Voisin, P.; Barquinero, F., Blakely, B., Lindholm, C., Lloyd, D., Luccioni, C., Miller, S., Pallitti, F., Prasanna, P.G., Stephan, G., Thierens, H., Turai, I., Wilkinson, D., Wojcik, A.: Cell. Mol. Biol. 48 (2002) 501. Waker, A.J.; Schrewe, U., Burmeister, J., Dubeau, J., Surette, R.A.: Radiat. Prot. Dosim. 99 (2002) 311.
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10.2 Radiological protection measurements: external exposure Radiological protection aims at the restriction of the doses to the human body, the effective dose and the equivalent dose in an organ or tissue by applying constraints and limits. The assessment of these doses is therefore fundamental to the practice of radiological protection. However, neither the equivalent dose in an organ or tissue nor the effective dose can be measured directly. Values of these quantities must be inferred from measurable quantities with the aid of models. Radiological protection measurements therefore include measurements related to the system of radiological protection and the interpretation of these measurements in the assessment of external and internal exposures. For monitoring external exposure specific operational dose quantities have been defined which normally provide an estimate of effective dose sufficiently accurate for the purpose of radiological protection (see Sect. 10.2.1). Details about the types of detectors, which can be used to measure these operational quantities are described in Sect. 10.1. The primary justification of any monitoring program is to achieve and demonstrate an appropriate level of protection. Further objectives of monitoring programs are to • estimate the actual radiation exposure level, • demonstrate compliance with legal requirements, • demonstrate good working practices, • provide data for use in reviewing optimization programs, • provide data for medical purposes as required, • provide data for use in epidemiological studies. Monitoring programs can be distinguished with regard to the objectives and the location of monitoring. While area monitoring provides a dose or dose rate which enable to estimate the dose a person would receive when staying for a specified time period at the location of interest, individual monitoring provides an estimate of the dose a person has already received. Monitoring can be organized as routine, task-related and special monitoring. Local monitoring can be performed at the workplace, e.g. by means of area monitoring. Individual monitoring can be performed by measuring the external exposure, the internal exposure, and the skin contamination. Routine monitoring is associated with continuous operation and is largely of confirmatory nature. Operational individual monitoring is associated with a particular operation. It may make use of supplementary dosimeters in addition to those used for routine monitoring. Special individual monitoring will be applied in actual or suspected abnormal conditions including incidents and accidents. The result of monitoring may be used to initiate certain actions when a pre-defined dose level is exceeded.
10.2.1 Operational quantities The International Commission on Radiation Units and Measurements (ICRU) has defined a set of operational dose quantities for area and individual monitoring of external exposure [85ICR, 92ICR, 93ICR] which were designed to provide an estimate of the protection quantities defined by ICRP [77ICR] and to serve as calibration quantities for dosimeters used in monitoring. More information about the definition of the operational quantities is given in Sect. 4.5. For area monitoring, the appropriate operational quantities are the ambient dose equivalent H*(10) for strongly penetrating radiation, and the directional dose equivalent H'(0.07,Ω ), for weakly penetrating radiation (see Sect. 4.5.3.3). For individual monitoring, the quantity personal dose equivalent Hp(d ) was defined, which is the dose equivalent in ICRU soft tissue, at an appropriate depth d below a specified point on the body where the individual dosimeter is worn [92ICR, 93ICR]. For strongly penetrating radiation a depth of 10 mm, denoted by Hp(10), and for weakly penetrating radiation a depth of 0.07 mm, denoted by Hp(0.07), is used. A depth of 3 mm, denoted by Hp(3), was also proposed for monitoring the exposure of the lens of the eyes but has never been used in practice.
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Personal dose equivalent is defined in the human body and may, therefore, in a given exposure situation vary between individuals. The value may also depend on the position of the dosimeter worn on the body. Consequently, the personal dose equivalent can be expected to vary between locations of any given individual and is anticipated to be a multi-valued quantity [96ICR, 98ICR, 99Zan]. For routine monitoring in cases where the readings are far beyond the corresponding legal limits those values are seen to provide a sufficient approximation to the corresponding protection quantity, e. g. effective dose, if the dosimeter is worn at a position representative for the exposure. To make this quantity single-valued in a given exposure situation, both a particular location of the dosimeter on the human body and a particular phantom of the body need to be specified for evaluation. More information is given in Chap. 6. Table 10.5 summarizes the objective of dose control and the corresponding operational dose quantities used and specifies their application. Table 10.5. Operational dose quantities and their objectives in external monitoring Dose quantities for Objective area monitoring individual monitoring control of effective dose ambient dose equivalent, H*(10) personal dose equivalent, Hp(10) control of skin equivalent dose directional dose equivalent, personal dose equivalent, Hp(0.07) H'(0.07) control of equivalent dose of directional dose equivalent, H'(3) personal dose equivalent, Hp(3) the eye lens The operational quantities for area and individual monitoring of external exposure are chosen to approximately assess the effective dose under most exposure conditions. Because of the different models used in the definition of the quantities the ratio of the operational quantities and the effective dose depends on the type and the energy of the radiation considered and on the direction of radiation incidence on the body. Figs. 10.23, 10.24, and 10.25 show the ratios E/H*(10) and E/Hp(10) for photons and neutrons under various exposure conditions. 1.6
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AP PA RLAT LLAT ROT
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AP PA RLAT LLAT ROT ISO
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0.2 0 10 2
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Fig. 10.23. Ratio of effective dose E and ambient dose equivalent H*(10) versus photon energy for various directions of photon radiation incidence on the human body [96ICR]. AP: frontal incidence, LLAT: left lateral incidence, RLAT: right lateral incidence, PA: incidence from the back, ROT: incidence rotational to the vertical axis, ISO: isotropic incidence.
0 10-2
1 10-1 Photon energy [ MeV ]
10
Fig. 10.24. Ratio of effective dose E and personal dose equivalent Hp(10) versus photon energy for various directions of photon radiation incidence on the human body and the dosimeter worn in front of the lung [96ICR, 99Zan]. AP: frontal incidence, LLAT: left lateral incidence, RLAT: right lateral incidence, PA: incidence from the back, ROT: incidence rotational to the vertical axis.
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Fig. 10.25. Ratio of effective dose E and ambient dose equivalent H*(10) versus neutron energy for various directions of neutron radiation incidence on the human body [96ICR]. AP: frontal incidence, LLAT: left lateral incidence, RLAT: right lateral incidence, PA: incidence from the back, ROT: incidence rotational to the vertical axis, ISO: isotropic incidence.
10.2.2 Reference levels Reference levels are values of measured quantities above which specific actions or decisions should be taken. In the context of this section the most important reference levels are the “Investigation Level” and the “Intervention Level” also called „Action Level“ (see Sect. 4.8). In practical implementation of monitoring programs additional reference levels might be required. They may include levels for recording the monitoring results (“Recording Level”) and for their reporting (“Reporting Level” or “Notification Level”). Measured values above the “Investigation Level” require an investigation of the reason and the implication of the measured value. “Investigation Levels” are specifically defined by the operating management and they can apply both to the individual and the working environment. It is appropriate to select “Investigation Levels” for individual dose and intake on the basis of expected levels or on the basis of a selected fraction of the relevant dose limit. “Investigation Levels” should be defined at the planning stage of any practice although they may need to be revised on the basis of operational experience. In the medical field specific “Diagnostic Reference Levels” for patients and for standard applications of ionizing radiation and radioactive substances are defined which characterize a dose level corresponding to the technical and operational state of the art which should be considered for avoiding situations where the level of dose to a patient or the administered activity is unusually high. Intervention applies to those situations where the source, pathways and exposed individuals are already in place when the decisions about control or remedial measures are being considered. “Intervention Levels” are set by competent authorities and are often mandatory. Typical examples of interventions are actions taken after a radiological emergency to protect the members of the public. There may also be the need to undertake an intervention to protect workers involved in accidents at the workplace. Intervention may also be necessary to decrease the exposure of workers in de-facto situations, e.g. to elevated levels of natural radiation.
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10.2.3.1 Occupational exposure Radiation work is defined [91ICR1, 96EU] as work in which the annual effective dose of an exposed worker of age 18 or over from radiation sources at work may exceed the annual dose limits for members of the public, e.g. an effective dose of 1 mSv or an equivalent dose of the lens of the eye of 15 mSv or an equivalent dose of the skin of 50 mSv averaged over any 1 cm2 area, regardless of the area exposed. For occupational exposure the ICRP has recommended a limit on the effective dose of 20 mSv per year averaged over 5 years (100 mSv in 5 years) [91ICR1] with the further provision that the effective dose should not exceed 50 mSv in any single year. The limit on equivalent dose for the lens of the eye is 150 mSv in a year, the limit on equivalent dose for the skin is 500 mSv in a year. It is implicit in these recommended dose limits that the dose constraint for optimization should not exceed 20 mSv in a year. The limits on effective dose for apprentices and students aged between 16 and 18 years who, in the course of their studies are obliged to use radioactive sources, is 6 mSv per year. Special protection is required during pregnancy: the equivalent dose to the child to be born is limited to 1 mSv to the reminder of the duration of pregnancy (see also Sect. 4.8). The decision to provide individual monitoring for an individual or a group of workers depends on three major factors: the expected dose in relation to the constraint or limit, the likely variations in the dose in time and space, and the complexity of the measurement and of the interpretation procedures. Individual monitoring is required for category A workers. It should be established and monitored by an approved dosimetric service. Category A includes any radiation work in which the annual effective dose is or might be higher than 6 mSv, or the annual equivalent dose of the lens of the eye, the skin or hands and feet is or may be higher than 3/10 of the dose limit stipulated for these tissues or organs. Category B includes all other radiation work. For practical reasons, monitoring of category B workers is often treated similar to category A workers by individual monitoring. In many cases the individual monitoring of external exposure is fairly simple and does not require a heavy commitment of local resources (see 10.2.6). However, in mixed radiation fields, e.g. neutron/photon fields, monitoring is much more complex (see 10.2.6). For special groups of workers, e.g. to the air craft crew, doses caused by cosmic radiation are determined by calculations based on their flying hours and the flight plans rather than by individual measurement. In situations where individual monitoring is not appropriate or feasible the occupational exposure shall be assessed on the basis of the results of area monitoring at the workplace and on information on the location within the area considered and the duration of exposure. The control of occupational exposure can be simplified by the designation of work places as “controlled” and “supervised” areas. “Controlled Areas” are subject to special rules for the purpose of radiation protection and to which access is controlled. In “Supervised Areas”, a minimum radiological surveillance of the working environment will be organized. Outside these designated areas, the dose rates from sources and the risk of contamination by unsealed radioactive material will be low enough to ensure that the level of protection for those who work in the premises will be comparable with the level of protection required for the public. In several areas of medicine the control of occupational exposures is of particular importance, e.g. nursing of brachytherapy patients, palpation of patients during fluoroscopy, and interventional radiology [96ICR]. In these cases individual monitoring with careful scrutiny of the results is always important. 10.2.3.2 Public exposure The control of public exposure in all normal situations is exercised by the application of control at the source. Almost all public exposure is controlled by the procedures of constrained optimization and the use of dose limits. In particular, appropriate monitoring equipment and surveillance programs are required from the licensee to assess public exposure related to any practice or source and to demonstrate that the dose to members of the public does not exceed authorized dose limits. This includes environmental monitoring systems measuring doses or dose rates in the vicinity of a source or widely distributed in a country for the purpose of surveillance or early warning. Routine individual monitoring of public exposure is not necessary under normal conditions. Landolt-Börnstein New Series VIII/4
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In all areas of medicine there are no restrictions on the public access to non-designated areas. Public access to controlled areas should be limited to visitors of patients only, who should be advised of any restrictions on their behavior. All reasonable steps shall be undertaken to assess any exposure incurred by members of the public as a consequence of an accident. This assessment will be based on data of area dosimeters installed in and around the facility involved, on model calculations based on plant status or on information about any environmental contamination, and on results of environmental and individual (physical and biological) monitoring. For practical purposes decisions on interventions are often based on derived secondary limits of dose rate and on values of the contamination level in agricultural and other environmental products, which can easily be measured [99IAE]. Emergency response personnel, although not normally occupationally exposed may have to carry out their duties in areas where there is a potential for elevated radiation exposure. Protection of this personnel should be treated as part of the occupational exposure incurred in a practice.
10.2.4 Types of monitoring programs In the context of this Section, two types of monitoring programs will be discussed, e.g. individual and area monitoring. In many cases where photon radiation is dominant both programs can be considered as independent means to estimate the effective dose a person would receive or has received. In situations with significant contribution of β-radiation or in mixed neutron-photon radiation fields, data and information from both types of monitoring programs might be required to arrive at reliable estimates of the total dose. 10.2.4.1 Individual monitoring for external exposure External dosimetry deals with radiation that originates outside the body. The external exposure may result from photon irradiation (X- or γ-rays), particle irradiation (electrons, neutrons, protons, heavy particles) or from mixed irradiation (e.g. γ-rays and neutrons). The exposure may involve the whole body or may be confined to a sizeable part of the body. It may be localized, from a narrow beam irradiation or a small radiation source near to the body. The design of a monitoring program should include the specification of the type(s) of dosimeter to be used and how and where they should be worn (see Sect. 10.2.5). In complex and inhomogeneous fields it will often be necessary to use more than one dosimeter. In particular, operations involving manipulations of radioactive sources may call for dosimeters worn on the fingers. In radiation fields with both penetrating and weakly-penetrating radiations, e.g. γ- and β-rays, a two component dosimeter is required. Sometimes neutrons may contribute to the total dose from occupational exposure. In situations where neutron exposures are likely to significantly contribute to the effective dose special neutron dosimeters are necessary for monitoring. A detailed overview on individual monitoring of external radiation is given in [01Bar]. In the case of radiological accidents with low external exposures only, the assessment of effective and equivalent dose would be covered by routine monitoring programs. In cases where highly elevated dose levels can not be excluded additional dosimeters, preferentially with direct reading of the dose and dose rate and with the option of an audible or acoustical warning should be considered. 10.2.4.2 Area monitoring for external exposure The purpose of area monitoring at workplaces is to ascertain that a working area is free of significant levels of radiation and contamination. Area monitoring allows the warning of personnel to avoid hazardous areas. The nature and frequency of workplace monitoring shall be sufficient to evaluate the Landolt-Börnstein New Series VIII/4
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radiological condition at the workplace and to assess the exposure in controlled areas. The routine workplace monitoring program will usually involve the use of repeated survey measurements. Such a program may include the use of continuously operating monitors installed at fixed and representative locations to monitor the normal radiation level and to identify the onset of abnormal or emergency conditions. The frequency with which routine monitoring will be conducted is determined by the stability of the radiation environment. If the radiation fields are liable to increase rapidly and unpredictably to significant levels, the monitoring program should include instrumentation with technical provisions for fast early warning. Most of the instruments used at the workplace will measure dose rate rather than dose. Particular care is required in the selection and calibration of instruments used to measure neutrons, β-rays and low energy photons. Task related monitoring will provide forecast of the doses likely to be accumulated during a task. For this purpose portable instruments will preferably be used. Particular care is needed when working with beta and other weakly penetrating radiations. Special care should be given to the measurement of the dose rate adjacent to surface or point sources. 10.2.4.3 Calibration Calibration aims at establishing the relationship between values indicated by a measuring instrument or system (see Sect. 10.1) and the corresponding true values of a quantity to be measured. The radiation types used for the calibration of dosimeters are mainly photons, neutrons and beta particles. Calibrations for each of these types are performed differently using different instrumentation and techniques. Calibration should closely follow the recommendations of the International Organisation for Standardisation (ISO) dealing with reference radiations and be based on the methods described in these standards [96ISO1, 96ISO2, 97ISO, 98ISO, 99ISO, 00ISO1, 01ISO]. A detailed description of the calibration procedures can be found in [94Alb, 00Die]. The calibration of personal dosimeters or area survey meters used for radiation protection purposes is mostly a three step process. First, the value of a physical quantity such as air kerma rate or particle fluence rate of which primary standards usually exist, is determined by a reference instrument at a reference point in the radiation field used for calibration. Second, the value of the appropriate operational quantity is determined by application of a conversion coefficient relating the physical quantity to the radiation protection quantity. Conversion coefficients used to determine operational quantities for neutrons and photons were evaluated by international committees and finally accepted for general use by international agreements (see Sect. 6.12, 6.3 and 6.4). Third, the device being calibrated is placed at this reference point to determine the response of the instrument to the operational quantity, e.g. the personal, ambient or directional dose equivalent or their corresponding rates. While area dosimeters are generally calibrated free in air, personal dosimeters are always calibrated in front of a standardized phantom (details see Sect. 4.5.3.4). The primary physical quantity used to specify a photon radiation field is exposure or air kerma, and the primary standard instruments used for its measurement are air-filled ionization chambers. For photon energies up to about 150 keV, mostly a free-air chamber is used as a standard instrument to measure air kerma. For higher photon energies, air-equivalent walled cavity chambers are generally employed. Properties of radiation fields used for the calibration of photon dosimeters are described in ISO standard 4037 [96ISO1]. Calibrations of dosimeters and survey instruments for the measurement of beta radiation are performed using standard reference beta sources as specified in ISO standard 6980 [96ISO2]. Determination of the conventional true value of the absorbed dose, and hence the directional dose equivalent, is achieved with a thin-window extrapolation ionization chamber [97Amb]. The primary quantity measured for neutrons is the fluence. In monoenergetic neutron fields the fluence is measured either directly by a reference instrument (e.g. proton recoil telescope, proportional counter or Long Counter) or by applying the associated particle method. As regards radioactive neutron sources, e. g. 252Cf without or with a surrounding D2O-sphere, the neutron fluence is determined from the source emission rate which is usually determined from comparative activation measurements performed Landolt-Börnstein New Series VIII/4
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by a national standards laboratory. The emission rate is then used to compute the neutron fluence or fluence rate. In addition, the neutron energy spectrum must be known. With the known spectral fluence mean conversion coefficients can be calculated and applied to determine the ambient dose equivalent at a reference point [98ISO] (see Sect. 6.4). The calibration of a personal dosimeter or area survey meter is not complete without the calibration being documented. National regulations often specify the details and format of both calibration records and certificates, as well as the frequency of calibration and the period of time for which calibration records are to be kept. The following list gives a general guideline for calibration records or certificates. A certificate should include: 1. 2. 3 4. 5. 6. 7. 8.
Date and place of calibration, Description of dosimeter or instrument (type and serial number), Owner of device, Descriptions of reference radiation sources and standard instruments, Reference conditions, calibration conditions or standard test conditions, Results with statement of uncertainties, Names of the person who performed the calibration and of the reviewer, Any special observations.
10.2.5 Requirements for individual monitoring of external exposure The basic requirement for personal dosimeters are to provide a reliable measurement of the appropriate quantities, i.e. Hp(10) and Hp(0.07) for almost all practical situations, independent of type, energy and direction of incidence of the radiation and with prescribed overall accuracy. These basic requirements can be expressed in terms of operational and technical parameters influencing the performance of the dosimeter, e.g. its response to radiation type, spectral and directional distribution, environmental influences and practical aspects. The most important ones are described below. The following minimum requirements apply to all types of personal dosimeters: • • • • • • • • • •
convenient in size and shape, low weight, easy to wear inexpensive mechanically robust easy to handle adaptable to various applications (assessment of whole body dose or extremity dose) a broad range of doses should be measured response should be reasonably independent of the radiation energy and dose rate response should not be strongly influenced by normal changes of environmental conditions (minimum temperature range 10 °C to 40 °C, relative humidity: 10 to 90 %) the measured dose should not be influenced by other unconsidered types of radiation the dosimeter reading should be independent of any delay between irradiation time and time of evaluation.
Electronic personal dosimeters are usually capable to measure dose and dose rate and additionally include an immediate warning capability. For extremity dosimeters where, due to the close proximity to the source large variations in the dose rate may occur, small-sized detectors such as detectors kept in finger rings are required (cf. Fig. 10.16). Detailed technical requirements are specified for detectors suitable for measuring whole body exposures and partial exposures in [94EU]. Some are summarized in Tables 10.6 and 10.7.
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Table 10.6. Technical requirements for measuring whole body exposures Photons Neutrons Measurement range Max. range Measurement range Max. range Dose range 0.2 mSv to 1 Sv up to 10 Sv 0.2 mSv to 1 Sv up to 10 Sv 10 keV 10−3 keV to 1.5 MeV up to Energy range(s) 15 keV to 250 keV or to or 100 MeV 70 keV to 1.5 MeV 10 MeV 1.5 MeV to 15 MeV Dosimeter 0° to ± 60° 0° to ± 180° 0° to ± 60° 0° to ± 180° orientation Table 10.7. Technical requirements for measuring partial body exposures Photons Beta radiation Measurement range Max. range Measurement range Max. range Dose range 1 mSv to 10 Sv 0.1 mSv to 10 Sv 1 mSv to 10 Sv 0.1 mSv to 10 Sv 15 keV to 1.5 MeV 10 keV to 10 MeV 0.2 MeV to 0.5 MeV1) 0.06 MeV to Energy range 1.0 MeV1) Dosimeter 0° to ± 60° 0° to ± 180° 0° to ± 60° 0° to ± 180° orientation 1)
mean beta energy
10.2.5.1 Operational requirements Two principal questions have to be answered before a measurement of Hp(10) and Hp(0.07) can be interpreted for radiation protection purposes. The personal dosimeter has to be worn at a “representative” place on the surface of the body and it will measure the dose at this point for a predefined period of time. In many practical cases the radiation field will be inhomogeneous and multidirectional. The value measured will therefore often depend on the orientation of the body in the field. While the use of several dosimeters could in principle improve the situation and lead to a more representative assessment of the effective dose, mostly one dosimeter measuring Hp(10) is sufficient for routine monitoring. In order to measure Hp(10) for assessment of effective dose, a personal dosimeter is usually worn in front of the body between the shoulders and the waist. Dosimetry of the skin dose or an extremity dose can be performed by using finger batches worn on the hand as rings. An estimate of the eye equivalent dose is obtained by wearing a whole body dosimeter placed at the collar. This is a reasonable location to measure both the eye and the whole body dose. Alternative locations for dosimetry may be necessary in the course of certain types of work. An example is the use of lead aprons in X-ray applications. In this case two dosimeters may be necessary, one under the apron to measure Hp(10) and one at the collar to measure the dose to the head. Other situations may necessitate relocation of dosimeters including fetal monitoring where the monitor should be placed in front of the abdomen to assess the uterus or fetal dose. In many practical cases of routine monitoring, dosimeters will be worn over a period of one month. In low dose environments this period could be extended up to 6 months. In cases with highly variable radiation fields and in situations where there are indications that dose limits could be reached or exceeded, an evaluation of the dose within a shorter time period could be required. If for operational reasons daily monitoring is required, a direct reading dosimeter with sufficient sensitivity should be used in addition to the routine dosimeter. Direct reading dosimeters are frequently used to monitor the dose received during a particular task, e.g. one working day or one shift. In the case of a pregnant woman an electronic dosimeter would be an appropriate way of individual monitoring (cf. Fig. 10.13).
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10.2.5.2 Accuracy requirement Basic rules in the description of uncertainties in measurements are given in a joint document of BIPM/ /IEC/ISO/OIML [95ISO]. The errors and uncertainties in the use of monitoring to provide estimates of individual doses and intakes result from the measurements and from the models used to link the measured and the required quantities. Different types which contribute to the overall uncertainties can be distinguished: random uncertainties due to counting statistics, systematic errors due to calibration errors and errors in dosimetric and metabolic models and errors in practical application of the models. For most assessments, the systematic errors in modelling result in a bias towards over-estimation of the true dose. Basic recommendations on the acceptable uncertainty in routine individual monitoring are given by ICRP Publication 60 (par. 271) [91ICR1], Publication 75 [97ICR] and by ISO [00ISO2]: for annual doses of the order of the relevant annual limit, the apparent annual dose to an individual as indicated by routine dosimeters should not differ from the annual dose equivalent indicated by an ideal dosimeter by more than –33 % or +50 % at the 95 % confidence level. The 95 % confidence level means that the given requirement must be fulfilled for 19 of 20 different measurements. For dose values equal to or close to the annual dose limit, e. g. 20 mSv (see 10.2.3.1), the relation between the measured and the true value may thus vary between 1.5 and 1/1.5 times this value in 19 of 20 different measurements. For individual doses much below the annual limit the accuracy requirements are less than the values given above. For characterising the acceptable uncertainty in performance tests of individual dosimeters trumpet curves as proposed by Böhm et al. [90Boe] have been defined by ISO [00ISO2] describing the requirement by an interval for Hm/Ht (measured dose value/conventionally true dose value) as a function of dose. A detailed overview on requirements for photon dosimeters and dosimetry services as published in the various international recommendations and standards is given by Ambrosi et al. [98Amb, 01Amb]. As example, Fig. 10.26 shows the recommendation of the European Commission [94EU] and the IAEA [97IAE] for photon dosimeters. While many countries, e.g. Germany, Italy, Sweden and Switzerland, use this procedure in performance tests, other countries, e.g. Spain, UK and USA, uses a statistical evaluation of the measured values together with criteria for a bias setting [01Amb]. 2.0
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Fig. 10.26. Requirements in performance tests of individual dosimeters for photons showing upper and lower limits for the ratio of the measured dose to the true dose value, Hm/Ht, as a function of dose [90Boe, 98Amb] where 95 % of all measured values must be within these limits. Full lines: monthly monitoring period; broken lines: bi-weekly monitoring periods. (a) limits for Hp(10); (b) limits for Hp(0.07).
The accuracy requirements for individual dosimeters for neutrons are mostly less than those for photon dosimeters because of the difficulties in realising a dosimeter response sufficient independent of neutron energy (cf. Sect. 10.1.7). Often it may be necessary to use some information about the spectral Landolt-Börnstein New Series VIII/4
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distribution of the neutrons in order to apply an approximate calibration factor to the dosimeters used. In situations where the contribution of the neutron dose is substantial and where the total dose may approach the dose limit, the radiation field should be characterised by application of more sophisticated equipment, e.g. neutron spectrometry, to get better information about the neutron spectrum and its directional dependence.
10.2.6 Personal dosimeters for individual monitoring in different radiation fields The most abundant types of detectors used for the purpose of personal dosimetry in routine monitoring are the film dosimeter, the photoluminescence (glass) dosimeter (PLD), thermoluminescence dosimeters (TLD) and ionization chambers. Also electronic dosimeters, based e.g. on GM-counters, proportional counters or semiconductor detectors (see Sect. 10.1), are used. Film dosimeters and ionization chambers are used to monitor X-, γ- and high energy β-ray exposures by measuring Hp(10). Glass and TLD´s are able to measure both Hp(10) and Hp(0.07). They are used to monitor β-, γ-, X-rays, and neutron radiations. Etched-track detectors and bubble detectors measuring Hp(10) are mainly sensitive to neutrons only. Small TLD badges, e.g. finger badges are designed to be worn on the finger to record the dose to the hand. They are sensitive to X- and γ-rays and high energy beta rays. A detailed description of the various detector types is given in Sect. 10.1. In many practical situations where task monitoring is required in environments with the possibility of elevated radiation levels, “active” dosimeters are needed which include immediate warning capabilities, e.g. electronic dosimeters. Advanced methods of active neutron dosimetry in mixed radiation fields are of particular importance. Potential fields of application of active dosimeters of this kind are at nuclear power plants and particle accelerators. 10.2.6.1 Photon dosimetry In most work situations with exposures by strongly penetrating electron/photon radiation, an estimate of Hp(10) can be obtained from a single dosimeter sensitive to electrons and photons. The overwhelming share of occupational exposure is caused by photons. Mostly film or TL dosimeters are used in routine monitoring, sometimes also PLD´s are in use [01Bar]. Since 1980, the application of TLD´s has increased considerably. For partial body dosimeters nearly always TLD´s are used. In a few cases where the workers doses are at or near the limits, it may be necessary to obtain additional information about the exposure conditions, e.g. from field measurements at the workplace to better estimate the effective dose equivalent. Film dosimeters (Sect. 10.1.6) as compared with solid state dosimeters involve a somewhat greater uncertainty of measurement in the lower dose range (<0.4 mSv) in the case of very hard gamma radiation. Test measurements have shown that all three dosimeter types fulfill the performance requirements mentioned in Sect. 10.2.5.2 (see Fig. 10.27). If well designed the dosimeter types are suitable for dose measurements up to photon energies of 15 MeV. Their response to neutrons with energies up to about 1 MeV is usually very low, i.e. they measure the photon dose independent of the neutron dose.
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2.0
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Fig. 10.27. Results of test measurements with different personal dosimeters. The ratio of the measured dose to the true dose as a function of photon dose is shown [01Amb]. The lines mark the accuracy limits for a monthly monitoring period. ▲ Film dosimeter ● Thermoluminescence dosimeter ♦ Photoluminescence dosimeter.
10.2.6.2 Beta dosimetry In most cases the dose generated by weakly penetrating beta radiation is a partial body dose of the skin of uncovered extremities. Although the depth of the sensitive layer of the skin vary between individuals and over the body of individuals, Hp(0.07) is considered to be a reasonable quantity to apply for the assessment of doses if the dosimeter is worn at a position representative for the exposure. Monitoring of weakly penetrating radiation is predominantly achieved by partial body dosimeters with thermoluminescence probes. In the case of hard beta radiation, e.g. from 90Sr/90Y or 204Tl, the dose is detected with a sufficient degree of reliability with film dosimeters. Sensitivity and energy-dependence of the film are sufficient for all practical purposes even in the low dose range. For a detection of beta radiation with intermediate energies above 100 keV the film dosimeter is, in principle, well suited. In practice, most radiation fields are mixed photon/beta fields and when measuring Hp(0.07) correctly it is generally difficult to obtain a separate assessment of the fractions of the dose from photon and beta radiation. In this case an interpretation of the measured data must be based on additional information by the licensee. The detection of weakly penetrating beta radiation to the skin of the hands with standard finger dosimeters is not always satisfactory. Many of these dosimeters are intended to be used for measuring photon radiation. They may considerably underestimate exposures from beta radiation. Sensitive thin layer thermoluminescence dosimeters need to be used for this purpose. In most practical situations the skin will be exposed to weakly penetrating radiation together with strongly penetrating radiation and an estimate of the skin dose will have to take account of both types of radiation. 10.2.6.3 Neutron dosimetry Personal dosimeters for neutrons have not yet reached the quality standards of photon dosimetry. This is mainly because their sensitivity and their variation of response with neutron energy and with the angle of radiation incidence are unsatisfactory. The nuclear track film which has frequently been used is only suited for fast neutrons with energies >1 MeV. A special type of thermoluminescence probes, e.g. the albedo dosimeter, is suitable to measure both the photon and the neutron dose. The photon energy range of this dosimeter type is 15 keV to 10 MeV, for neutrons from thermal neutrons up to 20 MeV while the response to neutrons, however, is strongly decreasing with neutron energy (see Sect. 10.1.7). A reasonable approach for individual neutron dosimetry is to use more than one type of detector to cover the whole energy spectrum, e.g. an Albedo dosimeter for neutrons in the low energy region together with a solid state etched-track dosimeter to cover the energy range above approximately 100 keV Landolt-Börnstein New Series VIII/4
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(cf. Sect. 10.1.7). Even with this detector system neutrons with intermediate energies may not be measured to full satisfaction. More recently, electronic personal dosimeters for neutrons based on semiconductor devices became available which partially improve the situation. In cases where the neutron dose contributes significantly to the total dose and the total dose is likely to approach dose limits, a more elaborate approach may be necessary. In such situations the use of area monitors and of neutron spectrometers is recommended to better characterize the radiation field. 10.2.6.4 Dosimetry in mixed field situations (photons and neutrons) In a mixed field situation with photons and neutrons the personal dose equivalent, Hp(10), includes the contributions of both photon and neutron dose. In mixed field situations an improvement of the measuring methods is still required. There are several ways to estimate the total dose and some commercial devices exist to perform appropriate measurements. Either two dosimeters are used each of them sensitive to photons or neutrons only, or one detector which measures the total dose directly. But often the energy response of these devices fails where their physical properties would be attractive, i.e. inside the containment of nuclear power plants, where low energy neutrons dominate. Bubble detectors for neutrons are available on the market but in the attempt to make them robust their unique response properties have been neglected. Tissue equivalent proportional counters (TEPC), which measure the total dose equivalent in a mixed field rather than being just neutron dosimeters have been developed but without real breakthrough on the market. There have been ongoing developments in recent years [99Alb] and a few most promising monitoring techniques may be ready for routine application in near future. The TEPC personal dosimeter can determine neutron dose equivalents down to 10 µSv with sufficient accuracy. The dosimeter offers the option of a detailed quantification of any radiation exposure in terms of a microdosimetric spectrum. Combined neutron and photon dosimeters of very small dimension, e.g. ionization chambers with direct ion storage (DIS) are recommended for application in radiation fields with high contributions of neutron doses and in places where light-weight and small dimensions are important because of the type of work performed. Dosimeters based on superheated drop detectors reveal a high sensitivity to neutrons with no sensitivity to photons.
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10.2.7 References for 10.2 77ICR 85ICR 90Boe 91ICR1 92ICR 93ICR 94Alb
94EU 95ISO 96EU 96ICR 96ISO1
96ISO2 97Amb 97IAE 97ICR 97ISO
98Amb 98ICR 98ISO
International Commission on Radiological Protection: ICRP Publication 26. Oxford, UK: Pergamon Press, 1977. International Commission on Radiation Units and Measurements: ICRU Report 39, Bethesda, MD: ICRU Publications, 1985. Böhm, J., Ambrosi, P.: Mandatory type tests of solid state dosimetry systems as an appropriate aid to quality assurance in individual monitoring. Radiat. Prot. Dosim. 34 (1990) 123-126. International Commission on Radiological Protection: ICRP Publication 60. Oxford, UK: Pergamon Press, 1991. International Commission on Radiation Units and Measurements: ICRU Report 47. Bethesda, MD: ICRU Publications, 1992. International Commission on Radiation Units and Measurements: ICRU Report 51. Bethesda, MD: ICRU Publications, 1993. Alberts, W.G., Böhm, J., Kramer, H.M., Iles, W.J., McDonald, J., Schwartz, R.B., Thompson, I.M.G: International standardisation of reference radiations and calibration procedures for radiation protection instruments. Proc. German-Swiss Radiation Protection Association Meeting 1994, Karlsruhe, 1994. European Commission: Radiation Protection 73, Technical recommendations for monitoring individuals occupationally exposed to external radiation. EUR 14852 EN, EC, Luxembourg, 1994. International Organization for Standardization: Guide to the expression of uncertainty in measurement. Geneva, Switzerland: ISO, 1995 (corrected of first print in 1993). Council Directive 96/29/EURATOM of 13 May 1996 laying down basic safety standards for the protection of the health of workers and the general public against the dangers arising from ionising radiation; EC Journal Series L 159 (1996) International Commission on Radiological Protection: ICRP Publication 73. Oxford, UK: Pergamon Press, 1996. International Organization for Standardization: X and gamma reference radiations for calibrating dosemeters and dose rate meters and for determining their response as a function of photon energy. Part 1: Radiation characteristics and production methods. ISO/4037-1, Geneva, Switzerland, 1996. International Organization for Standardization: Reference beta-radiations for calibrating dosimeters and dose rate meters and for determining their response as a function of betaradiation energy. ISO/6980, Geneva, Switzerland, 1996. Ambrosi, P.: Improved beta secondary standard; PTB-News 97.1 (1997) 3 IAEA Safety Series: Draft safety guide: Assessment of occupational exposure to external radiation. NENS-12, IAEA, Vienna, 1997. International Commission on Radiological Protection: General principles for radiation protection of workers. ICRP Publication 75. Oxford, UK: Pergamon Press, 1997. International Organization for Standardization: X and gamma reference radiations for calibrating dosemeters and dose rate meters and for determining their response as a function of photon energy. Part 2: Dosimetry for radiation protection over the energy range 8 keV to 1.3 MeV and 4 MeV to 9 MeV. ISO/4037-2, Geneva, Switzerland, 1997. Ambrosi, P., Bartlett, D.: Dosimeter characteristics/ Dosimeter and service performance requirements. PTB-Dos-27, PTB, Braunschweig, 1998. International Commission on Radiation Units and Measurements: Conversion coefficients for use in radiological protection against external radiation. ICRU Report 57, Bethesda, MD: ICRU Publications, 1998. International Organization for Standardization: Reference neutron radiations - Part 3: Calibration of area and personal dosimeters and the determination of their response as a function of neutron energy and angle of incidence. ISO/8529-3 Geneva, Switzerland, 1998.
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99IAE 99ISO
99Zan 00Die 00ISO1 00ISO2 01Amb 01Bar 01ISO
10 Measuring techniques Alberts, W.G., Arend, E., Barelaud, B., Curzio, G., Decossas, J.L., d´Errico, F., Fiechtner, A., Grillmaier, R., Meulders, J.-P., Menard, S., Roos, H., Schuhmacher, H., Thevenin, J.-C., Wernli, C., Wimmer, S.: Advanced methods of active neutron dosimetry for individual monitoring and radiation field analysis (ANDO), Report PTB-N-39, Braunschweig, 1999. Generic procedures for monitoring in a nuclear or radiological emergency, IAEA-TECDOS1092, Vienna, 1999. International Organization for Standardization: X and gamma reference radiations for calibrating dosemeters and dose rate meters and for determining their response as a function of photon energy. Part 3: Calibration of area and personal dosemeters and the measurement of their response as a function of energy and angle of incidence. ISO/4037-3 Geneva, Switzerland, (1999). Zankl, M.: Personal dose equivalent for photons and its variation with dosimeter position. Health Phys. 76 (1999) 162. Dietze, G.: Dosimetric concepts and calibration of instruments, IRPA 2000, Hiroshima, May 2000. International Organization for Standardization: Reference neutron radiations - Part 2: Calibration. Fundamentals of radiation protection devices related to the basic quantities characterizing the radiation field. ISO/8529-2 Geneva, Switzerland, 2000. International Organization for Standardization: Radiation protection – criteria and performance limits for the periodic evaluation of personal dosemeters for x and gamma radiation. ISO/14146, Geneva, Switzerland, 2000. Ambrosi, P.: Dosimetric performance requirements for the routine dose assessment of external radiation. Radiat. Prot. Dosim. 96 (1-3) (2001) 67-72. Bartlett, D.T., Böhm, J., Hyvönen, H. (eds): Individual monitoring of external exposure. Proc. Europ. Workshop 2000. Radiat. Prot. Dosim. 96 (1-3) (2001). International Organization for Standardization: Reference neutron radiations - Part 1: Characteristics and Methods of Production. ISO/8529-1 Geneva, Switzerland, 2001.
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10.3 Radiological protection measurements: Internal exposure 10.3.1 Measurement of radon and its progeny Radon (222Rn) and thoron (220Rn) are gaseous radionuclides in the U- and Th-decay chain, respectively, occurring naturally in the ground and escaping from there into air and water (see Chapter 11). Their decay products are metallic radionuclides (see Tables 10.8 and 10.9). In air usually a mixture of radon/thoron and short-lived radon/thoron progenies exist. The progenies are mostly attached to aerosols with sizes of about 0.01 - 10 µm in diameter. Some (few percent only), however, are “non-attached” (cluster <0.005 µm in diameter). Due to the short half-life of 55 s of thoron this nuclide and its decay products are less important for dose from inhalation than radon (T1/2 = 3.825 d) and its short-lived progeny. In special situations, however, they may contribute to the total potential alpha energy concentration from radon and thoron progenies up to 50 %. While the inhaled radon is mostly exhaled again, the progeny are deposited in the respiratory tract where their decay by alpha-particle emission is seen to be most relevant for the dose to the lung and hence for lung cancer induction. Their contribution to the dose is generally 2 to 3 orders of magnitude greater than that of 222Rn. In measurements, however, the radon is given the highest interest, because action levels and reference values are nearly always specified in terms of radon concentration (in Bq m−3) or radon exposure (in Bq m−3 h) and not in terms of dose quantities (see Sect. 4.8). If doses wanted to be specified, the radon progeny concentrations in air or the equilibrium factor F need to be determined. For radon in homes often an equilibrium factor F = 0.4 is applied, if no measurements are available. Table 10.8. 226Ra decay chain with radon and its progeny (data from [98NN]). Radionuclide Half-life T1/2 Radiation energy and relative emission probabilities α-particles β-particles(1) γ-rays MeV (%) MeV (%) MeV (%) 226 4.59 (4.16) 0.186 (3.51) Ra 1 600 a 4.78 (94.5) 222 Rn 3.825 d 5.49 (100) 218 Po 3.10 min 6.00 (100) 214 Pb 26.8 min 0.67 (48) 0.242 (7.4) 0.73 (42) 0.295 (19.3) 1.02 (6) 0.359 (37.6) other (4) 214 Bi 19.9 min 1.00 (23) 0.609 (46.1) 1.51 (40) 1.120 (15.1) 3.26 (19) 1.764 (15.4) other (18) 214 Po 7.69 (100) 164 µs 210 Pb 22.3 a 0.015 (81) 0.047 (4.05) 0.061 (19) 210 Bi 5.013 d 1.161 (100) 210 Po 138.4 d 5.30 (100) 206 Pb stable (1)
The energy given is the maximum energy of β-particles emitted in the specific decay channel.
Specific quantities have been defined taking care of the complex decay chain of radionuclides (see Sect. 3.4.1) and the importance of the progeny for internal dosimetry. While in measurements of radon/thoron the actual activity concentration or its mean value over a longer period is usually determined, for the progeny the potential alpha energy concentration (PAEC), the equilibrium equivalent concentration (EEC), the equilibrium factor and the potential alpha energy exposure are measured (see Sect. 4.6).
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For radon the concentration in air above ground, in the air, in the ground, and in water from the ground are of interest, while the progeny are important only when produced in air above ground. In buildings often the exhalation rate of radon from the ground or wall materials is looked at. This can be determined by measuring the increase of the radon concentration in a closed volume after it has been well ventilated. The diffusion of radon in air is very fast and its concentration in a closed room is, therefore, usually homogeneous. Mostly there is no equilibrium with its progeny because of a continuous deposition of aerosols on the walls and other surfaces. Often an equilibrium factor between 0.3 and 0.7 is achieved. Local radon concentrations in air, however, may strongly vary with time depending on environmental parameters like weather conditions, pressure, wind, temperature as well as ventilation in rooms. Care must, therefore, be taken when using data from short time measurements. Also the positioning of instruments for measurement of radon needs attention in order to avoid erroneous results. In rooms detectors should usually not be positioned near windows or doors or directly on a wall. For thoron, the situation is generally different. Free in air, there is mostly a strong decrease of the thoron concentration with height above ground due to the short half-life of 55 s, and equilibrium with thoron progeny is never achieved. Table 10.9.
228
Th decay chain with thoron and its progeny (data from [98NN]).
Radionuclide
Half-live T1/2
228Th
1.913 a
224
3.66 d
Ra
220
Rn Po 212 Pb
55.6 s 0.15 s 10.64 h
212
Bi
60.6 min
212
Po Tl
304 ns 3.04 min
Pb
stable
216
208
208
Radiation energy and relative emission probabilities α-radiation β-radiation(1) γ-radiation MeV (%) MeV (%) MeV (%) 0.0837 (1.2) 5.34 (27.0) 0.216 (0.25) 5.43 (73.0) 5.45 (4.9) 0.241 (4.1) 5.68 (95.1) 6.29 (100) 0.55 (0.1) 6.78 (100) 0.331 (83) 0.239 (43.3) 0.569 (12) 0.300 (3.3) other (5) 6.05 (25) 1.55 (5) 0.040 (1.1) 6.09 (10) 2.26 (55) 0.727 (6.6) other (40) 1.620 (1.5) 8.78 (100) 1.28 (23) 0.511 (22.6) 1.52 (22) 0.583 (84.5) 1.80 (51) 0.860 (12.4) other (4) 2.614 (99.1)
(1) The energy given is the maximum energy of β-particles emitted in the specific decay channel.
Radon/thoron and its short-lived progeny form a decay chain and the relation of their activity concentrations is generally described by a set of differential equations (see Sect. 3.4.1.4). If, for example, a closed chamber is filled only with radon with an activity A at a time t = 0, the activity of the progeny will increase from 0 Bq until after about 3 h it reaches an equilibrium (F = 1), where the activities of the progeny equal that of radon (see Fig. 10.28). Another example may be the build up and decrease of the activity of radon progeny absorbed in a filter flowed by air during a fixed period. In this case, the activities of the decay products on the filter are described by a set of differential equations: dAi dt = ciV& + λi ( Ai-1 − Ai )
i=1-4
(10.3.1.1)
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with Ai activity of radionuclide i on the filter in Bq, ci activity concentration of radionuclide i in air in Bq m−3, V& air flow through the filter in m3 h−1 and λi decay constant of radionuclide i in h−1. Fig. 10.31 shows a typical example. The same equations can be used when the activity A is exchanged by the number of the corresponding particles N. 1200 1000
1
Activity [Bq ]
800 3
2
4
600 400 200
0
30
60
120 90 Time [min ]
150
180
Fig. 10.28. Example for the decay of radon and built up and decay of radon progeny in a closed chamber. Activity at t = 0: ARn-222= 1000 Bq, APo-218 = APb-214 = ABi-214 = APo-214 = 0 Bq. (1) 222Rn, (2) 218Po, (3) 214Pb , (4) 214Bi and 214Po.
Different types of measurements are performed in the assessment of radon and radon progeny concentrations. Measurements within a period of up to 2 days are called short-time measurements. They are performed by either a sequence of single short measurements, by a continuous measurement or by a measurement with an integrating device. Such measurements provide the actual radon concentration at a place of interest but are usually not sufficient for the estimation of a representative mean annual or monthly value because of the possible environmental variations. Long-term measurements ranging for some months or a year are usually performed with integrating passive detectors which at the end deliver a value integrated over the selected period and hence a mean radon concentration which can be related to reference values, action levels or annual limits. Continuous measurement means that an instrument is continuously measuring and periodically delivering data where the smallest time period depends on the instrument and the accuracy needed. Those measurements are performed with active electronic devices and are used if short-time or daily variations are investigated. While radon and thoron are mainly emitting α-particles, their various progenies decay by emission of α- or β-particles accompanied by γ-rays (see Tables 10.8 and 10.9). Therefore, depending on the aim and type of measurement very different detector systems are applied, ranging from simple and cheap passive ones e.g. for screening measurements in houses up to detectors with complex electronic devices used as reference instruments. In the following an overview is given on the measurement techniques and devices in use for detection of radon and its progeny (see also [88NC, 88Naz, 02SSK]). For thoron similar instrumentation may be applied if care is taken of the short half-life of thoron which limits its diffusion time and, for example, has the effect that a strong decrease of the thoron concentration exists with increasing height above ground. In any way, care must be taken that devices may be sensitive to both radon and thoron and their progenies, and then the influence of thoron on measurements of radon should be estimated. Often a sufficiently long diffusion time through a filter can avoid such problems.
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10.3.1.1 Measurement of radon in air General Radon measurements in air are performed in homes or dwellings or at work places as well as outside in areas where the exhalation from the ground may be high. Most methods for the measurement of the radon concentration in air are based on the principle that radon is entering a closed chamber or is adsorbed in carbon material, while the radon/thoron decay products in air which would strongly influence the measurement are absorbed before entering the system by a filter. This is realized either by pumping air through the filter to the chamber or by diffusion of radon through the filter. The radiation from radon and its progeny from the decay of radon in the chamber or in the carbon absorber are then contributing to a detector signal. If the diffusion time is longer than some minutes, the contribution from thoron to the detector response is small due to its short lifetime (T1/2= 55 s). Depending on the detector device either the charge produced in the chamber by ionisation is measured or α-particles from both radon and its progeny or β-particles and γ-rays from its progeny are detected individually. Passive integrating systems Passive integrating systems are those which do not allow a continuous or quasi continuous read out but deliver a value of the radon concentration integrated over a longer time period (in Bq m−3 h). These are either systems with a passive detector, e.g. a detector chamber with an etched track detector (see Sect. 10.1.7) which detects the α-particles emitted by the radionuclides in the chamber gas, or an electret detector, where the discharge of a charged electret by ionisations in the chamber gas is determined by a voltage measurement [88Kot], or systems with a radon adsorbing material (an activated carbon absorber) where radon is trapped on the active sites of the carbon beds, and afterwards the γ-radiation which is emitted from the radon decay products 214Pb and 214Bi produced in the absorber is measured by a gamma spectrometer (see Sect. 10.1.4) [84Geo, 90Geo]. All systems are equipped with a filter in front of the chamber for retaining radon progeny from entering the chamber (see e.g. Fig. 10.29). Because water vapor and also temperature influences the radon collection efficiency the activated carbon collector systems should be used indoors only. Table 10.10 gives an overview on different systems with passive detectors in use and provides some further information. More details about special systems available on the market are given in a review by George [96Geo] and a report on a European intercomparison of passive radon detectors [00How]. Depending on the diffusion time of radon through the entrance filter or plastic foil the system may also be sensitive to thoron or not. Integrating detectors are usually mailable and relatively cheap. They are often used for medium and long time measurements in dwellings where knowledge about radon concentration values averaged over 3 to 12 months are most important. The analysis of the detector response is mostly performed at a central laboratory and needs more expensive additional equipment.
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Cover Gasket Filter
Drying agent
Detector
Filter Carbon absorber
a
b
Fig. 10.29. Schematic diagram of passive devices for radon measurement in air; (a) chamber with activated carbon absorber; (b) diffusion chamber with either an etched track detector or an electret detector.
In principle, also active systems which allow to nearly continuously providing data may be used as an integrating device. For long term measurements, however, they become much more expensive. Table 10.10. Integrating systems with passive detectors for the measurement of radon in air Method
Measured radiation
Absorption of radon in γ-rays a carbon filter Absorption of radon in γ-rays or a carbon collector α-particles with entrance filter Diffusion chamber with entrance filter
α-particles
Diffusion chamber with entrance filter
charge from ionisations
Detector
Lower detection limit
γ-spectrometer
∼ 5 Bq m−3 in 0.5m3 air ∼ 10 Bq m−3 (3 h measurem.)
γ-spectrometer or liquid scintillation counter etched track ∼ 105 Bq h m−3 detector (CR-39, LR-115 etc.) electret detector ∼ 103 Bq h m−3
Influenced by air humidity temperature air humidity temperature
Main application short time measurement short time measurement up to 3 d
air pressure, temperature,
long time measurement 1-12 months long time measurement 1w-1a
air pressure, dose rate, γradiation
Active detector systems These systems allow a single measurement within a short time period or quasi continuous measurements to register temporal variations of the radon concentration. All systems are supplied with an entrance filter for absorption of the progeny from outer air. Usually, the α-particles emitted in the chamber (mostly from progeny produced in the chamber and deposited on the inner chamber wall) are directly measured (see Fig. 10.30). The use of a scintillation cell, where the walls are coated with a scintillating material (mostly silver activated ZnS(Ag) powder), is one of the eldest methods (Lucas chamber [57Luc]). Others are using silicon surface barrier detectors or diffused junction detectors for the detection of α-particles. In other systems the charge or charge pulses from ionisations in the chamber gas are measured [92Bal]. Table 10.11 gives an overview on different systems with active detectors in use and some additional information. More specific details on devices already used in practice are given by George [96Geo].
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Table 10.11. Active detector systems for the measurement of radon in air Lower detection limit
Method
Measured radiation
Detector
Ionisation chamber with entrance filter
α- and βparticles
Multi-wire ionisation chamber with entrance filter Scintillation chamber with entrance filter Chamber with entrance filter, electrostatic deposition Diffusion chamber with α-spectrometry Two filter method (also for thoron)
α-particles
∼ 5 Bq m−3 (103 cm3 volume, 3 h measurem.) ∼ 5 Bq m−3 (103 cm3 volume, 3 h measurem.) ∼ 20 Bq m−3 (250 cm3 volume, 3 h measurem.) silicon surface ∼ 5 Bq m−3 barrier detector (103 cm3 volume, 3 h measurem.) charge or charge pulse measurement charge pulse measurement, α-spectrometry ZnS(Ag) scintillator
α-particles α-particles
activity of 2nd filter
Pump
Main application
air humidity
single or continuous measurement single or continuous measurement single or continuous measurement single or continuous measurement
air humidity vibration
air humidity
air humidity silicon surface ∼ 100 Bq m−3 barrier detector (102 cm3 volume, 3 h measurem.) activity ∼ 10 Bq m−3 determination (105 cm3 volume, of filter, vari10 h measurem.) ous detectors
α-particles
Filter
Influenced by
single or continuous measurement single measurement
Drying filter Pump
P
Filter
P
Detector
Chamber wall
ZnS Quartz glass PM Amplifier
Amplifier
HV Multichannel analyser
a
Counter
b
Electrometer
c
Fig. 10.30. Schematic diagram of active devices for radon measurement in air; (a) Lucas chamber [57Luc] (PM: photomultiplier); (b) ionisation chamber with charge measurement (HV: high voltage); (c) diffusion chamber with semiconductor detector.
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10.3.1.2 Measurement of radon progeny in air Nearly all devices for the measurement of radon progeny are using the filter method which can in general be described by the following. A pump is sucking air through a filter where the aerosols with the radon progeny are absorbed. A flow meter is measuring the air volume flowing through the filter and the activity of the filter is determined by measuring the α-, β- or γ-radiation emitted from the progeny either quasi-continuously or after the collection period at different time intervals. As shown in Fig. 10.31 the activity of 218Po, 214Pb and 214Bi on the filter vary differently with time, while the activity of 214Po equals that of 214Bi because of the short lifetime (T1/2 = 164 µs). From these data the potential alpha energy concentration (PAEC in J m−3 or WL) or the activity concentration of the different decay products can be calculated using the set of differential equations (see equ. (10.3.1.1)). 14000 12000
Activity [Bq]
10000 8000 6000
Fig. 10.31. Build up and decrease of the activity of radon progeny deposited on a filter with 100 % collection efficiency which for 2000 s is flowed by 103 cm3 min−1 of air with a radon activity concentration of 1 kBq m−3 and an equilibrium factor F = 1. (A) 218Po (α-emission), (B) 214Pb (β, γ-emission), (C) 214 Bi (β, γ-emission) and 214Po (α-emission).
4000 A
B
C
2000 0
4000 6000 Time [s ]
2000
8000
10000
The sampling unit and the activity measurement unit are either separated or combined in one system (see Fig. 10.32). The absorbing filter should have a high efficiency. For β- or γ-radiation measurements a total absorption efficiency of >98 % is sufficient. For α-particle detection, however, it is additionally important that the progeny are adsorbed at the filter surface. This affords membrane filters with pore sizes of less than 3 µm. Detector
Filter
Detector
FM
P
a
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FM
P
Computer
b
Electronics
Computer
Fig. 10.32. Schematic diagram of devices for radon progeny measurements [96Por] (FM: flow meter, P: pump); (a) with separated sampling and detector systems; (b) with a combined sampling and detector system.
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The sensitivity of a system depends strongly on the power of the pump. Mostly pumps with flow rates from 0.1 m3 h−1 to 1 m3 h−1 are used, but for special short time measurements also pumps with flow rates up to 100 m3 h−1 are applied. Care must be taken that the aerosols are not already be deposited in the entrance region of the device in front of the filter. If a flow meter is included in the system, its position is important. On the one hand it should not be influenced by the pump (e.g. by vibrations), but on the other hand it should not influence the flow in front of the filter and the deposition of aerosols. Immediately behind the filter might be the best position. Determination of the activity concentration of the radon progeny Three independently measured total count numbers for decay by α-particles during different counting intervals after the sampling period can be used to calculate the progeny concentrations using equ. (10.3.1.1) [96Por]. This method has been developed by Tsivoglou [53Tsi] and was further improved by various groups [72Tho, 80Bug, 81Sco, 84Naz] in order to improve its application also for low count rates. Mostly ZnS(Ag) scintillation detectors or proportional counters are applied (see Sect. 10.1.3 and 10.1.2) but also other active detectors are in use. Raabe and Wrenn have developed a method measuring the total alpha particle count number in many successive intervals and fitting the obtained decay curve with a theoretically calculated decay function using the maximum likelihood method [69Raa]. As a result the three activity concentrations of the radon progeny are obtained. A further development is the use of α-spectrometry [69Mar, 80Por] or even α- and β-spectrometry [97Ruz] for the identification of the different radon progeny. Silicon surface barrier or diffused junction detectors are usually used with multichannel analysis. These methods are generally more complex, however, with α-spectrometry measurements in two intervals and with α- and β-spectrometry a single measurement is sufficient. It allows also a correction for the contribution of thoron progeny. In general, the detection limit is lower than in methods with gross alpha counting. The spectrometry allows also a continuous measurement, if the filter and the detector are combined in one unit. The continuous detection of α-particles from 218Po and 214Po are not sufficient to determine the concentrations of the 4 radon progeny, if no further information is available. Very often, however, an equal ratio of the activity concentrations cPo-218/cPb-214 = cPb-214/cBi-214 is fulfilled. Most algorithms for the calculation of concentrations from measured data assume constant progeny concentrations during the sampling period. If this is not the case, more complex calculations are necessary and the uncertainty will be higher. Determination of the potential alpha energy concentration cp (PAEC) In principle, PAEC can be calculated from measured activity concentration data of the progeny (see Section 4.6.2). Often, however, methods are used which are simpler in instrumentation and optimised with respect to direct PAEC determination. A simple method is based on gross α-counting during a single period [56Kus]. After a short sampling time, a measurement of the number of α-particles from 218Po and 214Po during some hours, where most of these radionuclides decay, multiplied with the mean α-energy provides a PAEC value with an uncertainty of at least 10 % due to the difference in the α-energies from 218Po and 214Po. For these measurements mainly active detector devices (e.g. solid state detectors, electret ionisation chambers or scintillation detectors) are used. If shorter counting periods are used, this needs a correction of the calibration factor and an assumption on the equilibrium status because of the different decay times of the progeny (see Fig. 10.3.1). For quasicontinuous measurements the filter and the detector must be combined in one unit. A short sampling time may then be followed by a counting time of about 1 hour. After about 3 measurement cycles a correct PAEC value is achieved. This may, however, not be the case, if the progeny concentrations vary strongly in time. Landolt-Börnstein New Series VIII/4
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Due to the short lifetime of 218Po (T1/2= 3.1 min) the α-particle measurement needs a precise timing. This is not the case, if instead β- or γ-radiation from 214Pb and 214Bi is measured. Then, however, corrections need to be applied for the influence of environmental background radiation. While the foregoing methods provide an actual PAEC value, often a value averaged over a long time period or the exposure Ep is wanted. In this case, in addition to the above mentioned active devices also passive detectors, e.g. etched track and TL detectors, are used. While etched track detectors are sensitive to α-particles only, TL detectors are also sensitive to beta and γ-radiation, e.g. from environmental background. In this case, two TL-detectors are applied, where one measures the background only. The thoron progenies are also deposited on the filter. Especially the long-living progeny 212Pb (T1/2= 10.64 h) may therefore be enriched on the filter during long sampling times and will together with its progeny 212Bi influence the PAEC measurement. This can be checked when an additional measurement of the filter is performed at about 10 h after the end of the sampling period, where the radon progeny are already decayed. Determination of the unattached fraction of the progeny fp A separate measurement of the unattached fraction of the progeny is based on the different diffusion properties. Due to the small size of the unattached progeny (<5 nm) compared to the aerosols they have a much higher diffusibility and are deposited on surfaces much faster than those. For measurements instead of a filter a diffusion battery – a system of small diffusion tubes or a wire screen – is used where mainly the unattached fraction is deposited [96Por]. The further measurement of its concentration or PAEC is similarly performed as described for the attached progeny. If a filter is deposited after the diffusion battery, the PAEC of the attached fraction can additionally be measured. The unattached fraction fp is then given by fp =
cpf c + cpf a p
(10.3.1.2)
where c pa is the PAEC of the progeny attached to aerosols and c pf is that of the unattached fraction.
Determination of the equilibrium equivalent concentration ce (EEC) and the equilibrium factor F. The equilibrium equivalent concentration (EEC) ce of radon for a non-equilibrium mixture with its progeny in air is the fictitious activity concentration of radon which is in radioactive equilibrium with its short-lived progeny and has the same PAEC as the actual non-equilibrium mixture. The equilibrium factor F is then defined as the quotient of ce and cRn (see Sect. 4.6.2). It is always ce ≤ cRn and hence F ≤ 1. ce and F are not directly measurable. A determination needs the measurement of the concentrations of radon and their progeny. While cPo-214 can be ignored (see Sect. 4.6.2), ce of radon and F are given by the equations ce = 0.106 cPo-218 + 0.513 cPb-214 + 0.381 cBi-214. F = ce / cRn The unit of ce equals that of cRn, but is often marked Bq m−3 (EEC).
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10.3.1.3 Measurement of radon in the ground and in water Radon and thoron are continuously produced in the ground by the decay of 226Ra and 224Ra, respectively. The radon/thoron is partially transferred from minerals to pores of air in the ground (emanation) and then it may diffuse to the surface and exhale to the atmosphere. The radon concentration in the ground air (effective radon concentration in the ground) is about 100 to 10 000 times higher than in the open air while for thoron the ratio is even higher. Near the surface (<1 m) the concentration decreases due to the exhalation at the surface. The actual exhalation rate is also influenced by the air pressure and the temperature in the atmosphere. The radon concentration in the ground can be indirectly determined by measuring the specific activity of 226Ra in the soil together with the determination of the emanation probability (usually 0.2 - 0.5 depending on the soil type and on the humidity of the ground). The 226Ra content in a soil sample can be measured by γ-spectrometry. If the sample is deposited for about 25 days in a closed box, equilibrium of radon with 226Ra is achieved and the total content of radon in the soil sample can also be determined by γ-spectrometry (γ-rays from 214Pb and 214Bi). The effective radon content in the sample can be determined when equilibrium is achieved by flowing radon-free gas through the chamber and adsorbing the radon in an activated carbon absorber afterwards. The radon content in the absorber is then determined by γ-spectrometry (γ-rays from 214Pb and 214Bi). The same can be achieved by measuring the radon content in the sample before and after the flow of gas through the chamber and taking the difference. A direct measurement of the effective radon concentration in the ground is performed by inserting a probe into the ground, at minimum 1 m below the surface in order to avoid surface effects. The probe may be a special diffusion chamber for local radon measurement (see Tables 10.10 and 10.11) or a system connected with a pump to collect a known amount of ground air into an external radon measuring device. In this case the ground air is often returned to ground in a closed loop. In any way care is needed to avoid atmospheric air entering the probe. Most of the methods can also be used for the measurement of the thoron activity concentration in the ground, if special care is taken considering the fast decay of thoron. It should also be looked at, if thoron is influencing the radon measurement. Larger diffusion times or waiting some time before measuring a probe may avoid these problems. The specific activity concentration of radon dissolved in water can be directly measured in a water sample (after 3 h when equilibrium exist between radon and its progeny) by detecting γ-rays from 214Pb and 214Bi with a γ-spectrometer. Also a liquid scintillation detector may be used. The radon of a water sample may also be exhaled with a radon-free gas, dried and either absorbed in a carbon filter or transferred to a detector chamber. The radon measurement is then be performed by methods given in Table 10.11. A fast and simple technique which avoids the sampling of water is the use of a diffusion chamber of a material (e.g. a membrane tube of polypropylene) which allows a fast diffusion of radon. The radon concentration can then be conventionally measured. Glass fibre filters of a certain brand were found to be very efficient for the adsorption of short-lived radon decay products during filtration of water [97Phi] where they are in equilibrium with radon. The β-radiation from 214Pb and 214Bi on the dried filter can then simply be measured using a proportional counter with a thin window.
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10.3.1.4 References for 10.3.1 53Tsi 56Kus 57Luc 69Mar 69Raa 72Tho 80Bug 80Por
81Sco 84Geo 84Naz 88Naz 88Kot 88NC 90Geo 92Bal 96Geo 96Por 97Phi 97Ruz 98NN 00How 02SSK
Tsivoglou, E.C., Ayer, H.E., Haladay, D.A.: Occurrence of non equilibrium atmospheric mixtures of radon and its daughters. Nucleonics 11(9) (1953) 40. Kusnetz, H.L.: Radon daughters in mine atmospheres. A field method for determining concentrations. Am. Ind. Hyg. Assoc. J. 17 (1956) 85. Lucas, H.F.: Improved low-level alpha scintillation counter for radon. Rev. Sci. Inst. 28 (1957) 680. Martz, D.E., Hollemann, D.F., McCurdy, D.F., Schiager, K.J.: Analysis of atmospheric concentrations of RaA, RaB and RaC by alpha spectroscopy. Health Phys. 17 (1969) 131. Raabe, O.G., Wrenn, M.E.: Analysis of the activity of radon daughter samples by weighted least squares. Health Phys. 17 (1969) 593. Thomas, J.W.: Measurement of radon daughters in air. Health Phys. 23 (1972) 783. Bugsin, A., Phillips, C.R.: Uncertainties in the measurement of airborne radon daughters. Health Phys. 39 (1980) 943. Porstendörfer, J., Wicke, A., Schraub, A.: Methods for a continuous registration of radon, thoron and their decay products indoors and outdoors, in: Gesell, T.F., Lowder, W.M. (eds). Natural radiation Environment III. CONF-780422 Vol. 2, DOE, Washington D.C., 1980, p. 1293-1307 Scott, A.G.: A field method for measurement of radon daughters in air. Health Phys. 41 (1981) 403. George, A.C.: Passive integrated measurement of indoor radon using activated carbon. Health Phys. 46 (1984) 867. Nazaroff, W.W.: Optimizing the total alpha three count technique for measuring concentrations of radon progeny in residences. Health Phys. 46 (1984) 395. Nazaroff, W.W., Nero, A.V.: Radon and its decay products in indoor air. New York, Chichester, Brisbane, Toronto, Singapore: John Wiley & Sons, 1988. Kotrappa, P., Dempsey, J.C., Hickey, J.R., Stieff, L.R.: An electret passive environmental 222 Rn monitor based on ionization measurement. Health Phys. 54 (1988) 47. National Council on Radiation Protection and Measurement: Measurement of Radon and Radon Daughters in Air. NCRP Report No. 97, Bethesda, MD, 1988. George, A.C., Webber, T.: An improved passive activated carbon collector for measuring environmental 222Rn in indoor air. Health Phys. 58 (1990) 583. Baltzer, P., Gorsten, K.G., Backlin, A.A.: A pulse counting ionization chamber for measuring the radon concentration in air. Nucl. Inst. Meth. Phys. Res. A317 (1992) 357. George, A.C.: State-of-the-art instruments for measuring radon/thoron and their progeny in dwellings – a review. Health Phys. 70 (1996) 277. Porstendörfer, J.: Radon: measurements related to dose. Environ. Int. 22, Suppl. 1 (1996) 563. Philipsborn, H. von: Efficient adsorption of waterborn short-lived radon decay products by glas fiber filters. Health Phys. 72 (1997) 451. Ruzer, L., Sextro, R.: Measurement of radon decay products in air by alpha and beta spectrometry. Radiat. Prot. Dosim. 72 (1997) 43. NNDC: Nuclear Data, Decay Radiations. National Nuclear Data Center, Brookhaven National Laboratory, Upton, NY, 1998. Howarth, C.B., Miles, J.C.H.: Results of the 1998 European Commission intercomparison of passive radon detectors. European Commission, Report EUR 18835 EN, Luxembourg, 2000. Strahlenschutzkommission: Leitfaden zur Messung von Radon, Thoron und ihren Zerfallsprodukten. Veröffentlichungen der SSK, Band 47, München, Jena: Urban & Fischer, 2002.
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10.3.2 In vivo measurements 10.3.2.1 Introduction Internal exposures due to incorporation of radioactive materials may be detected both by in vivo and in vitro measurement: the in vivo measurement involves the measurement by detectors external to the body, thus mainly radionuclides emitting γ or X-rays can be detected by this way. The in vitro method is based on the measurement of the activity excreted with the urine, faeces and exhaled air and thus can be applied in principle for any material. In many cases the in vitro method is more sensitive but the interpretation of the results is in general more difficult because of the lack of information with respect to the individual metabolic behaviour of the incorporated materials. The in vivo method is not as sensitive as the in vitro method but the evaluation of the results is more easy because not that much information is needed for the assessment of dose from body or organ activity data. So the in vivo measurement is considered to be the best method for the detection of γ emitting radionuclides i.e. most of the fission and activation products and few actinides such as 235U or transuranium radionuclides such as 241Am, whereas the in vitro method is applied for the detection of all other radionuclides where the in vivo method is not sensitive enough. In few cases the in vivo technique can be applied also for α-emitters with γ emitting daughters, such as 226Ra or 238U. The first in vivo measurements of Radium have been carried out in 1927 by Blumgart and Weiss using ionisation chambers for blood flow studies [27Blu]. The lower detection limit of those measurements was reported to vary in the range from 5 to 100 µg (0.18 to 3.7 MBq) 226Ra, assuming radiological equilibrium between 226Ra and its γ-emitting daughter products [29Sch]. In 1931 Schlundt took into account the geometry and the self-absorption of the body by inserting sealed radon sources in a phantom, thus providing the first calibration standard for in vivo measurements [31Sch]. In those years three lethal cases of radium poisoning were established in the radium manufacturing industry in Germany, this enforcing the further development of detection procedures at the Max-Planck Institute for Biophysics in Frankfurt, Germany. So in 1938 the “Untersuchungsstelle für die physikalische Diagnostik der Radiumvergiftungen” was founded at the Kaiser-Wilhelm-Institut for Biophysics in Frankfurt. At the same time researchers in the United States applied high sensitive Geiger-Müller tubes also and the lower detection limit was reported to be 1 µg (37 kBq) 226Ra [37Eva]. Few years later Rajewsky and Dreblow performed first partial body measurements using a so-called gamma-ray stethoscope which enabled the diagnosing physician to contact directly the individual parts of the body thus allowing for the localization of radium deposits in the body [41Raj]. About one decade later, Sievert developed a 4π - geometry whole-body counter using an arrangement of 10 long ionisation chambers surrounding the subject. For achieving a low background, the measuring device was installed into granite rocks 50 m below ground surface, this resulting in a detection limit of about 1 nCi (37 Bq) of 226Ra for a measurement of 3 to 4 hours [51Sie, 57Sie]. In the fifties a lot of progress was achieved due to the development of new types of detectors such as NaI(Tl) scintillation detectors as well as liquid scintillation detectors [57And] and organic scintillation detectors [58Bir]. In 1956 the first technical conference on in vivo measurement was organised in Leeds in order to discuss the state of the art of measuring techniques. Since that time further improvements in the sensitivity and reliability of in vivo measurements have been achieved by the use of arrays of scintillation detectors and the application of anticoincidence techniques for the reduction of the detectors' background. In 1968 Laurer presented the first dual NaI(Tl)/Cs(Tl) scintillation crystal detector (phoswich) which was designed especially for low energy photon detection [68Lau]. In the late sixties semi-conductor detectors were introduced for in vivo measurement, starting with Lithium-drifted Germanium (Ge(Li)) detectors, which later on have been replaced by high purity Germanium (HPGe) detectors.
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With improving measuring techniques, whole-body counting has become a standard method for radiation protection. Personnel working in nuclear installations or in laboratories handling radioactive materials may receive internal radioactive contamination by inhalation, ingestion or intake through wounds. Techniques of in vivo measurement may be employed in the monitoring and control of such contamination, either as the sole means of evaluation or in conjunction with in vitro measurement of activity excreted by the subject or present in the working environment. 10.3.2.2 Requirements Techniques of in vivo measurement depend on the detection outside the body of photons originating from internally deposited radioactive materials. They are useful, as a means of controlling exposure to a given nuclide, if a significant deposit of that nuclide leads to a detectable signal at or near the surface of the body. In this context, a significant deposit would be one implying an intake at or exceeding some level requiring administrative action. Their feasibility in the case of a given nuclide thus depends on the yield and energy of photons emitted by the nuclide, on its pattern of deposition in the body and on the relevant limits on internal exposure. Many fission and activation products emit abundant penetrating gamma radiation, thus allowing for the assessment of intakes small in relation to annual limits, with relatively simple equipment. By contrast, long-lived α-emitters with only weak low-energy photon emissions may escape detection with the most sensitive and elaborate equipment, even when present in levels far exceeding annual limits on intake; in such instances the technique will find application only in the investigations of major acute intakes or in the monitoring of uptake following long-term chronic exposure. In most applications, photon detectors are located at selected sites near or on the body. Usually, at least partial shielding of the detector and/or of the subject will be needed to reduce the interfering response from ambient radiation; in some cases anticoincidence techniques may be required to achieve sufficient background discrimination. Electrical signals from the detectors must be amplified and processed, leading to a gamma-ray spectrum which will most conveniently be stored in computer based systems. Procedures are necessary to separate the response attributable to a given nuclide in the body from that due to ambient radiation and to components associated with other sources of body radioactivity. The extracted response must be converted into an assessment of body or organ radioactivity through appropriate calibration procedures. Three detection features define the requirements for in vivo measurement of radionuclides in the human body: 1. Selectivity: the capability to measure the activity of a radionuclide in the presence of other radionuclides. 2. Sensitivity: the response of the measuring system with respect to the level of radioactivity within the body, i.e. the capability to measure internal exposures below the limits. 3. Accuracy: the mean deviation of the result in terms of radioactivity from the actual radioactivity in the body or in a phantom, respectively. Selectivity The monitoring method has to be prepared to provide nuclide specific information, which is basic requirement for internal dose assessment. Except for circumstances where there is only one radionuclide handled by the workers to be monitored, the measurement system should provide identification of radionuclides. This means that in case of direct monitoring and also when indirect methods are applied for measuring gamma emitting radionuclides, gamma spectrometry is the method, which can meet this requirement. In this respect the semiconductor spectrometry provides better selectivity compared to scintillation spectrometry. Sensitivity The monitoring sensitivity should be high enough to be able to determine with proper safety an internal exposure corresponding to 1 mSv annual effective dose or 10 % of organ dose limits, respectively. To meet this requirement one has to consider the minimum detectable activity for the radionuclides to be expected, the selected monitoring frequency and a reasonable measuring time. This is illustrated by Landolt-Börnstein New Series VIII/4
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Tables 10.12 and 10.13 which show the characteristics for some selected fission and activation products and some selected actinides, respectively. Both the energy and the yield of the predominant photon radiation of most of the fission and activation products are relative high. On the other hand the dose coefficients of these radionuclides are relatively low. So the intake corresponding to 1 mSv is rather high and thus the respective body or organ activities are high even at the end of long monitoring intervals. Contrary, the energy and the yield of the photon radiation of the actinides are typically low and the dose coefficients are high as compared to the fission and activation products. Table 10.12. Radiological and monitoring characteristics for selected fission and activation products [83ICR, 94ICR, 04BMU] Nuclide 22
Na
57
Co
60
Co
125
I
131
I
134
Cs
137
Cs
Predominant photon radiation Typical absorption Energy [keV] Yield [%] type 511 180 F 1275 100 122 86 M 136 11 1173 100 M 1333 100 40 F 27.2 (Kα2) 74 27.5 (Kα1) 14 31.0 (Kβ1) 6.7 35.5 365 82 F
Routine monitoring Intake1) corresponding to 1 mSv Interval [d] Required seneff. dose [Bq] sitivity [Bq] 180 200 5 ⋅105 (Whole body) 180 20000 2.6 ⋅106 (Whole body) 180 1000 1.4 ⋅105 (Whole body) 120 700 1.4 ⋅105 (Thyroid)
9.1 ⋅104
14
569 605 796 662
F
1 ⋅105
180
F
1.5 ⋅105
180
15 98 85 85
100 (Thyroid) 6000 (Whole body) 10000 (Whole body)
1) Inhalation of aerosols with 5 µm AMAD particle size
Table 10.13. Radiological and monitoring characteristics for selected actinides [83ICR, 94ICR, 04BMU] Predominant photon radiation Typical Routine monitoring Nuclide Intake1) corresabsorption ponding to 1 mSv Interval [d] Required senEnergy [keV] Yield [%] type eff. dose [Bq] sitivity [Bq] 235 U 145 11 M 550 180 3 186 57 (Lungs) 239 Pu 1.6 S 120 180 2 13.6 (Lα) 2.3 (Lungs) 17.1(Lβ) 0.6 20.3 (Lγ) 241 13.6 M 37 180 0.2 Am 13.9 (Lα) 18.6 (Lungs) 17.6(Lβ) 0.3 20.3 (Lγ) 35.9 (Skeleton) 59.6 1) Inhalation of aerosols with 5 µm AMAD particle size
The range of the photons in the body governs the sensitivity of in vivo measurement. For soft tissue the range of the 17.1 keV X-rays of 239Pu is 0.89 cm, and so most of the photons emitted from an internal contamination are absorbed within the body. The range of the 1332 keV γ-rays of 60Co, however, is 16.5 cm and so most of the photons emitted in the body will reach the body surface. Thus, the photon flux Landolt-Börnstein New Series VIII/4
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at the body surface due to a 60Co deposition will be at least 3 orders of magnitude higher than that due to a 239Pu deposition with the same activity. Moreover, when taking into account the different dose coefficients, the photon flux due to an internal 60Co exposure will be more than 6 orders of magnitude higher than that due to a 239Pu exposure with the same committed effective dose, this illustrating the range of sensitivity required for in vivo measurements. Accuracy In monitoring of occupationally exposed workers for radiation protection purposes, procedures must be established to ensure that workers have exposures measured and recorded with a reasonable degree of accuracy. General requirements for the overall accuracy of the dose assessment have been recommended by the ICRP [97ICR]. Special requirements for in vivo measurement have been defined for example by the U.S. Department of Energy in its Laboratory Accreditation Program DOELAP [99USD]. These recommendations have been adopted by national guidelines, as for example the German Guideline for Internal Monitoring [04BMU]. The various national and international guidelines have been harmonised in the framework of the European project IDEAS [03DOE]. The accuracy of a monitoring result reflects to the quality of measurements and is usually characterized by two quantities namely by the bias and precision (repeatability). The relative bias (Br) is a measure of how close the assessed activity is to the actual activity in the organ(s) or in the whole body. Since the actual activity in the person is rarely known, this criterion applies to measurements on suitable phantoms that simulate the person. The relative bias statistic (Bri) is defined for the purposes of performance testing of a finite number of measurements in each category of analysis by Br =
1 N
N
∑B
ri
i =1
with Bri =
Ai − Aai Aai
(10.3.2.1)
where N is the number of test measurements in a given category (N ≥ 5) Ai is the value of the i th measurement in the category being tested Aai is the actual quantity in the test mock-ups (phantom) for the ith measurement For service laboratories the performance requirement for the relative bias should be 0.25 ≤ Br ≤ 0.50. This requirement can only be considered if all values of Aai exceed the lower limit of detection by at least factor 5. The relative precision (SB) describes the relative dispersion of the values of Bri from their mean Br and is defined as N
∑ (B
ri
SB =
i =1
− Br ) 2
N −1
(10.3.2.2)
For service laboratories the relative precision should be SB ≤ 0.4 for the conditions mentioned above, i.e. if all Aai exceed the lower limit of detection significantly.
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10.3.2.3 Principles of γ spectrometry Photon interactions For understanding γ spectrometry a knowledge of the basic processes by which a photon interacts with matter is essential. Three fundamental processes govern the interaction: photoelectric effect, Compton scattering and pair production. Due to these interactions the intensity I(x ) of the photon flux is decreasing along the pathway x according to the function I ( x) = I 0 ⋅ e − µ ⋅x
(10.3.2.3)
where µ is the total linear attenuation coefficient. For illustration Fig. 10.33 shows the linear attenuation coefficient of water and NaI, respectively. Water is a typical example for a low-Z material, the absorption behaviour of which is similar to soft tissue, and NaI is a typical example for a high-Z material, which is widely used for γ spectrometry. More detailed information about photon interaction can be found in Section 3.5.2.3 and elsewhere [80Tai, 99Kno]. 10 2
10 2
H2 O
5 -1
]
2
10
Linear attentuation coefficients [cm
Linear attentuation coefficients [cm
-1
]
5
5 2
1
Total
5 2
10-1 5 2
Compton
10-2
Photoelectric
5 2
10-3 10-2
2
5
10 2 5 Energy [MeV ]
1
2
10 5 2
1 5 2
10-1
Total
5
Compton
2
10-2 5
Photoelectric
Pair production
2
Pair production -1
Nal
2
5
10
10-3 -2 10
2
5
10-1 2 5 1 Energy [MeV ]
2
5
10
Fig. 10.33. Linear attenuation coefficient of H2O and NaI.
The photoelectric capture predominates for low photon energies and the photons are absorbed much more strongly in high-Z materials than in low-Z materials. If the incident photon is absorbed by photoelectric effect in a detector, the resulting pulse contributes to the so-called full absorption peak or photo-peak, which provides the key information about the energy and the intensity of the incident photon radiation. If the photon undergoes Compton scattering, the resulting pulse contributes to the so-called Compton continuum of the detector spectrum which consists of two components: if the Compton scattering occurs inside the detector the resulting pulse contributes to the detector-specific Compton continuum which covers according to the energy of the scattered electron the energy range from zero for θ = 0° up to the so-called Compton edge for θ = 180°. If on the other hand the Compton scattering occurs in the environment and the scattered photon is subsequently absorbed in the detector, then the resulting pulse contributes to the environment-specific Compton continuum which covers the energy range from the so-called backscatter-peak for θ = 180° up to the photo-peak for θ = 0°. It is important to know that anticoincidence techniques can reduce only the detector-specific Compton continuum but not the environment-specific Compton continuum. More detailed information about photon interactions is given in Section 3.5.2.3.
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Energy resolution The energy resolution of a detector describes its ability to distinguish between photon energies. When exposed to a radioactive source, the detection system processes a number of monoenergetic photons, which results in a spectrum consisting of the photo peak and the Compton continua. For illustration Fig. 10.34 and Fig. 10.35 show the spectra of a NaI(Tl) scintillation detector (see Section 10.3.2.4.4.1) for low-energy photons (5.9 keV Kα X-rays from 55Fe) and for high-energy photons (662 keV γ-rays from 137 Cs), respectively. The low-energy photons are fully absorbed in the detector because at 5.9 keV the probability for photoelectric absorption is more than two orders of magnitude higher than the probability for Compton scattering (Fig. 10.33). On the other hand, the high energy photons undergo mainly Compton scattering in the detector because at 662 keV the probability for photoelectric absorption is one order of magnitude less than that for Compton scattering. Thus, in the spectrum for low-energy photons there is only the photo peak whereas in the spectrum of high-energy photons there is in addition a broad Compton continuum including the backscatter peak and the Compton edge. Ideally, all fully absorbed monoenergetic photons would be assigned exactly the same pulse height (or channel) in the measured spectrum. However, the photo-peak in the measured spectrum is a distribution of pulse heights with a peak width that reflects the detector resolution. The full width at half maximum (FWHM) of the photopeak is used to characterise the resolution of the detector. The FWHM is the energy width of the distribution at half the maximum of the photo-peak when the background has been subtracted. For lowresolution detectors, such as scintillation detectors, resolution is defined as the FWHM divided by the photo-peak energy, and is usually expressed as a percentage. The 662 keV gamma ray from 137Cs is usually used as the reference for this purpose. For high-resolution semiconductor detectors, the energy resolution is usually specified as the FWHM (keV or eV) for a specified energy. Manufacturers normally also provide the width of the photo-peak at one-tenth and one-fiftieth of the maximum, with reference values for the 1.332 MeV gamma ray from 60 Co or the 122 keV gamma ray from 57Co, depending on the type of detector. 70000
1400 59
Photomultiplier 1200 tube noise
600 FWHM 77 channels
Peak channel = 70 Resolution =6/70 = 8.6%
40000 30000 20000
FWHM 6 channels Backscatter peak
10000
200 0
50000
Counts per channel
800
400
50
100 150 200 250 Channel number
300
350
Cs source
60000
Peak channel = 177 Resolution = 77/177 = 43.5%
1000
Counts per channel
137
Fe source
400
Fig. 10.34. Spectrum of a NaI(Tl) scintillation detector for low-energy photons (5.9 keV Kα X-rays from 55Fe).
0
20
Comption edge 60 40 Channel number
80
100
Fig. 10.35. Spectrum of a NaI(Tl) scintillation detector for high-energy photons (662 keV γ-rays from 137Cs).
Detection efficiency The detection efficiency is defined as the ratio of the number of photons detected to the number of photons emitted by a radiation source during a given time interval. The detection efficiency is made up by four factors:
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1. Geometrical attenuation factor: the fraction of all emitted photons which are emitted in the direction of the sensitive volume of the detector, depending on the solid angle covered by the detector with respect to the source. 2. Material attenuation factor: the fraction of those photons emitted in the direction of the sensitive volume, which actually reach it, depending on the attenuation of the material in between the source and the detector. 3. Interaction efficiency: the fraction of photons reaching the sensitive volume that react with it, depending on the attenuation of the detector material. 4. Data recording efficiency: the fraction of photons interacting with the sensitive volume, which produce recorded events, depending on the type of recording (i.e. photo-peak counting or total counting) and the data acquisition system. 10.3.2.4 Equipment 10.3.2.4.1 Detectors 10.3.2.4.1.1 Scintillation detectors (see also Section 10.1.3) There are three groups of scintillation detectors: crystals, glasses and gases. For in vivo measurements, however, only crystal type scintillation detectors are applied. Each photon, which interacts with the sensitive volume of the crystal, generates a single scintillation pulse. This is a very weak pulse, typically consisting of less than 1000 photons of few eV, so it has to be viewed by a highly sensitive photomultiplier tube (PMT) and the whole assembly must be enclosed in a light-tight housing to isolate the scintillation from the ambient light. The characteristic properties of some selected scintillation materials are summarized in Table 10.14. Table 10.14. Properties of NaI(Tl), CsI(Tl), Bi4Ge3O12, and organic (Polystyrenetetraphenylbuta-diene) scintillators. Property Density [g/cm3] Light output relative to NaI(Tl) Wavelength of maximum emission [nm] Decay constant [µs] Hygroscopic Energy resolution at 662 keV [FWHM in %]
NaI(Tl) 3.67 1 415 0.23 Yes 7 - 10
CsI(Tl) 4.51 0.45 550 1.0 No 10 - 12
Bi4Ge3O12 7.13 0.12 - 0.20 480 0.3 No 10 - 12
Organic 1.0 0.14 450 0.005 No 25 - 50
Sodium iodide activated with thallium (Nal(Tl)) provides the best properties with respect to light output, decay constant and energy resolution. However, NaI(Tl) is hygroscopic, and absorption of water results in a loss of energy resolution. So these crystals must be placed in gas-tight housings. A typical design of a NaI(Tl) scintillation detector is shown in Fig. 10.36. In between the crystal and the housing there is a MgO reflector and so the scintillation light can leave the crystal only via the optical window. Thus most of the scintillation light is collected on the photo cathode of the photomultiplier tube behind the optical window. A magnetic shield protects the photomultiplier tube in order to avoid deflection of the secondary electrons in the tube by external magnetic fields. Both the reflector and the magnetic shield make sure that the amplitude of the output signal is proportional to the energy absorbed in the NaI(Tl) crystal.
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Æ 62
Magnetic shield
187 Æ 216
130 mm Photo tube Optical window
25
Mounting holes - 6 × M6 on 200 B.C. Reflector - packed MgO
H + 13
NaI (Tl) crystal Spun body - 0.8 Al 2S Æ 184.2
Fig. 10.36. Typical design of a NaI(Tl) scintillation detector (Harshaw matched window assembly with crystal dimensions: 178 mm ∅; 130 mm height).
Caesium iodide activated with thallium (CsI(Tl)) can also be used as scintillator material, but the light output is smaller and thus the energy resolution is not as good as that of NaI(Tl). So CsI(Tl) crystals are not commonly used for spectrometric measurements but they are frequently used as anticoincidence detectors in order to reduce the background of NaI(Tl) crystals, this requiring high efficiency but no highenergy resolution. The combination of NaI(Tl) and CsI(Tl) crystals as a dual phosphor sandwich (phoswich) allows for high sensitive detection of low-energy photons (see Section 10.3.2.4.3.2). Because of the poor energy resolution of the scintillation detectors the photo peaks are relatively broad and thus the background prediction can be difficult, especially in the low-energy region. This is illustrated by Fig. 10.37 which shows the spectrum of a NaI(Tl)/CsI(Tl) phoswich detector for a subject with 0.25 kBq 241Am in the liver, 1.5 kBq 241Am in the skeleton and 12 kBq 137Cs in the whole body. The photo peak due to the X-rays of 137mBa (daughter of 137Cs) is overlapping to some extend the photo peak due to the γ-rays of 241Am, and thus the separation of the photo peaks from each other and from the Compton continuum requires high sophisticated spectrum evaluation procedures (see Section 10.3.2.5). 4000 3500
Counts per channel
3000
241 137m
Am:59.6 keV
Ba:31.8 /32.2 keV
2500 2000
Fig. 10.37. Spectrum of a NaI(Tl)/CsI(Tl) phoswich detector (Harshaw 208 mm ∅ matched window assembly with 1 mm thick NaI(Tl) crystal and 51 mm thick CsI(Tl) crystal) for a subject with 0.25 kBq 241Am in the liver, 1.5 kBq 241Am in the skeleton and 12 kBq 137 Cs in the whole body (detector arranged over the liver, measuring time 2000 s).
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Bismuth germanate (Bi4Ge3012), often abbreviated to BGO, provides the highest efficiency of all available scintillation materials because of its high density and effective atomic number. The energy resolution, however, is relative poor and so BGO is applied only in those cases where small detectors with high photo peak efficiency are required, as for example wound measurements. Organic scintillation detectors Solutions of organic liquids [61Lan], and solid organic scintillators [62Bur], have also been used for in vivo measurement applications. Solid organic scintillators can be made by impregnating plastic materials with anthracene. They can be made in very large sizes (e.g. 60 × 40 × 10 cm3) but require several photomultiplier tubes to achieve even a modest energy resolution, and in consequence have not been widely adopted for radiological protection purposes. They could be considered only where the interest was in a single nuclide, or in a mixture whose composition was reliably known, and where interference from the body's natural 40K could be either neglected or inferred from measurements prior to the subject's exposure. Organic scintillators can also be incorporated into liquid solvents; geometries approaching 4π can be produced with such solutions contained in annular tanks, but they suffer from the same restrictions as organic scintillators. Organic scintillation detectors have almost the same photon absorption behaviour as soft tissue and thus they can be applied for direct assessment of the actual internal dose rate due to incorporated γ-emitting radionuclides [95Doe].
10.3.2.4.1.2 Semiconductor detectors Semiconductor detectors are solid-state ionisation chambers, the principle of which being described in detail in Section 10.1.4 and elsewhere [92Del, 99Kno, 80Tai]. For in vivo measurements most commonly Germanium detectors are used, starting in former times with Li drifted Germanium detectors (Ge(Li)), which have been replaced since 1976 by high purity Germanium detectors (HPGe) [76Fal]. In the eighties mainly p-type HPGe crystals have been used. These crystals have a lithium diffusion zone to form the n-contact, which results in an insensitive layer of about 0.6 mm at the crystal surface. So only photons with energy higher than about 50 keV could be detected. Since the nineties also n-type detectors are been applied which have a boron ion implantation to form the p-contact. The implantation results in a very thin insensitive layer (0.3 µm) at the front of the crystal, thus allowing also for measurement of low-energy photons with energies of several keV. Semiconductor detectors have major advantages in energy resolution, the FWHM being typically below 0.6 keV for low energy photons or 2 keV for high energy photons, respectively (Table 10.15). Thus, semiconductor detectors allow almost unambiguous identification of the radionuclides in a mixture, but most of them are inconvenient in that they need cooling to liquid nitrogen temperatures. High purity germanium (HPGe) detectors can tolerate cycles to room temperature but need cooling during operation. Furthermore, many semiconductor detectors are available only in fairly small sizes, so that their geometrical efficiency is small as compared to inorganic crystals and other scintillators. Compact arrays of three to six detectors are becoming standard for monitoring contamination in specific organs, such as the lungs. Miniature semiconductor detectors, in particular those using cadmium telluride (CdTe) operating at room temperatures, are becoming increasingly available. CdTe detectors offer high sensitivity for detection of low energy photons. Their small size (approximately 10 mm in diameter and 2 mm thick) makes them ideal for localized wound monitoring.
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Table 10.15. Properties of some selected semiconductor detector materials. Property Density [g/cm3] Band gap at 300 K [eV] Energy per electron-hole pair at 77 K [eV] Requires cooling Energy resolution at 5.9 keV [FWHM in keV] Energy resolution at 1332 keV [FWHM in keV]
Si 2.33 1.12 3.61 Yes1)
Ge 5.33 0.67 2.98 Yes 0.6 1.8
CdTe 6.06 1.47 4.43 No
1) If high resolution is required
The excellent energy resolution of HPGe detectors is illustrated by Fig. 10.38., which shows the spectrum for a subject with 0.25 kBq 241Am in the liver, 1.5 kBq 241Am in the skeleton and 12 kBq 137Cs in the whole body. A comparison of this spectrum with the corresponding phoswich spectrum measured at the same subject in the same geometry at almost the same time (Fig. 10.37) reveals the major advantage of the HPGe detectors especially for the in vivo measurement of low-energy photon emitters. 200 180 241
Counts per channel
160
Am:59.6 keV
140 120
137m
Ba:31.8 /32.2 keV
100 80 60 40 20 0
200
400 600 Channel number
800
1000
Fig. 10.38. Spectrum of a HPGe detector (Silena 50 % coaxial n-type HPGe crystal) for a subject with 0.25 kBq 241 Am in the liver, 1.5 kBq 241Am in the skeleton and 12 kBq 137Cs in the whole body (detector arranged over the liver, measuring time 2000 s).
10.3.2.4.1.3 Gas-filled detectors There are three types of gas-filled detectors used for radiation protection measurements: Geiger-Müller tubes for counting of absorption events, ionisation chambers for dosimetry, and proportional counters for spectrometry. Large-area proportional counters with anticoincidence guard layers offer energy resolutions intermediate between those of scintillation counters and semiconductor detectors and, with acceptable detection efficiency at energies below 30 keV, they were in former times seen as the most profitable approach to the assessment of plutonium in lungs. Several designs were produced by individual laboratories using argon-methane or xenon-methane counting gas at normal or high pressure. The energy resolution was 13-15 % and the absorption probability was 30-85 % for the 13.6 and 17.2 keV uranium L X-rays from the decay of 239Pu. High sophisticated designs were developed using full space anti-coincidence techniques for background suppression [76Sch]. However, the proportional counters fell into disuse with the development of the phoswich, which offered greater sensitivity and robustness.
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10.3.2.4.2 Electronics (see also Section 10.1.3) Electronic equipment is required to extract, amplify and sort electrical signals from the detector, converting them ultimately into a pulse-height distribution. It therefore includes a chain of preamplifier and main amplifier for pulse shaping and amplification followed by an analogue-to-digital converter and multi-channel analyzer for pulse height analysis; besides, units with specialized functions are required for phoswich detectors, or for controlling the movement of detectors scanning, high voltage power supplies, etc. Preamplifier The preamplifier being located close to the detector serves as a gain stabiliser and an impedance matcher. For scintillation detectors mainly voltage-sensitive preamplifier is used whereas for semiconductor detectors charge-sensitive preamplifiers are preferred. The preamplifier adds no or little amplification to the signal but it enables the signal to pass without loss of information through a coaxial cable to the main amplifier, which in general is located out of the shielding in several meters distance from the detector. Main amplifier The main amplifier serves for amplification and proper shaping of the pulses from the preamplifier. Modern amplifiers provide many functions for smoothing and shaping of the pulses by integration and differentiation, baseline restoration, pole-zero adjustment and linear amplification, the gain varying from a factor 10 to 5000. The output signal can be chosen to be unipolar (either positive or negative) or bipolar (first positive and then negative), the standard for spectroscopic applications being a positive unipolar pulse with amplitude up to 10 V. The length of the pulses is controlled by the shaping time constants and should not exceed 10 µs in order to avoid superposition of pulses at higher count rates which would result in broadening of the photo peaks and thus in a loss of energy resolution of the detector system. Pulse-height analysis For determination of the amplitude the pulses are first digitised by an analogue-to-digital converter (ADC) and then fed into a multichannel analyser (MCA). The MCA reads the pulses according to their height into a memory consisting of a certain number of channels, thus generating a spectrum i.e. a frequency distribution of the pulses as a function of their amplitude. The total number of channels per spectrum is defined by the width of the photo peaks. It should be large enough to provide full information about the peak shape in order to allow for proper separation of the photo peak from the background continuum or from overlapping photo peaks due to other radionuclides, respectively. On the other hand the number of channels should not be too large in order to have a sufficient number of pulses per channel or good counting statistics, respectively. For scintillation detectors 0.25k (256) channels are sufficient to cover the energy range from zero to 3000 keV, whereas for HPGE detectors up to 8k (8096) channels may be required to provide full information in this energy range. At present time, both ADCs and MCAs are available as plug-in components of personal computers and customized computer codes for the qualitative and quantitative analysis of the measured spectra are available for all gamma spectrometry systems. The computer codes include basic functions such as background subtraction, and also more ambitious operations such as adjustment for instrumental drift, resolution of a spectrum into its several components by linear regression analysis and peak search, evaluation and identification procedures (see Section 10.3.2.5). 10.3.2.4.3 Shielding The purpose of shielding is first of all to reduce the background radiation to the level necessary for the sensitivity required; but also to reduce perturbations in the counter background response, which occur because the subject's body distorts the ambient radiation field through absorption, scattering and other processes. The background is governed by cosmic radiation and the radiation of radioactive materials present in the local environment (see Chapter 11). The cosmic radiation consists primarily of charged particles of solar or galactic origin, which produce mesons, electrons, photons and activation products such as 7Be or 14N and
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O due to interaction with the earth atmosphere. The radioactivity in the local environment is mainly due to natural uranium and 40K, but also due to man-made 137Cs from the nuclear weapons fallout and the Chernobyl accident, respectively. Also airborne activity such as 222Rn and 220Rn progeny may contribute significantly to the background radiation in the environment of the detector device. In addition, radionuclides in the detector device itself may give rise for some background components, i.e. 60Co in steel components, uranium in aluminium and beryllium components, and 40K in photomultiplier tubes. The background may be characterized by the background index that gives the count rate per unit detector volume, typically in units of counts per minute per cm3 over the energy range from 200 keV to 2 MeV. The reduction of the background is characterized by the background reduction factor indicating the ratio of count rates measured with a detector in a well defined energy range without and with applying a certain procedure for background reduction, as for example outside and inside a shielding room. In principle, there are passive and active methods for background reduction: the passive methods are based on the absorption of environmental radiation with appropriate shielding materials and/or on the selection of construction materials with very low intrinsic radioactivity in the environment of the detector device whereas the active methods are making use of special anticoincidence techniques for reducing the detector specific background signal. 10.3.2.4.3.1 Passive shielding The requirements for primary shielding materials are: high attenuation of gamma rays, requiring high atomic number and density; freedom from unacceptable concentrations of natural or artificial radionuclides; and suitable mechanical properties for fabrication and assembly. Steel or lead are most commonly used. If lead is chosen, there will be characteristic X-rays, induced by ambient radiation or by the subject's gamma-ray emissions; if these interfere in a critical energy region, they may be removed through an inner lining of a few mm of cadmium or tin, with the ensuing Cd or Sn X-rays eliminated if necessary by a further lining of steel or copper. For major installations typical thicknesses for the primary material are 5-10 cm lead or 10-20 cm steel. Those thicknesses seem to be optimum due to the fact that the background from the environment is about two orders of magnitude higher than the background from the subject due to the natural 40K in the body. So it is making no sense to reduce the environmental background far below the background from the subject. Thus, a background reduction factor of 100 seems to be reasonable with respect to in vivo measurements in the energy range up to about 3000 keV. Such a background reduction factor is achieved for example by 14.6 cm steel (Table 10.16). Behind the shielding the environmental radiation is scattered down to a broad Compton continuum with a maximum around 200 keV. The Compton continuum may be attenuated by a factor of 100 with an inner lining of about 5 mm Pb. In the Pb lining, however, characteristic X-rays are produced with energy between 72 and 88 keV (Kα1 at about 75 keV). These X-rays then may be absorbed by an additional inner lining of about 2 mm Sn, and the X-rays from the Sn (Kα1 at about 25 keV) then may be absorbed by a third inner lining of about 0.25 mm Cu, this being the principle of the so-called “graded Z lining” which provide optimum shielding for in vivo measurements (Table 10.16). Table 10.16. Design parameters for a “graded Z shielding”. Critical radiation
Reference energy [keV] Background photons 3000 Compton scattered photons from Fe 200 75 X-rays from Pb (Kα1) 25 X-rays from Sn (Kα1)
Absorption material Fe Pb Sn Cu
Linear absorption coefficient [cm−1] 0.315 10.7 25 180
1 % thickness [cm] 14.6 0.43 0.18 0.025
Shielded room The most effective, convenient arrangement is a wholly shielded enclosure, to accommodate both the subject and the detector system. For a given thickness of the chosen shielding material, this design offers the greatest reduction in background and it offers also the smallest dependence of background response on body size. Landolt-Börnstein New Series VIII/4
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Occasionally, subjects may react to isolation in a shielded room; such instances are much rarer where access to the counter is via an open shielded labyrinth rather than through massive hinged or sliding doors. In practice the background reduction factors are smaller than expected from the theoretical consideration above, mainly due to scattering of high energy photons (Eγ >3000 keV) down into the energy range of in vivo counting (Table 10.17). Table 10.17. Shielding parameters for some selected shielded rooms for in vivo measurement devices Shielding materials
Reference Background Background reduction factor index [cpm/cm3] 0.43 85Sch 70 cm silica sand, 0.3 cm Pb, 0.04 cm Cd, 0.1 cm Cu, 73 0.55 cm plastic 15 cm Fe, 1 cm Pb, 0.2 cm Fe 15 cm Fe, 0.9 cm Pb
100 40 - 57
0.36 0.34
85Sum 61Kie
Partial shielding Some installations comprise more open structures, which eliminate direct paths for radiation between the detector and the laboratory. Examples are the “shadow shield” design used for assessing whole-body radioactivity in which the subject lies on a bed moving under a fixed detector in a central turret. Other arrangements embodying the same principle can be devised to assess the radioactivity of individual organs or regions. In another simple arrangement the detector and the back of a chair or bed holding the subject are shielded for the investigation of radioactive deposits in larger regions. With all partially shielded counters the background response below 200 keV is likely to be much larger than in a shielded room, because they respond to photons scattered by the subject into the detector. For this reason a shielded room is essential for the sensitive assessment of low-energy photon emitters. More information about shielding is given for example in the IAEA “Directory of whole-body radioactive monitoring” [70IAE]. In addition to the shielding of the environmental radiation it is very important that all materials in the detector systems, the mounting facilities and the shielding are selected for low level of intrinsic radioactivity. Also the natural radioactivity in the construction materials of the surrounding building is important for the background. The natural activity concentration of 40K can vary in between 200 Bq/kg and 800 Bq/kg in bricks and 320 Bq/kg and 800 Bq/kg in cement, respectively [92Zik]. Thus, there is a large potential for background reduction by proper selection of the construction materials. Moreover it is important to minimise the amount of material close to the detector system in order to minimise the background component due to Compton scattering in the direct vicinity of the detectors. Last but not least there is a need for air filtration in order to reduce the background component due to airborne radioactive materials. 10.3.2.4.3.2 Active methods There is different kind of active methods for the reduction of the detector background signal, most of them being based on anticoincidence techniques. Firstly this technique was applied for proportional counters in order to reduce the background due to β particles or Compton electrons from the environment. For this purpose the counting volume of the proportional counter was surrounded by guard counters and all coincident absorption events in both the counting volume and one of the guard counters were discriminated. Thus only the events due to photoelectric absorption in the counting volume were processed, this resulting in very good detection features especially for low energy photons such as the plutonium L X-rays. On the other hand the sensitivity of these detectors for photons with higher energy as for example the 59.6 keV γ-rays of 241Am is very low, even when the proportional counters were operated with heavy counting gases such as Xe under high pressure of 2 or 3 bar. However, for in vivo measurement of plutonium via the L X-rays it is essential to measure simultaneously the 241Am activity present, because the L X-ray yield of 241Am is one order of magnitude higher than that of plutonium. The energy resolution of the proportional counters does not allow for discrimination of the 241Am L X-rays Landolt-Börnstein New Series VIII/4
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from the plutonium L X-rays and so the contribution from 241Am must be determined by measurement of 241 Am via the 59.6 keV γ-rays in the same measuring geometry. It was mainly due to this reason that the proportional counters were replaced in most laboratories by the phoswich detectors, which were commercially available since the late sixties.
Photo multiplier Other source of radiation
CsI (Tl)
(4) NaI (Tl) (1)
(2)
(3)
Be. entrance window Source material
Fig. 10.39. Measuring principle of a phoswich detector.
The phoswich detector has been developed by Laurer for the in vivo measurement of low energy photon emitters such as 210Pb, 239Pu and 241Am [68Lau]. The detector consists of a large area NaI(Tl) crystal the thickness of which being just enough to fully absorb the low energy photons (typically 1-3 mm). The NaI(Tl) crystal is backed by a CsI(Tl) crystal for the detection of scattered photons due to Compton effects in the NaI(Tl) crystal (Fig. 10.39). The photomultiplier tubes detect the scintillation light from both crystals. However, because of the different scintillation decay times of the materials, it is possible to assess by pulse shaping techniques whether the scintillation light is originating from the NaI(Tl) crystal, the CsI(Tl) crystal or from both crystals. Thus it is possible to discriminate the Compton scattering events (in both crystals) from the full absorption events, this resulting in a significant reduction of the detector specific Compton continuum of the NaI(Tl) crystal. When comparing the count rates of a phoswich detector in the low energy range (10 - 100 keV) with and without applying the pulse shape discrimination technique a reduction of about one order of magnitude is observed. This figure, however, does not correspond to the actual reduction of the Compton continuum from the NaI(Tl) crystal because most of the scintillations detected by the photomultipliers are due to absorption events in the CsI(Tl) crystal only. Actually the reduction of the NaI(Tl) Compton continuum is less than a factor 2 because the CsI(Tl) anti Compton shield covers less than a 2π space angle. Phoswich detectors provide good detection features for low energy photon emitters if no high-energy photon emitters such as 137Cs are present. Compton continuum, backscatter peaks and characteristic X-ray peaks due to those high-energetic emitters may influence the spectrum in the low energy region significantly, this giving rise for problems in background prediction.
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10.3.2.5 Spectrum evaluation The interpretation of a photon-energy spectrum of body radioactivity will involve initially the identification of radionuclides responsible for its individual features. The next stage, unless the spectrum is dominated by the contributions from a single radionuclide, will generally involve resolution into the constituent components. In a further process, the response attributable to a particular contributor will be translated into an estimate of body or organ radioactivity; this is accomplished by reference to a spectrum representing a known radioactivity of the nuclide measured in the same conditions. Methods of deconvoluting photon-energy spectra of body radioactivity do not differ in principle from those applied in X- and gamma-ray spectrometry generally, except that account must often be taken of the effect on spectral shape of scatter in a large attenuating mass. The process is at its simplest in the estimation of peak areas from semiconductor detectors. The good energy resolution of such instruments allows the effective background response underlying a spectral peak to be reliably deduced from the adjacent continuum. In the case of scintillation counters, the width of spectral peaks makes this approach often inapplicable especially in case of multiple peaks. It will generally be necessary to first subtract an appropriate spectrum of counter background and a more rigorous analytical procedure will generally be required. The activity of the given radionuclide (q) can be expressed as follows q=
where
N t ⋅ y ⋅η
(10.3.2.3)
N is the number of net counts in the full energy peak area, t is the measuring time, y is the yield of γ (or X)- ray, η is the counting efficiency at the given energy for the respective measuring geometry
This is the most dominating evaluation method in gamma spectrometry. Stripping method When deconvolution of spectra from scintillation counters is required, a “stripping” process is sometimes followed. Reference spectra are derived for each nuclide present, each representing the response from known amounts of the nuclide in appropriate measuring geometry and in relevant absorbing media. The reference spectrum containing the peak with the highest energy is selected, and it is normalised to the subject's spectrum on the basis of count rate in an energy region where only that nuclide contributes. Subtraction of the normalised spectrum gives a residue representing the remaining components, which is treated in the same way. The activity of each radionuclide is calculated directly from the fraction of its reference spectrum, which must be subtracted. In principle, the process can be repeated until the residue consists of the response from a single nuclide only. In practice, unacceptable errors are likely to accumulate if the number of stages exceeds two or three, particularly in relation to minor components in a spectrum; moreover, the method will generally be inapplicable when the dominant peaks of different components overlap. Linear regression analysis A more satisfactory procedure in many situations is to adopt a method of linear regression analysis, to derive the proportions of each reference spectrum which, when combined, gives rise to the best fit to the subject's spectrum. Facilities for such analyses are embodied in several commercially available computer programmes for processing γ ray spectra; alternatively, they can be developed locally. Utilizing a much larger portion of the spectrum, instead of the restricted regions successively considered in the stripping process, this method gives improved statistical accuracy in the estimates of the various components; moreover, realistic estimates of this accuracy may be derived in the matrix-inversion procedures. As with other methods of deconvolution, this approach has its limitations. In particular, it demands stability of the spectrometer during the measurement, especially if the nuclides present possess overlapping spectral features. It is also important that the locations of peaks in the subject's spectrum should coincide with those in the relevant reference spectra; where minor drifts occur between the measurements, adjustments can
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often be made prior to the analysis if appropriate routines are available. However, the validity of the analysis depends also on the spectral shapes of the reference standards according with those of the corresponding components in the subject's spectrum. This method is used practically exclusively in scintillation gamma spectrometry, however the procedure can be extended to the deconvolution of any kind of complex distribution of measured data. Such application is unfolding profile scanning data to quantify the measured distribution pattern in terms of activity deposited in different body regions or organs. 10.3.2.6 Measuring geometries In principle, the in vivo measuring systems can be allocated to two different types of systems, namely geometry dependent and geometry independent systems. Here the geometry is defined as the detector configuration in relation to the photon-emitting source in the body. 10.3.2.6.1 Geometry depending systems Static geometry The most common kind of geometry dependent systems are those having detectors that are positioned close to the subject looking to specific organs or tissues. The advantages include high efficiency, better subject positioning and less space requirement for the system. The use of such a static geometry dependent system is extremely important in measuring low energy photons where the efficiency of the detection system needs to be maximized. Isotopes of iodine, and also 99mTc may concentrate in the thyroid gland. The range of photon energies encountered is 27 keV (125I) to several hundred keV. Some HPGe detectors used for assessment of actinides in lungs are of suitable diameter (ca. 50 mm) in relation to the size of the thyroid, and are large enough to provide adequate detection efficiency over most or all of this energy range. Alternatively, a planar germanium detector or thin NaI(Tl) crystal may be adopted for photon energies <100 keV, with thicker crystals, either NaI(Tl) or co-axial germanium detectors, used if necessary to secure efficient photon detection at high energies. More accurate assessment of an easily-detectable deposit would require better shielding, with the detector recessed in a suitable collimator. Inhalation is the most common route for intake in occupationally exposed personnel, with the respiratory tract being the site of initial deposition. If the deposit persists for a sufficient time, monitoring of pulmonary activity may offer the most sensitive and reliable means of assessing the intake. Indeed, in the case of certain actinides which subsequently re-locate to organs absorbing virtually all low-energy photon emissions, it offers the only remotely practicable means of assessing an intake by external counting. A large-diameter (150-300 mm) stationary NaI(Tl) detector recessed in a cylindrical collimator may be used. The response would preferably be recorded in two locations, the detector viewing in turn the anterior and posterior surfaces of the thorax; ideally, if two such detectors were available, the measurements could proceed simultaneously. The sensitive measurement of pulmonary deposits of low-energy photon emitters (<100 keV) is most commonly required for 241Am and for isotopes of uranium and plutonium. It requires equipment giving a better signal-tobackground ratio than is provided by NaI(Tl) detectors. This is achieved either through partial suppression of background response as in the phoswich detector, or through the improved energy resolution offered by semiconductor detectors [00Lop]. Fig. 10.40 - Fig. 10.42 show as typical examples the arrangements of 2 phoswich detectors and 4 HPGe detectors as used in the Research Centre Karlsruhe for the in vivo measurement of low energy photon emitters in the lungs.
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Fig. 10.40. Typical arrangement of 2 phoswich detectors (Harshaw 20 cm diam. 1 mm NaI(Tl) / 50 mm CsI(Tl) crystals) for in vivo measurement of low energy photon emitters in the lungs
Two other sites of deposition, which frequently attract specific interest, are bone and liver. Arrays of detectors viewing the skull have been employed to assess skeletal deposits of 210Pb (47 keV photons) and 241 Am (59.6 keV). The levels of contamination in liver and skeleton may also be of interest in regard to their interfering contributions when lung deposits are assessed. An extreme, but fairly common example of a localized deposit is the presence of poorly soluble radioactive material at the site of a puncture wound, investigated shortly after an accident, before important quantities have become systemic. With fission or activation products giving abundant and energetic photon emissions and with high limits of intake, improvised arrangements employing any spectrometrically suitable scintillation or semiconductor detector are likely to be satisfactory. Profile scanning geometry Another kind of geometry dependent systems is using moving detectors or subjects. The so-called scanning systems can be used for identifying the organ or tissue where the radionuclide in question is deposited. The so-called profile scanning systems are providing information on the activity distribution pattern in the body. Better spatial resolution can be obtained by using collimators in the front of the detectors however it can only be done at the expense of the counting efficiency. Profile scanning measurement may also point to the presence of surface contamination, which has to be removed. It should be noted that profile scanning with simple slit-collimated detectors would often give reliable indications of the relevant sites of deposition, but that only a rough evaluation of that deposition can be obtained if a single detector viewing only one aspect of the body is employed. Quantitative assessment of radionuclide distribution by profile scanning requires either paired detectors or, if only one is available, separate traverses of the anterior and posterior surfaces, and further improvements may result if focused collimators are used and computer-aided evaluation techniques are applied.
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Fig. 10.41. Typical arrangement of 4 HPGe semiconductor detectors (Silena HPGe sandwich with 20 cm diam. planar p-type HPGe crystal backed by 50 % coaxial n-type HPGe crystal) for in vivo measurement of low energy photon emitters in the lungs.
10.3.2.6.2 Geometry independent systems Geometry independent systems are characterized by the detectors being located in a certain distance from the subject to be measured, either above or below or both. In those arrangements the detectors have an almost uniform response for any distribution of gamma emitters in the body. Thus, arrangements characterized by this geometry independent feature were originally called as whole-body counter. Arc geometry This arrangement is capable of high accuracy if the levels of internal radioactivity and the space available inside a shielded room are sufficient. The subject lies on a curved frame forming the arc of a circle centered at the detector, so that all parts of the body are roughly equidistant from it. The detection efficiency is of course poor and, even with a large detector in a heavily shielded room, the technique will not generally be applicable to the determination of body burdens below several kBq; it would seldom be feasible with semiconductor detectors. In the context of radiological protection, therefore, the arc method will be used primarily in the investigation of established cases of internal contamination rather than as a regular means of control.
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Chair geometry In this arrangement the subject reclines in a tilted chair with the detector supported typically 0.4 m above the abdomen. The detector will commonly be a NaI(Tl) scintillator, e.g. 200 mm diameter × 100 mm, and in a heavily shielded room would be capable of detecting as little as 50 Bq of most common fission and activation products in a counting time of 15 min. Similar arrangements have been employed with semiconductor detectors. The response will however depend markedly on the location of the radioactive deposit, e.g. in the lungs or liver, may differ by a factor of two or more from that applying to material widely dispersed in the body. The potential for systematic error is accordingly much greater than with the arc technique, but this will be unimportant in many routine applications. Recently, fully automated chair type whole body counters have been developed providing very good detection features [02Sin]. The chair geometry is also applied in most of the mobile whole body counters used for monitoring of staff of nuclear power plants. Scanning geometry With this design of counter, the response is accumulated from a single detector while traversing the subject's length at a fixed distance above or below the supine body, or in some versions, in a corresponding disposition relative to the erect body. Alternatively, the supine subject may be moved in relation to a fixed detector. Accuracy will be improved if a second traverse is performed with the subject's posture reversed and the evaluation is based on the combined response in the two positions. When employing two detectors (one above and one below the supine subject), a more representative sample of the photon flux is obtained by moving the supine subject relative to the counters during accumulation of the spectrum. However, while scanning arrangements can provide more uniform detection geometry compared with that given by most other configurations, they might not yield significantly greater accuracy than a well-designed multi detector array of stationary detectors. Useful indications of the distribution of a radionuclide within the body may be secured if the system can display a profile of the response according to position. Stretcher geometry An alternative approach, preserving the good detection efficiency given by the chair technique but offering improved uniformity of response, is to adopt a “stretcher” geometry, with the subject in a supine posture and several detectors distributed about the body. Commonly 4-8 detectors are employed, disposed above and below the stretcher so that their combined response is acceptably independent of the source distribution. Complete uniformity of efficiency is of course unattainable, however, for activity, which is not concentrated in small organs or regions, such an array can yield results for energetic (>100 keV) photon emitters accurate to within 20 % or better. Fig. 10.42 shows as an example a stretcher type whole body counter with 4 NaI(Tl) scintillation detectors as used in the Research Centre Karlsruhe for in vivo measurement of photon emitters with energy between 100 and 3000 keV. Disadvantages are the need to provide several independently adjustable supporting mechanisms for the various detectors, and the requirement of a larger shielded room than would generally be necessary to accommodate only a chair. Table 10.18. Comparison of various whole-body counter geometries Geometry
Mechanical arrangement
Uniformity of response
Information on Sensitivity distribution
MDA of 137Cs. [Bq]
Arc
fixed
very good
no
low
300
Chair
fixed
poor
no
high
100
Static array
fixed
good
possible
high
70
Scanning (collimated detectors)
moving detector or moving bed
good
yes
high
100
Shadow shield
moving bed
good
yes
medium
130
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Fig. 10.42. Stretcher type whole body counter with 4 NaI(Tl) scintillation detectors (Bicron 20 cm diam. × 10 mm crystals) for in vivo measurement of medium energy photon emitters (100 - 3000 keV)
10.3.2.7 Calibration The purpose of whole-body counter calibration is to determine the relationship between the detector response and the radioactivity in the body. Gamma emitting radionuclides can be characterised by the shape of their spectra or by the location of spectral peaks, which correspond to emitted photon energies. In certain circumstances, for example when spectra from scintillation counters are analyzed by leastsquares fitting, it is essential to use the same nuclides for calibration as those to be measured. In the situations it may suffice to derive a calibration factor by interpolation of data measured individually for a series of monoenergetic photon emitters covering the energy regions of interest. With high-resolution semiconductor detectors, a single nuclide emitting photons at several energies, or a suitable mixture of nuclides, may be more convenient. 10.3.2.7.1 Energy calibration As has already been mentioned energy calibration establishes the relationship between spectral peak location (channel number) and emitted gamma ray energy. This relationship may then be used to identify radionuclides from the location of their spectral peaks. Energy calibration is performed using radioactive sources emitting gamma rays with known energies. The gamma energies emitted by the calibration source(s) should cover the range of energies likely to be encountered during whole-body measurements.
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Table 10.19 lists a selection of radionuclides suitable for calibration purposes. The analysis systems in modern whole-body counters automatically derive the energy calibration relationship using fitting routines and also plot the function for inspection. Many software techniques require that the peak shape has to be characterised as a function of energy. Even if the software does not require it, peak shape is a useful tool both to confirm proper operation of the detector and to identify the presence of multiple nuclides with similar gamma energies. Peak shape calibration, characterised by FWHM (Full width at half maximum height), is usually performed at the same time and with the same spectrum as energy calibration. Table 10.19. Characteristic parameters of radionuclides suitable for photon energy and efficiency calibration Nuclide Half-life
Photon emission Energy [keV]
55
Fe
241
109
57
Am
Cd
Co
141
Ce
Nuclide Half-life
5.89 (Mn Kα2) 5.90 (Mn Kα1) 6.49 (Mn Kβ1)
Emission probability [%] 8.4 16.6 3.4
13.93 (Np Lα) 17.61 (Np Lβ) 21.00 (Np Lγ) 26.35 59.54
13.2 19.4 4.9 2.4 36.0
463 d
88.03
3.65
271.84 d
122.06 136.47
85.59 10.58
145.44
48.9
2.73 y
432.0 y
32.50 d
Energy [keV]
137
Ce
137.65 d
165.85
80.0
203
Hg
46.61 d
279.20
81.3
51
Cr
27.71 d
320.08
9.58
22
Na
950.4 d
511.00
180.7
1274.54
Emission probability [%] 99.94
Cs
30.0 y
661.66
85.0
54
Mn
312.5 d
834.84
99.98
46
Sc
83.80 d
889.28 1120.55
99.98 99.99
60
Co
1925.5 d
1173.24 1332.50
99.90 99.98
88
Y
106.66 d
898.04 1836.06
94.6 99.24
4939 d
121.78 244.69 344.37 778.89 963.38 1085.78 1112.02 1407.95
28.37 7.51 26.58 12.96 14.62 10.16 13.56 20.58
152 139
Photon emission
E
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10.3.2.7.2 Efficiency calibration Efficiency calibration allows one to convert the measured count rate in a region of the spectrum to the emission rate of the corresponding gamma photon of the radionuclide. Emission rate can then be converted to disintegration rate or activity using the probability of photon emission per decay. The conversion factor between count rate and activity is energy dependent and must be measured for the range of photon energies, which are expected to be present in the whole-body measurements. The variation in the efficiency factor as a function of photon energy may be obtained by simple interpolation between the measured points on the spectrum or by fitting an empirical function to the points. If one knows in advance the radionuclide to be measured, a situation frequently encountered in medical applications, then the efficiency factor is simply determined by using the same nuclide for calibration. In general, however, more complex spectral techniques are needed to derive efficiency factors for the spectra encountered in radiation protection applications. Different spectral analysis techniques should be applied to high or lowresolution detector systems. To achieve accuracy in calibration the following factors must be observed: • The same detector and hardware configuration should be used both for calibration and whole-body counting; the possible influence of necessary changes on the detection efficiency must be checked. • Identical spectral analysis methods must be used. • The source - detector geometry must be adequately simulated. • The photon attenuation conditions applying in the body should be adequately reproduced As a general rule the lower the energy of the photons to be detected the greater the care required to achieve a specified accuracy. 10.3.2.7.2.1 Point source calibration This procedure may be adopted for whole-body counters whose “geometrical” counting efficiency shows little dependence on the location of the source, for example those disposed in distant arc geometry and in certain scanning arrangements. With the arc geometry, calibration accurate to within a few percent can be achieved with a standardised point source suspended in a tank of water or located between stacked plates of a solid absorber with comparable attenuation properties, the appropriate location of the source in the tank or stack is deduced from examination of the spectral shape and evaluation of the relative photon fluxes emerging from the subject's surfaces. 10.3.2.7.2.2 Phantom calibration In practice systemic radionuclides tend not to be concentrated in a single anatomical region and such counters are generally calibrated, at least in the first instance, with the aid of whole-body phantoms simulating the human form and containing a standardised aqueous solution of the relevant nuclide. The most convenient general-purpose phantom is the so-called BOMAB (BOdy Manikin ABsorption) phantom, which consists of a collection of polyethylene vessels of circular or elliptical cross-section (Fig. 10.43). These are available commercially, or can be made by a workshop with experience of plastics. It is useful to have available two or three such phantoms with dimensions suitably scaled to represent individuals of different sizes; for intermediate physiques calibration factors may be derived by interpolation according to functions of anatomical parameters e.g. weight/height.
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Fig. 10.43. The BOMAB phantom for simulation of homogeneous activity depositions in the whole body 1.5 1.4
Counting efficiency [%]
1.3 1.2 1.1 1.0 0.9 0.8
Fig. 10.44. Counting efficiency of a standard whole body counter (stretcher type with 4 NaI(Tl) scintillation detectors as shown in Fig. 10.43) for homogeneous activity depositions in the whole body according to calibration with the BOMAB phantom.
0.7 0.6 0.5 0
500
1000 1500 Photon energy [keV ]
2000
Alternatively makeshift arrangements, e.g. using plastic reagent bottles, may suffice; and in some situations more versatile in simulating specific physiques and postures (Bottle phantom). Conversely, much more elaborate whole-body phantoms (REMCAL, REMAB) can be purchased, some of them provided with discrete organs, which can be labelled independently of a dispersed deposit. Such complicated devices would ordinarily be required only in the calibration of equipment for assessment of deposits in specific organs. Recently, phantoms were developed which do need to be filled by water solution of radionuclides. Organic gels with dissolved radionuclides are used for the filling of BOMAB phantoms or large number of point sources is inserted into polyethylene bricks from which phantoms of different body height and weight could be easily built (IGOR). Comparison of various phantom characteristics is shown in Table 10.20
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Table 10.20. Comparison of the characteristics of different phantoms used in whole-body counting. Type of Phantom
Availability
Price
Versatility Antropomorphity
Handling
Decontamination
Bottles
very good
cheap
good
satisfactory easy
not necessary
BOMAB
good
medium
bad
satisfactory easy
difficult
REMCAL
good
very expensive
good
good
complicated
very difficult
REMAB
good
very expensive
good
good
complicated
very difficult
Presswood
very good
cheap
bad
bad
easy
not necessary
Besides phantoms for whole-body counting there are several phantom constructions simulating certain body regions such as thyroid in neck phantom (IAEA/ANSI neck phantom) or chest phantom containing organs like lungs, pulmonary lymph nodes, liver etc. (LLNL realistic chest phantom, JAERI phantom for Asian men). These latter phantoms were very sophisticatedly constructed which are especially suitable for calibration of low energy photon emitting radionuclides like transuranium elements deposited in the human respiratory tract. The LLNL chest phantom has been developed at the Lawrence Livermore National Laboratory and then manufactured by Humanoid Systems Inc. (Fig. 10.45). The phantom made from tissue equivalent materials represents the thorax of a male adult with 177 cm height and 76 kg weight. Lungs, tracheo-bronchial lymph nodes and the liver can be exchanged by active components with homogeneous deposition of plutonium and other low energy photon emitters. Overlay structures are also provided for the simulation of chest walls with different thickness (1.6 - 4.1 cm) and different muscle/adipose composition. At present there are 3 generations of the LLNL phantom, which are used as international standard for lung counter calibration [02Kra]. However, the LLNL phantom cannot be used for calibration for inhomogeneous deposition in the lungs or other organs. For this reason a second anthropomorpheous phantom has been provided by Humanoid Systems with the organs having a hole matrix for inserting small cylindrical sources in order to simulate any kind of inhomogeneous organ deposition (Fig. 10.46).
Fig. 10.45. The LLNL chest phantom for simulation of homogeneous activity depositions in the lungs, tracheobronchial lymph nodes and in the liver (Humanoid Systems Inc.)
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Fig. 10.46. The “Fission product phantom” for simulation of activity depositions in all relevant organs of the trunk (Humanoid Systems Inc.)
10.3.2.7.2.3 In vivo calibration On occasion, known activities of short-lived tracers have been administered to volunteers expressly in order to calibrate whole-body counters, e.g. 132Cs and 42K. It may sometimes be possible to make use of subjects who have received intakes in medical diagnosis or other metabolic studies. This would only be possible under supervision of ethics committee. If the measurement is delayed after intake of the tracer, e.g. to allow its distribution to stabilise, excreta voided in the interim may need to be collected and assessed for their content of radioactivity. When taking into account the loss of activity, the results of in vivo calibration are in very good agreement with the results of phantom calibration, as demonstrated by Kaul for 42K [64 Kau]. 10.3.2.7.2.4 Mathematical calibration Methods of calibration using phantoms are relative methods; absolute methods do not require a radioactive standard for calibration, however standards must be used to confirm a calibration. Mathematical phantoms, using Monte Carlo method may be used for such calibration, as demonstrated by Mallet [95Mal]. The advantage of such phantoms is that any distribution of the radionuclide in the phantom could be used. However, thorough comparison of calculated examples with the measured values has to be performed as to ensure good quality of the mathematical phantom. It is especially important when low energy gamma emitters are subject of interest and the radionuclide in the body is nonhomogeneously distributed, as shown by Hunt for some cases with 241Am in the lungs and at the bone surface or with natural uranium in the lungs, respectively [03Hun].
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Fig. 10.47. The “Yale phantom” as used by Hunt for the Monte Carlo simulation of the radiation transport from the activity deposition in the body (i.e. 241Am at the bone surface) to the detector outside the body (i.e. 8” × 4” NaI(Tl) scintillation detector over the lungs) [03Hun].
10.3.2.8 Uncertainties and detection limits Generally, the uncertainties in the measurement are difficult to estimate. When activity levels are low and close to the limit of detection, uncertainties due to counting statistics may dominate the overall uncertainty. For radionuclides that are easily detected and present in sufficient quantity, uncertainties due to counting statistics will be small compared to other sources of uncertainty. Consideration must also be given to systematic uncertainties in other parts of the measurement procedure, e.g. calibration, or correction for body size of in vivo measurements, etc. Typically, the components of uncertainty are grouped in two categories: Type A comprises those components, which can be described by the Poisson distribution (i.e. counting error, to some extend also the variation of background signal and the variation of the subject positioning) whereas Type B comprises all other components (i.e. variation of body dimensions, overlaying structures, distribution of activity within the body, and the uncertainty of the calibration standard). The Type B components cannot be expressed in terms of Poisson statistics, and thus they cannot be associated with the Type A components in order to derive the total uncertainty of the measurement. Table 10.21 lists some typical values for the various components of uncertainty. Table 10.21. Typical values for the components of uncertainty for the in vivo measurements of radionuclides emitting low, intermediate and high photon energy radiation Uncertainty (%) Source of uncertainty (Type) Low photon energy Intermediate photon High photon energy E <20 keV energy E >100 keV 20 keV < E <100 keV Counting statistics (A) 50 % 30 % 7% Variation of detector positioning (A) 20 % 5% <5 % Variation of subject background (A) 50 % 10 % <5 % Variation in body dimensions (B) 50 % 12 % 7% Variation of overlaying structures (B) 30 % 15 % 12 % Variation of activity distribution (B) 30 % 5% <5 % Calibration (B) 5% 5% 5% Spectrum evaluation1) (B) 15 % 5% 3% 1) HPGe detector spectra
The statistical quantities for describing the Type A uncertainty and the corresponding detection limits are analogous to those used in all other kind of radioactivity measurement. Therefore, in recent years mainly the ISO definitions [98ISO] are applied for the calculation of uncertainties and detection limits of in vivo measurements. There are two basic terms being complementary to each other, i.e. decision threshold and detection limit.
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Decision threshold The decision threshold (frequently also referred to as decision level or critical level) is an a posteriori calculated value at which the decision can be made, whether the registered pulses include contributions from the body or are solely due to background. If this decision rule is observed, a wrong decision that there is a contribution from the body when actually only a background effect exists (Type I error), occurs with a well-defined probability α. By definition the decision threshold is given in terms of pulses but for practical application it is frequently transferred to the corresponding activity value. Detection limit The detection limit (frequently also referred to as minimum detectable activity or lower limit of detection) is an a priori calculated value, which specifies the minimum body contribution that can be detected by a defined measurement procedure. The detection limit is complementary to the decision threshold, i.e. when considering the detection limit the wrong decision that there exists only a background effect when there is in fact a contribution from the body (Type II error), occurs with a well-defined probability β. Thus, the detection limit is closely related to the decision threshold defined by the Type I error probability α. By definition the detection limit is given in terms of body or organ activity and it can be compared directly with guideline values. The choice of the values of α and β depends on the aim of the measurement. For the purpose of radiation protection typically the values α = β = 0.05 are used (i.e. 5% probability for both Type I and Type II errors). With these values the following formula for the decision threshold may be derived from the general concept given by Altshuler and Pasternack [63Alt]: N DT = 1.645 ⋅ N B ⋅
where NDT NB tB tS
tS tB
tS 1 + t B
(10.3.2.4)
is the decision threshold in terms of net counts in the full energy peak region for α = β = 0.05, is the total number of background counts in the full energy peak region, is the background measuring time, is the subject measuring time,
For the detection limit the generic formula derived by Currie [68Cur] may be used: N DL = 2.71 + 2 ⋅ N DT
(10.3.2.5)
where NDL is the detection limit in terms of net counts in the full energy peak region for α = β = 0.05. When the background count rate is high enough, Eq. (10.3.2.5) can be simplified and the following expression is derived for the detection limit: q DL =
t 3.3 ⋅ NB ⋅ S tB tS ⋅ y ⋅η
tS 1 + tB
(10.3.2.6)
where qDL is the detection limit in terms of body or organ activity for α = β = 0.05 y is the yield of γ (or X)- ray, η is the counting efficiency at the given energy for the respective measuring geometry
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Table 10.22. Typical detection characteristics of a standard whole body counter (stretcher type with 4 NaI(Tl) scintillation detectors as shown in Fig. 10.40) for homogeneous depositions of some selected fission and activation products in the whole body (subject counting time tS = 300 s; background counting time tB = 1800 s) Nuclide 57
Co Cs 134 Cs 60 Co 22 Na 137
Detected photon radiation Energy [keV] Yield y [%] 123 86 662 85 796 85 1173 100 1275 100
Counting efficiency η [%] 1.5 0.9 0.81 0.66 0.61
Background count rate NB/tB [cps] 14.0 7.4 4.9 3.9 4.0
Lower limit of detection [Bq] Required Achieved (Table 10.12) 20000 60 10000 73 6000 66 1000 62 200 67
Table 10.23. Typical detection characteristics of a lung counter (4 HPGe semiconductor detectors as shown in Fig. 10.3.2.7) for homogeneous depositions of some selected actinides in the lungs (HPGe AntiCompton counter of the Nuclear Research Centre Karlsruhe; subject counting time tS = 3000 s; background counting time tB = 30000 s) Detected photon radiation Counting Lower limit of detection [Bq] Nuclide Background count rate Energy [keV] Emission pro- efficiency Required Achieved η [%] NB/tB [cps] bability [%] (Table 10.12) 239 Pu 2.3 0.017 0.0088 2 1500 17.1(Lβ) 241 Am 59.6 35.9 0.47 0.0090 0.2 3.6 235 U 186 57 0.24 0.0041 3 3.0 1) Calibration with LLNL chest phantom for 25 mm chest wall thickness and 50/50 muscle/adipose tissue composition
10.3.2.9 Measurement procedure Subjects for direct measurements should be free of external surface contamination and in fresh clothing, often disposable paper garments. Accessories such as jewellery, watches and spectacles should be removed. Such precautions help to avoid false identifications of internal activity, and also prevent the transfer of contamination to the counting equipment. Individuals should, to the extent practicable, be in a defined counting position, to ensure reproducibility in serial measurements and to improve comparison with calibration results. In some cases the subject will need to remain stationary for periods up to an hour for satisfactory precision in the measurement. Some means of communication should be provided for subjects in enclosed shielding, particularly when extended counting periods are necessary. For counting systems based on scintillation counting (NaI(Tl) crystals or phoswich detectors), background counts for the detector system should therefore be determined using an appropriate phantom, as similar as possible to the subject to be counted and placed in the defined counting position. For whole body counting, background counts determined using uncontaminated subjects matched with respect to gender, height and weight would improve results. Measurements of background in the counter should be made as close as possible in time to the measurement of the subject, ideally just before and just after. When using semiconductor detectors, background counting with matching phantoms is not necessary. 10.3.2.10 Quality assurance and control Quality assurance (QA) for the measurement of internally deposited radionuclides includes all steps necessary to confirm the accuracy of the measurement and the validity of the dosimetrc interpretation. Guidance for quality assurance is provided by international institutions such as the International Standards Organization as well as by national authorities. The nature and extent of the QA programme should be consistent with the number of workers monitored, and the magnitude and likelihood of exposures expected in the workplaces to be covered by the monitoring programme. Landolt-Börnstein New Series VIII/4
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All persons involved in the internal exposure assessment programme are responsible for its quality and therefore for implementing its QA programme and quality control (QC) procedures. Responsibility for the quality of a particular operation should be delegated to the person actually performing the operation. Such persons should be actively involved in the development of QC procedures, and trained in methods of detecting non-compliance. A direct measurement facility should have a designated QA representative. This representative should monitor QC procedures, perform internal audits of the programme, and be responsible for training all personnel in QA, both in general terms and in the specific quality aspects of their individual work. The fundamental requirements for a complete programme of QA include: • • • •
compliance with general operational requirements stated in accepted written criteria a clearly documented in-house QA program periodic performance evaluations, including proficiency measurement tests documented procedures and quality assurance programme for services provided to customers
Quality control Quality control in a measurement process is important to ensure that assessment of intakes is as reliable as possible. Evidence of the validity of such assessments may be required for legal and/or regulatory purposes. Quality control programmes include the following activities: • • • • • •
procedures and protocols for proper management of the internal dosimetry programme, detector system verification, routine verification of proper instrument performance, data recording and archiving, audits and accreditation, intercomparison [01Ram, 00Doe].
Procedures and protocols The laboratory should prepare and maintain an operational manual that outlines responsibilities and provides requirements for data control, document control, maintenance/test, equipment calibration and checks, procedure, training, corrective action in the event of non-compliance, and traceability to standardizing bodies. The operations manual should include procedures to verify that the quality of the measurements meets the appropriate accuracy requirements. The quality control procedures should be carried out at appropriate intervals. Performance checks Performance checks of the system include energy calibration, energy resolution measurement and determination of the relative counting efficiency, generally using a point source. National regulations may require that facilities concerned with measurement and internal dose assessment be accredited. Such accreditation programmes will have specifications for QA and QC to be implemented.
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10.3.2.11 References for 10.3.2 27Blu 29Sch 31Sch 37Eva 41Raj 51Sie 57And 57Sie 58Bir 61Kie 61Lan 62Bur 63Alt 64Kau 64Raj 68Cur 68Lau 70IAE 72Par 75And 75ICR 76Fal 76Sch 78New 79Gri 80Tai 83ICR 83Too 85And 85New 85Sch 85Sum 87ISO
Blumgart, H.I., Weiss, S.: J. Clin. Invest. 4 (1927) 389. Schlundt, H., Barker, H.H., Flinn, F.B.: Am. J. Roentgenol. 21 (1929) 345. Schlundt, H., Nerancy, J.T., Morris, J.P.: Am. J. Roentgenol. 26 (1931) 112. Evans, R.: Am. J. Roentgenol. Radium Ther. 37 (1937) 368. Rajewsky, B.: Strahlentherapie 69 (1941) 438. Sievert, R.M.: Ark. Fys. 3 (1951) 337. Anderson, E.C.: Br. J. Radiol., Suppl. 7 (1957) 27. Sievert, R.M., Hultqvist, B.: Br. J. Radiol., Suppl. 7 (1957) 1. Bird, P.M., Burch, P.R.J.: Phys. Med. Biol. 2 (1958) 217. Kiefer, H., Maushart, R.: Kerntechnik 3 (1961) 228. Langham, W.H., in: Radioactivity in Man. Thomas, G.C. (ed.), Springfield, IL: Charles C. Thomas, 1961, p. 311. Burch, P.R.J., Hughes, D., Hnuma, T.A., Overton, T.R., Appleby, D.B., in: Proc. IAEA Symp. On Whole Body Counting, Vienna, Austria, 12-16 June 1961, Vienna: IAEA, 1962, p. 59. Altshuler, B., Pasternack, B.: Health Phys. 9 (1963) 293. Kaul, A., Schoeppe, W., Koch, K.M., Hierholzer, K.: Biophysik 2 (1964) 87. Rajewsky, B., Kaul, A. and Heyder, J., In: Assessment of Radioactivity in Man 1, Vienna: IAEA STI/PUB/84 (1964) 15. Currie, L.A.: Anal. Chem. 40 (1968) 586. Laurer, G.R.: The in vivo measurement of lung burdens of radionuclides emitting soft, penetrating radiations, available from: UMI, 300 N. Zeeb Rd. Ann Arbor, MI, USA, Order 6904570, 1968. International Atomic Energy Agency. Directory of whole-body radioactive monitoring. STI/PUB/213, Vienna: IAEA, 1970, ISBN 02-0-112070-2. Parr, R. M., Dudley, R. A., Fedorov, G. A., In: Assessment of Radioactive Contamination in Man, Vienna: IAEA STI/PUB/290 (1972) 215. Andrasi, A., Kotel, G.: Int. J. Appl. Radiat. Isot. 26 (1975) 451. International Commission on Radiological Protection, ICRP Publication 23, Oxford and New York: Pergamon Press, 1975. Falk, R.B., Tyree, W.H., Wood, C.B., Lagerquist, C.R.A., in: Advances in Radiation Protection Monitoring. Proceedings of a Symposium, Stockholm, Sweden, 26-30 June 1978, Vienna; IAEA, 1979, p. 445. Schmitt, A., Fessler, H., in: Diagnosis and Treatment of Incorporated Radionuclides. Proceedings of a Symposium. STI/PUB/411, Vienna: IAEA, 1976, p. 285. Newton, D., Fry, F.A., Taylor, B.T., Eagle, M.C., Sharma, R.C.: Health Phys. 35 (1978) 751. Griffith, R.V., Dean, P.N., Anderson, A.L., Fisher, J.C., in: Advances in Radiation Protection Monitoring, Vienna: IAEA, 1979, p. 493. Tait, W.H.: Radiation Detection, London: Butterworth & Co., 1980. International Commission on Radiological Protection, ICRP Publication 38. Ann. ICRP 11-12, Oxford and New York: Pergamon Press, 1983. Toohey, R.E., Keane, A.T., Rundo, J.: Health Phys. 44, Suppl. 1 (1983) 323. Andrasi, A. Beleznay, E. and Urban, J. In: Assessment of Radioactive Contamination in Man, Vienna: IAEA STI/PUB/674 (1985) 165. Newton, D., Wells, A. C., Mizushita, S., Toohey, R. E., Sha, J. Y., Jones, R., Jefferies, S. J., Palmer, H. E., Riekst, G. A., Anderson, A. L., Campbell, G. W., In: Assessment of Radioactive Contamination in Man, Vienna: IAEA STI/PUB/674 (1985) 183. Shen, C., Wen, H., Zheng, W., Zhao, Y., Tang, M., Ye, C., Wu, D., In: Assessment of Radioactive Contamination in Man. Proc. Symp. Paris, France, 19-23 November 1984, Vienna: IAEA STI/PUB/674 (1985) 123. Sumerling, T.J., McClure, D.R., Massey, D.K.: NRPB-R188, London: HMSO, 1985. International Organization for Standardization, International Standard ISO9000, Geneva: ISO, 1987.
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88ICR International Commission on Radiological Protection, ICRP Publication 54. Ann. ICRP 19 (13), Oxford and New York: Pergamon Press, 1988. 91Too Toohey, R.E.: Health Phys. 60, Suppl. 1 (1991) 7. 92Del Delaney, C.F.G., Finch, E.C.: Radiation detectors, Oxford: Clarendon Press, 1992. 92Zik Zikovski, L., Kennedy, G.: Health Phys. 63 (1992) 449. 94And Andrasi, A., Henrichs, K., Bogner, L.: Report EUR 15395 EN, 1994. 94IAE International Atomic Energy Agency. TECDOC-746, Vienna: IAEA, 1994. 94ICR International Commission on Radiological Protection, ICRP Publication 68. Ann. ICRP 24(4), Oxford and New York: Pergamon Press, 1994. 94Skr Skrable, K. W., Chabot, G. E., French, C. S., LaBone, T. R.: Internal Radiation Dosimetry (RAABE, O.G., Ed.), Medical Physics Publishing, Madison, WI (1994) 431. 95Doe Doerfel, H.: Proc. Symposium on Radiation Protection in Neighboring Countries in Central Europe, Portoroz, 1995. 95Hub Hubbel, J.H., Seltzer, S.M.: National Institute of Standards and Technology, NISTIR 5632, 1995. 95Mal Mallett, M.W., Hickman, D.P., Knuchen, D.A., Poston, J.W.: Health Phys. 68 (6) (1995) 773. 96HPS Health Physics Society: Performance Criteria for Radiobioassay: American National Standards Institute HPS N13.30-1996, McLean, VA: Health Physics Society, 1996. 96IAE International Atomic Energy Agency.: Safety Series No. 114, Vienna: IAEA, 1996. 96Kra Kramer, G.H., Loesch, R.M., Olsen, P.C.: Proc. 1996 International Congress on Radiation Protection, Vienna,. 2 (1996) 409. 97ICR International Commission on Radiological Protection: ICRP Publication 78. Ann. ICRP 27(3/4), Oxford and New York: Pergamon Press, 1997. 98ISO International Organization for Standardization: International Standard ISO/WD 119298/ISO/TC85/SC2/WG5, Geneva: ISO, 1998. 99Kno Knoll, G.F.: Radiation Detection and Measurement, 3rd edn, New York: John Wiley & Sons, 1999. 99USD United States Department of Energy, DOE Standard DOE-STD-1112-98, 1999. 00And Andrasi, A.: Radiat. Prot. Dosim. 89 (3-4) (2000) 229. 00Doe Doerfel, H., Andrasi, A., Bailey, M.R., Birchall, A., Castellani, C.-M., Hurtgen, C., Jarvis, N., Johansson, L., LeGuen, B., Tarroni, G.: Forschungszentrum Karlsruhe GmbH, FZKA 6457, 2000. 00Lop López Ponte, M.A., Bravo, T.N.: Radiat. Prot. Dosim. 89 (3-4) (2000) 221. 01Ish Ishigure, N., Nakano, T., Enomoto. H.: Radiat. Prot. Dosim. 97 (3) (2001) 271. 01Ram Ramzaev, V., Ishikawa, T. Hill, P., Rahola, T., Kaidanovsky, G., Yonehara. H., Hille, R., Uchiyama, M.: Radiat. Prot. Dosim. 98 (2) (2002) 179. 02Kra Kramer, G.H., Hauck, B.M.: Radiat. Prot. Dosim. 102 (4) (2002) 323. 02Sin Singh, I.S., Suri, M.M.K., Vidhani, J.M., Garg, S.P., Sharma, R.C.: Radiat. Prot. Dosim. 102 (2) (2002) 145. 03Doe Doerfel, H., Andrasi, A., Bailey, M.R., Berkovski, V., Castellani, C.-M., Hurtgen, C., Jourdain, J.-R., LeGuen, B.: Radiat. Prot. Dosim. 105 (1-4) (2003) 645. 03Hun Hunt, J.G., de S. Santos, D., da Silva, F.C., Malatova, I., Foltanova, S., Dantas, B.M., Azaredo, A.: Radiat. Prot. Dosim. 105 (1-4) (2003) 549. 04BMU Bundesministerium für Umwelt, Naturschutz und Reaktorsicherheit: Richtlinie zur Ermittlung der Körperdosis bei innerer Strahlenexposition, Bundesanzeiger, to be published (2004).
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10.3.3 In vitro measurements: excretion analyses 10.3.3.1 Introduction In section 7.5.1.2 “Analysis of excreta and other biological materials” the general recommendations of the ICRP in its Publication 78 [78ICR] on excreta monitoring programmes are summarized. The selection, use and interpretation of various bioassay approaches are based on the physical and biokinetic characteristics of the particular radionuclides considered. In vitro analysis refers to the identification and quantification of radionuclides in the body by analysis of material excreted or removed from the body. The main sources of bioassay data are urine, faeces, breath and blood. Other samples such as hair, teeth, saliva and nails have been employed in special cases. These biological samples provide an indirect measure of the internal radionuclide deposition because there is no direct information about the body or organ burdens. Proper interpretation of these results requires knowledge of the relationship between the presence of a radionuclide in the various bioassay samples and the organ radionuclide burdens of interest. Various factors influence the applicability of any particular type of sample: • • • • • •
The chemical element involved, Its physical and chemical form, The magnitude of internal deposition, Biological and physical half-lives of the radionuclides involved, Time elapsed since the intake occurred, and Sensitivity of the analytical and measuring method used.
For radionuclides emitting non-penetrating radiation, i.e. radiation being absorbed in the body, excretion analysis is the common method for monitoring workers exposed to radioactive material. This is the case with radionuclides with alpha particle radiation (such as thorium, uranium, plutonium, americium, curium) and radionuclides with soft beta particle radiation (3H, 14C, 35S, and others). In vitro techniques for evaluating the internal contamination include also time-consuming procedures for processing urine or faeces samples. Therefore, this technique should be used when in vivo methods are not applicable and the exposure is at low levels. A critical point of the in vitro analysis is the sampling. For the application of common assessment models the excretion samples have to be collected in a well defined time interval. Another important point is that the collected bioassay samples have to be free of contaminations from outside. For the interpretation of the data the biological variability of the excretion of a person, which can lead to different fractions of total body activity to be excreted in daily samples, must be considered. Consideration should be also given to whether medical interventions (chelating therapy, administration of diuretes, blocking agents and so on) could have influenced the pathway or excretion rates of radionuclides. 10.3.3.2 Urine samples Analysis of urine samples for excreted radionuclides is the method used most frequently for routine monitoring and assessment of internal contamination. Urine samples are easy to collect and rather reliably interpreted for material readily absorbed in the gastrointestinal (GI) tract. A radionuclide in a relatively transferable (soluble) form entering the body reaches the bloodstream and a fraction of it is deposited in various body organs. The remainder is excreted predominantly in the urine. This biokinetic behaviour depends on the chemical form of the radionuclide involved entering the body, and from its metabolic behaviour in the body after incorporation. Typical radionuclides which will be monitored routinely via urine samples are for example 3H, 32P, 35S, 89Sr, 147Pm, thorium, americium and other alpha-emitting radionuclides (see also Table 10.24).
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The nominal daily excretion rate of urine amounts to 1.2 l for females and 1.6 l for males [89ICR]. The individual excretion rate depends strongly on physiological and environmental conditions but also on individual nutritional habits. Therefore, the general sampling practice for routine monitoring is to collect 24-h-urine samples or equivalent. Repetitive sampling helps determining the time-dependent rate of excretion of a radionuclide after intake of the radionuclide into the body. Radioactive material can be lost from solution by adsorption onto surfaces of some containers, and so on. For this reason samples must often be stabilised until analysis by refrigeration or freezing. Other methods are the addition of a carrier or of an acidic, basic or other preservative as is appropriate for the particular situation. 10.3.3.3 Faeces samples Collection and analysis of faecal samples is another means of obtaining an indirect assessment of possible internal contamination. For routine monitoring, faeces are not used as often as urine, but analysis of faeces can provide at least qualitative information, particularly for relatively insoluble radionuclides. Faeces samples are also very helpful for the quick assessment in the case of an extraordinary situation: The excretion of radionuclides by faeces specifically of those with low gastrointestinal absorption (see Chapter 7) is very often faster than by urine. When an intake occurrs by ingestion, the quantity of a radionuclide being excreted soon after ingestion represents the fraction of the radionuclide that has not been absorbed during the passage through the GI tract. In the case of inhalation there are two fractions; one fraction which is absorbed from the respiratory tract enters the bloodstream and is partly deposited in various organs. A part of it is subsequently excreted from the liver into the GI tract via the bile into the faeces. The second contribution comes from radionuclides translocated by swallowing from the respiratory tract directly into the GI tract and is partly retained from the small intestine or excreted via the faeces (see Chapter 7). Long-term excretion of a radionuclide by faeces after its intake into the body is originating from delayed clearance of insoluble material from the pulmonary region of the respiratory tract or from the clearance of material that has entered the bloodstream and is excreted from the liver into the GI tract via the bile. There are only few radionuclides for which a routine monitoring should be based on faecal excretion analysis: 90Y, 147Pm (inhalation type S), thorium, curium and other alpha-emitting radionuclides (see also Table 10.24). Such a monitoring is adequate to identify incorporations which occurred just before sampling. Additionally, annual or biannual monitoring by faecal sampling may be used to check the reliability of air monitoring. On the basis of air monitoring results of a hypothetical faecal excretion rate can be derived; the comparison of measured faecal analysis can help to exclude a significant underestimation. The general sampling practice for routine monitoring is to collect 24-h-faecal samples. The nominal transit time for material passing directly through the GI tract is about two days [30ICR], but this varies considerably with diet, health of the individual, and other factors. For this reason also a collection time of three consecutive days for faecal samples is recommended, in order to obtain reliable estimates of daily excretion rates. Faecal samples are particularly subject to biodegradation. Therefore, they should be analysed promptly, ashed or preserved by deep freezing. 10.3.3.4 Exhalation Breath samples can be useful in the case of incorporation monitoring for determining the amount of a radionuclide leaving the body by exhalation, i.e. in gaseous form such as 222Rn, 220Rn, 14C-labelled carbon dioxide or tritiated water vapour. In the case of tritiated water about one third of an intake is excreted via breath, whose specific activity rapidly reaches equilibrium with that in body water [00IAE].
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For example, 220Rn exhalation measurement [00Eis] allows the individual determination of 228Th body burdens without chemical preparation. That means, 228Th will be measured by its decay product 220Rn (daughter of 224Ra) in the exhaled air of a person with thorium burden. So the worker has for breath sampling only to breathe into a collecting apparatus for up to 30 minutes, depending on the volume required. The detection limit of this method is about 1 Bq of 228Th in the lung and thus comparable to that of urinary excretion analysis. So this exhalation measurement method is best used to complement other assessment methods. Quantification of 232Th by measurement of 220Rn in exhaled air requires additional information about the nuclide spectrum because of the unknown activity ratio between 228Th and 232Th in the body. For example, the diet and the mineral water could lead via 228Ra to an increased 228Th burden. 10.3.3.5 Other biological samples Other biological samples, such as blood, hair, teeth, nails and nose blows can be used only as indicators for intake of radionuclides. Due to lack of biokinetic models or uncertainties in those models and data used for qualitative assessment of any internal contamination the results of assessments are generally not useful as a basis for quantitative dose estimations. Blood samples provide the most direct source for estimating circulating internal contamination. But the majority of radionuclides are rapidly cleared from the blood. So, and because of recirculation in the body, measurements of activity in blood are generally only poor indicators of the total systemic content. The analysis of nose blow samples can supplement monitoring for the purpose of screening for intakes, and give valuable information on the nature of the inhaled contaminant. Usually this method triggers other types or complementary analyses, such as urine or faeces samples. Hair samples have been analysed for plutonium [81Too]. Caution is needed to ensure that hair care products have not resulted in contamination with naturally occurring radionuclides, such as uranium. Teeth incorporate many of the bone-seeking elements, such as strontium, and may provide an indication of long-term, e.g. childhood exposures [95Hen]. 10.3.3.6 Radiochemical analyses A large number of different in vitro techniques have been developed for the detection and quantification of low-level activity of radionuclides in excreta. Table 10.24 gives an overview on some typical radionuclides and the complexity of the analytical procedure needed. Table 10.24. F = faeces) Radionuclide 3 H 14 C 32 P 90 Sr 131
Examples for some typical radionuclides for monitoring by in vitro techniqes (U = urine, Inhalation type, Chemical compound HTO org. F, M F, S
I Pm 232 Th
F M M, S
239
M
147
Pu
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Biological sample
Analytical procedure
U U U U F U U U F U F
simple simple simple elaborate elaborate simple elaborate elaborate elaborate elaborate elaborate
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Explanation Radionuclide: Inhalation type:
The radionuclide possibly suitable for incorporation monitoring. Inhalation type of the appropriate radionuclide specified as F = fast lung absorption, M = moderate lung absorption, and S = slow lung absorption. Biological sample: Determination of the activity in urine (U) samples or in faeces (F) samples. Analytical procedure: A rough overview on the extent of the analytical procedure (simple, elaborate) necessary to determine the radionuclide in the appropriate sample.
In general it is possible to classify the procedure for the determination of the activity concentration of the radionuclides in the following four main groups: • Alpha particle spectrometry of urine samples, using elaborate and time consuming radiochemical procedures, e.g., for 232Th, 241Am, 239Pu; • Beta counting of urine samples without or with simple radiochemical procedures, e.g., for 3H, 14C, 32P; • Beta counting of urine or faecal samples using elaborate and time consuming radiochemical procedures, e.g., for 90Sr and 147Pm; and • Gamma counting of urine samples without or with simple preceding radiochemical procedures, e.g., for 60Co, 134Cs, 137Cs and 131I. Regardless of which in vitro technique is used, the sample activity is calculated after the measurement by the equation
Ac = (
N s Nb 1 1 1 − )⋅ ⋅ ⋅ ts tb R E V
(10.3.3.1)
where Ac = sample activity concentration in Bq per volume unit Ns = number of counts observed in the sample during the counting time ts Nb = number of counts of the background during the counting time tb ts = sample counting time tb = background counting time R = chemical recovery, expressed as a fraction E = counting efficiency, expressed as a fraction V = sample size as volume unit If the counting times are equal the equation above will reduce to
1 1 1 1 A c = ⋅ (N s − N b ) ⋅ ⋅ ⋅ t R E V
(10.3.3.2)
When an internal tracer is added, e.g., in the case of alpha particle spectrometry, the activity may be calculated using this equation
⎛ N −N b A = At ⋅ ⎜ s ⎜ N t − N b' ⎝
⎞ ⎟⋅K ⎟ ⎠
(10.3.3.3)
Where A = sample activity At = activity of the internal tracer added to the sample Nt = number of counts observed in the tracer region of interest (ROI) Nb‘ = number of counts of the background in the tracer ROI K = calibration factor applicable for the sample volume
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In this case it is not necessary to know the chemical recovery and the efficiency. The minimum detectable activity (MDA) corresponds to the level of activity which is required to ensure with some chosen level of confidence that the net signal will be detected. The definitions and equations are specified in the Safety Standards Series No. RS-G-1.2 [99IAE] and Safety Reports Series No. 18 [00IAE]. See also Section 10.3.2.8.
MDA = 3.3 ⋅
Nb ts
⎛ t ⎞ 1 1 1 ⋅ ⎜⎜1 + s ⎟⎟ ⋅ ⋅ ⋅ ⎝ tb ⎠ R E V
(10.3.3.4)
When α, the probability of a type I error (false positive), and β, the probability of a type II error (false negative), are both set equal to 0.05, MDA may be calculated in most cases as shown above (the symbols are the same as in the equations before). In excretion analysis samples usually contain only low activities of a radionuclide. To ensure the reliable detection of such small activities low detection limits are essential. Therefore the following technical items have to be taken into consideration: • The radiochemical recovery must be as high as possible. • In general the counting or detection efficiency for alpha or beta particles cannot be higher than 50 % in the case of measuring flat source discs, because the solid angle seen by the detector cannot be more than 2π. On the other hand the counting efficiency is a function of the distance between detector and source and varies for alpha particle spectrometry from a few percent up to 40 %. Therefore it is necessary to get an optimal adjustment between counting efficiency and energy resolution needed. • A higher volume of the sample leads to a lower MDA per volume unit (see equation (10.3.3.4)). • A lower number of background counts will result in a lower MDA. A main point to ensure this requirement is the careful radiochemical separation of the radionuclides which have to be determined. • The MDA is inversely proportional to the counting time. That means the sample counting time should be as long as possible, but it depends on the time available. Equations for the calculation of MDA for more complicated cases will be found in the Safety Reports Series No. 18 [00IAE]. Whatever technique is used for counting, most of the radionuclides analysed and especially the actinides, need to be isolated from the matrices. Numerous analytical procedures have been developed. They all are based on the same principle which consists of • Sample preparation, • chemical separation, and • source preparation. The sample preparation for urine involves wet ashing or co-precipitation of calcium phosphate, calcination and dissolution of the precipitate. Faeces are ashed and dissolved in acid. Insoluble materials such as silica are treated with fluorhydric acid. Similar procedures are used for tissue samples. The chemical separation of the radionuclide to be analysed includes separation and purification on ion-exchange resin or by solvent extraction, or a combination of these two techniques. The type of source preparation used is a function of the following measuring technique needed for the analysis. For the alpha particle spectrometry of actinides the source is prepared by • Direct evaporation, • co-precipitation with lanthanide fluoride, or • electrodeposition. Landolt-Börnstein New Series VIII/4
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For beta counting with a proportional counter, the source is generally obtained by precipitation and filtration of the insoluble salt. For liquid scintillation counting (LSC) the purified radionuclide solution is mixed with an appropriate cocktail for the measurement. The radiochemical analyses carried out routinely in the laboratory for monitoring potential incorporations of occupationally exposed workers has to be documented [00ISO2]. These written procedures shall include all steps starting with the receipt of the sample at the laboratory to measurement of the sample, or of an aliquot of the sample, and should contain all radiochemical procedures used. For example, the description of the procedures can be very different: • In the simple case of analysing tritium (see Table 10.24) it is only necessary to describe how to make the aliquot of the urine sample and the adding of an appropriate scintillation substance to be ready for the measurement by liquid scintillation counter. • But in most cases the analytical procedure is more complicated and time-consuming, as seen in Table 10.24. In connection with the sample preparation (urine, faeces, tissue, blood etc.) and after adding an internal tracer to determine the chemical recovery, several nuclide-specific radiochemical separation steps have to be done. After electrodepostion on a stainless steel disk the radionuclide activity in the sample can be determined, e.g., by alpha particle spectrometry. A lot of different analytical procedures have been published. A very fundamental publication in this field is the HASL-300 (Health and Safety Laboratory) document [97HAS]. It is well known as „The procedure manual of the Environmental Measurements Laboratory (EML)“ and covers the existing technology and procedures currently in use at EML (older procedures are updated and new procedures are added). This voluminous manual is also available as CD-ROM. The main task of this document is the analytical chemistry to be used for a wide range of radionuclides (e.g., 3H, Fe, Sr, Tc, Pb, Po) with different matrices and measuring techniques. Detailed descriptions for the determination of thorium in urine and faeces, used in different laboratories in Germany, are given by Riedel et al. [93Rie]. The contribution of Harduin et al. [96Har] describes the analytical determination, especially for actinides in biological samples. Instructions about sequence analysis of actinides and 90Sr are found in the work of Wihlidal et al. [98Wih]. A fundamental and very informative technical note about the electrodeposition of actinides is given by L. Hallstadius [84Hal]. An important application of these analytical procedures and measuring techniques is the monitoring of workers involved in the decommissioning of nuclear facilities, because of the great variety of the radionuclides present and the conditions of the exposure. Establishing appropriate monitoring programmes and procedures is currently in progress. Due to the presence of transuranic radionuclides the analysis of excretion samples is required. So, for example, the contribution of Robredo et al. [00Rob] describes an excellent radiochemical procedure for the determination of americium and plutonium in urine samples. Neudert et al. [99Neu] present a very fast and closely method by using inductively coupled plasma-mass spectrometry (ICP-MS). In case of monitoring potential incorporations of occupationally exposed workers, in some cases consideration has to be given to radionuclides incorporated from natural sources via food, especially drinking water. These intakes result in contributions of activities measured when monitoring workers and may mislead the dosimetry of occupationally exposed workers. Therefore it is necessary to have information on the natural contents, especially of thorium and uranium in human urine and faeces; see for example [97Rot] and [98Bey]. 10.3.3.7 Measuring techniques There exists a wide field of different measuring techniques to determine the activity of alpha and beta emitting radionuclides (alpha particle spectrometry, beta counting, liquid scintillation counting (LSC), fluorimetry, laser induced fluorimetry, gamma spectrometry, inductively coupled plasma-mass spectrometry (ICP-MS), neutron activation analysis (NAA) and delayed neutron activation analysis (DNAA)).
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Table 10.25 gives a rough overview of these different techniques for the determination of radionuclides in excreta, including the values for the minimum detectable activity (MDA). The main measuring techniques for the determination of radionuclides in excretion samples are summarised below. Alpha particle spectrometry This technique is the most commonly applied technique for measuring the isotopes of the different alpha emitters, such as thorium, uranium, plutonium, americium and curium. Alpha particle spectrometry is used to identify the isotopes and to quantify their activities. Because of the non-penetrating radiation of the alpha emitter, the alpha particle spectrometry requires prior to the measurement an elaborate and time-consuming radiochemical procedure, as described above. For a quantitative nuclide-specific separation a very thin source is very important to get a good energy resolution of the alpha particle spectrum. So the common source preparation technique used is the electrodeposition on a stainless steel disk. This flat source will be measured in an alpha chamber in connection with a multichannel analyser to determine the different isotopes [84Hol]. By using internal tracers such as 229Th, 232U, 242Pu and 243Am the radiochemical recovery is simple to determine. This also allows to measure all isotopes of the element analysed in the same spectrum with the same MDA. Beta counting Radionuclides such as 32P, 89Sr, 90Sr or 131I can be measured after the chemical separation with a low background proportional counter. The technique is applicable for beta emitters of relatively high energy. As for alpha particle spectrometry, this technique requires the same type of radiochemical procedure as described before. The preparation of the source is generally obtained by precipitation and filtration of the unsoluble salt. Table 10.25. Measuring techniques for the analyses of excreta Measuring Radionuclide Analytical technique procedure Alpha particle Th, U, Pu, Am, Cm elaborate spectrometry 32 Beta counting P, 89Sr, 90Sr elaborate 3 LSC H, 14C, 63Ni, no 241 Pu elaborate 32 P, 89Sr, 90Sr elaborate 60 Gamma Co, 134Cs, 137Cs no spectrometry Fluorimetry U nat simple Laser-induced U nat simple fluorimetry 232 ICP-MS Th, no 238 U, no 239 Pu, elaborate 240 Pu elaborate 238 NAA U, elaborate 232 Th elaborate 235 DNAA U, no U nat, no U nat simple
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Typical MDA 1.0 mBq/l <80 mBq/l 5 Bq/l 15 mBq/l 120 mBq/l 200 mBq/l 25 mBq/l 2 mBq/l 0.05 mBq/l 0.1 mBq/l 0.1 mBq/l 0.4 mBq/l 0.0025 mBq/l 0.1 mBq/l 60 mBq/l 25 mBq/l 0.5 mBq/l
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Measurement with a low background proportional counter will lower the MDA as compared to measurement by liquid scintillation counting (LSC). Liquid scintillation counting is especially useful for beta emitters of weak energy such as 3H, 14C and 63 Ni. For these radionuclides a direct measurement can be performed by mixing a small volume of the urine sample with the scintillation cocktail. Internal standards are often used to measure low-level radioactivity. Liquid scintillation counting is also used to determine pure beta emitters like 32P, 89Sr and 90Sr. For the measurements of these radionuclides an elaborate radiochemical procedure is needed for separating the analyte from the matrix. Fluorimetry Fluorimetry is a simple and fast technique for the determination of uranium in urine. Uranium is determined by the fluorescence produced when exposed to ultraviolet light. The urine sample may directly be fused in a platinium dish and measured with a fluorimeter. The limitation of this technique is its poor sensitivity and the measurement of total uranium only, mainly 238U. Laser-induced fluorimetry The laser-induced fluorescence excitation technique needs some treatment of the urine like co-precipitation and calcination, or wet ashing before measuring the uranium content. Gamma spectrometry In many cases gamma spectrometry can be directly applied on the urine samples. It should be noticed that for gamma emitters such as 60Co, 134Cs, 137Cs and 131I in vivo measurements are mainly performed. Inductively coupled plasma-mass spectrometry (ICP-MS) The sample is introduced into a mass spectrometer by means of a nebuliser in a plasma torch, and afterwards in a magnetic field that separates the atoms in their different masses and producing thus results for each of the isotopes of the element analysed. This technique has been used for uranium and thorium analyses in urine and human tissues. In fact, this technique is very sensitive for radionuclides with very long half-life such as 235U, 238U and 232Th. Beside acidifying the urine sample no other treatment is needed. This mass spectrometry is also a very sensitive technique for measuring isotopic composition of plutonium. For plutonium analysis a radiochemical procedure is needed to separate plutonium from the bulk of inorganic material and from 238U which causes interferences at the mass 239, before being subjected to the mass spectrometric analysis. This technique is capable of measuring 239Pu and 240Pu separately, which is not possible by alpha particle spectrometry. A clear advantage of ICP-MS is its rapidity. After treatment of the sample, if needed, the results can be obtained in a few minutes; see for example [98Kre], [02Sch] and [03Bou]. Neutron activation analysis (NAA) The neutron activation analysis has been used for the rapid determination of thorium in urine and in biological fluids, like serum. The technique involves some radiochemical procedures before and after irradiation of thorium with thermal neutrons, see for example [94Hub]. The photons of 233Pa obtained by irradiation of 232Th are counted. This technique is sufficiently sensitive to detect the natural content of 232 Th in the environment, especially in food.
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The NAA technique has also been reported for the determination of 238U in urine. In this case the photons of 239Np will be counted. Delayed neutron activation analysis (DNAA) Delayed neutron assay has been used for measuring 235U. As a screening method for uranium in urine, the evaporated urine was directly irradiated by thermal neutrons in a nuclear reactor. Then the delayed neutrons resulting from fisson of uranium were counted. Good compilations about this measuring technique are found in [84Gab], [89Gla] and [94Hub]. For the determination of uranium in urine there is nearly no sample preparation necessary. 232 Th could be fisured only by epithermal neutrons. For the analysis of thorium in urine samples a radiochemical separation technique is required. DNAA is a very fast technique but measures only 235U, requiring information concerning the isotopic composition of uranium. For natural uranium a lower MDA can be achieved if uranium is isolated by adequate radiochemical separation. 10.3.3.8 Quality assurance Quality assurance (QA) is an essential and integral element of the routine work in a laboratory to ensure the reliability of the bioassay data yielded in this laboratory. The QA practice in the laboratory can be achieved, e.g., by: • Establishing a certified quality management system • Accreditation of all analytical procedures and measuring techniques used in the routine work of this laboratory. The normative requirements on a certified quality management system are laid down in EN ISO 9001: 2000 [00ISO1]. For the accreditation normative requirements exist for testing laboratories (i.e. „producing activity values“) in EN ISO/IEC 17025: 2000 [00ISO2]. In the case of laboratories performing inspections (i.e. „producing activity values including interpretation, e.g., dose values“) normative criteria are specified in ISO/IEC 17020: 1998 [98ISO]. General and helpful support is given in the paper „Quality management systems for technical services in radiation safety“ [03IAE]. The laboratory has to establish an in-house quality assurance plan to prove its organisational and technical competence. This plan should consist of two sections: the more formal part of „administration support“, and the technical part of „quality control“. In the first section the plan should include, among others, the following items: • • • • •
Organisational responsibilities Corrective actions Personnel qualifications Adequate operational environment Documentation of all aspects of the bioassay monitoring programme. The technical part of the QA plan should include procedures as follows:
• • • • • •
Sample registration procedures Administrative procedure for each sample Procedure manuals for the radiochemical analyses and measuring techniques Instrumentation and calibration manuals Use of control charts and testing materials Participation in intra- and interlaboratory comparisons.
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More detailed instructions are given in IAEA Safety Reports Series No. 18 [00IAE] and in ISO 12790-1 [01ISO]. The design and implementation of a QA plan is described in the Safety Guide No. RSG-1.2 [99IAE]. In order to avoid systematic errors and ensure the quality of the analyses and techniques used, many additional measurements and methods have to be performed in the laboratory. For example, routine checks are necessary on: • • • • •
Radiochemical recovery Energy calibration Efficiency calibration Background measurements Blind analyses (blanks).
Based on the laboratory practice of many years, particularly the following problems have to be considered. Uncritical observance of analysis procedures Generally analysis procedures are given for a defined matrix, e.g. for urine samples with an average content of mineral salt. In practice, however, the salt content is occasionally very high, e.g. from the administration of calcium tablets or excess consumption of lemonades with high phosphate content. In this case procedures should be modified by introducing additional separation steps or by using more efficient separation methods. Avoiding cross-contamination Cross-contamination, particularly in monitoring incorporations of alpha emitters, may lead to error analyses with serious consequences. They occur when samples with very different concentrations are measured with the same laboratory equipment. When uranium and thorium isotopes are determined it is appropriate to conduct the measurement of urine and faeces samples in separate locations and with different equipment, because their natural excretion is quite different. Errors during internal standardisation In the course of internal standardisation being commonly used in radiochemical analyses, the standard may contribute to the lasting inaccuracy of whole series of measuring results. This may be caused from, among others, errors of production or inappropriate storage and application. In some cases a “cleaning” of the standard from daughter nuclides is required immediately before it is used for analysis procedures. Interference due to blind values from chemicals To assess natural 232Th excretions in urine, identification limits of a few µBq excretions per day are required. In this range even p.A qualities of the used chemicals may lead to relevant contamination. Thus in concentrated hydrochloric and nitric acid (each at p.A. quality) up to a maximum of 3 mBq 232Th per litre have been proved.
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10.3.3.9 Examples for dose estimations from in vitro measurements Case 1: Routine monitoring of depleted uranium (238U) Handling: Depleted uranium, ICRP inhalation type M, routine monitoring by analysis of the urine, two times per year (every 180 days) Measured value m = 21 mBq/d (24 h urine) Dose assessment by using the following standard assumptions: • • • •
Route of intake: acute inhalation Date of incorporation: in the middle of the monitoring interval AMAD: 5 µm Biokinetic and dosimetric data: ICRP 68 and ICRP 78.
The committed effective dose is calculated by the equation: E=
e(50) ⋅ m E (∆ t ) E
(10.3.3.5)
where E = committed effective dose in Sv e(50) = dose coefficient for the effective dose in Sv/Bq m = value of the 24 h excretion in Bq/d, corrected at the end of the collecting period EE (t) = excretion rate (EE = EU for urine), at the day t in Bq/d after acute inhalation of 1 Bq ∆t = in the case of dose assessment by using standard assumptions half-time of the monitoring interval in days Result: ⇒ Committed effective dose E = 0.28 mSv half-yearly, calculated for 238U with EU(t) = 1.2 ⋅ 10−4 Bq/d Bq−1 at t = 90 days for 238U, tabulated in [03Nos] e(50) = 1.6 ⋅ 10−6 Sv/Bq for 238U, tabulated in [78ICR] Case 2: Practical estimation of the date of incorporation of 35S Handling:
35
S, ICRP inhalation type F, routine monitoring by analysis of the urine monthly, i.e. every 30 days Data from the 1st monitoring measurement: Date of collecting the urine: 30.04./01.05.2001, 24 h urine. Collected urine sample: 1100 ml, measured value 37 Bq/5 ml urine. Measurement result m = 8.1 kBq/d Dose assessment by using the same standard assumptions as in case 1, but intake and committed effective dose will be calculated in two steps, derived from the equation (10.3.3.5) in case 1: intake I = m/ EU(t ) = 7.4 MBq and committed effective dose E = I ⋅ e (50) = 0.59 mSv monthly with EU (t ) = 1.1 ⋅ 10−3 Bq/d Bq−1 at t = 15 days, tabulated in [03Nos] e (50) = 8.0 ⋅ 10−11 Sv/Bq for 35S, tabulated in [03Nos]
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Problem: Notice of a significant content of 35S in the urine of one worker, but there was no remarkable irregularity (contamination, accident and so on) happened in the laboratory during the monitoring interval. Therefore further measurements were done to try an estimation of the date of the incorporation and the amount of the intake to get, on this basis, a more realistic effective dose value: Collecting date Amount of urine Collecting time Measured value Remarks [ml] [h] [Bq/5 ml urine] 30.04./01.05.01 1100 24 37 1st montoring 03.05./04.05.01
1300
24
2.1
special
07.05./08.05.01
600
12
2.1
measurements
02.06./03.06.01
1050
24
0.75
2nd monitoring
To obtain the date of incorporation and the amount of the intake of 35S the following information of the radiation protection officer were used for the interpretation: • acute inhalation, and • inhalation type F and AMAD 5 µm of the 35S compound involved. The procedure consists of the following steps: 1. Variation of the date of the incorporation. 2. Calculation of the belonging intake for each of the four measured values. 3. Calculation of the mean intake and the standard deviation. The mean value with the minimum standard deviation SD should be the realistic time of incorporation. 1. 2. 3. 4. Mean SD*) Measurement Measurement Measurement Measurement value Measured value 8140 Bq/d 546 Bq/d 504 Bq/d 158 Bq/d 27.04.2001: 4d 7d 11 d 36 d Time ∆t −1 −1 −1 0.0019 d 0.0016 d 0.0014 d 0.0005 d−1 EU (∆t ) 4284 kBq 341 kBq 360 kBq 316 kBq 1325 kBq 130 % I = m/ EU (∆t ) 28.04.2001: 3d 6d 10 d 35 d Time ∆t 0.0035 d−1 0.0017 d−1 0.0014 d−1 0.0005 d−1 EU (∆t ) 2326 kBq 321 kBq 360 kBq 316 kBq 831 kBq 104 % I = m/ EU (∆t ) 29.04.2001: 2d 5d 9d 34 d Time ∆t −1 −1 −1 0.028 d 0.0017 d 0.0015 d 0.0005 d−1 EU(∆t ) 291 kBq 321 kBq 336 kBq 316 kBq 316 kBq 5.1 % I = m/ EU(∆t ) 30.04.2001: 1d 4d 8d 33 d Time ∆t 0.29 d−1 0.0019 d−1 0.0015 d−1 0.0005 d−1 EU (∆t ) 28 kBq 287 kBq 336 kBq 316 kBq 242 kBq 52 % I = m/ EU (∆t ) *) standard deviation
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Result of the measurements: The date of incorporation is 29.04.2001, as shown in the table above (the standard deviation is minimal at this date). The mean value of the intake I = 320 kBq. The committed effective dose E = 0.25 mSv Using this practical procedure in many cases it is possible to get infomation about the date of incorporation (as shown in the example above), the route of intake and the value for the AMAD. Case 3: Accidental inhalation of 239Pu [99Kau] Situation: A worker inhaled contaminated aerosol for aproximately 10 minutes, 239Pu-oxide is involved The following measurement results were achieved: Time after the incident [d] 1 2 3 10 15
Faecal activity [Bq] 160 0.62 0.44
Urinary activity [mBq] 1.6 1.1 -
Each measurement value will be interpreted in terms of intake using the ICRP Publication 78 [78ICR]: • The first faecal measurement was a measurement of the pool of all faecal excretion during the first three days after the incident. To interpret this it is necessary to compare this value with the sum of the parameters for the first three days given in [78ICR]. • Unfortunately for 15 days no value is tabulated in [78ICR]. However, in the table for routine monitoring the value for the measurement period 30 days is suitable for this purpose because for routine monitoring an acute intake in the middle of the monitoring interval is assumed, i.e. 15 days before the measurement for a monitoring interval of 30 days. • Because 239Pu has been inhaled as an oxide, inhalation type S is assumed. With these assumptions the following intake values were derived, using equation (10.3.3.5): • for fecal excretion measurements I1 = m / [EF (1) + EF (2) + EF (3)] = 452 Bq I2 = m / EF (10) = 954 Bq I3 = m / EF (15) = 898 Bq • for urinary excretion measurements I4 = m / EU (1) = 696 Bq I5 = m / EU (3) = 1320 Bq
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Using the following data for EF (t ) and EU (t ), tabulated in [78ICR]: Time after intake Urinary excretion Faecal excretion [d] [Bq/d Bq−1] [Bq/d Bq−1] −6 1 1.1 ⋅ 10−1 2.3 ⋅ 10 2 1.6 ⋅ 10−1 3 8.3 ⋅ 10−7 8.4 ⋅ 10−2 10 6.5 ⋅ 10−4 ) 15* 4.9 ⋅ 10−4*) *) Data for routine monitoring with an time interval of 30 days Results: In this case urinary and faecal excretion analyses give similar results. If several successive results are available, the best estimation of the intake I is obtained by taking the geometric mean of the Ii values established from these measurements. So in this case the geometric mean of all intake values is 810 Bq. ⇒ Committed effective dose E = 6.7 mSv with e(50) = 8.3 ⋅ 10−6 Sv/Bq for 239Pu, tabulated in [78ICR] In the case that there would have been an excretion enhancement by DTPA it would be necessary to consider this by some modifications, for example by divison of the measurement values by an appropriate number.
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10.3.3.10 References for 10.3.3 30ICR 78ICR 81Too 84Gab 84Hal 84Hol 89Gla 89ICR 93Rie 94Hub 95Hen
96Har 97HAS 97Rot 98Bey
98ISO 98Kre
98Wih 99IAE
International Commission on Radiological Protection: Limits for intakes of radionuclides by workers; ICRP Publication 30, Oxford: Pergamon Press, 1979. International Commission on Radiological Protection: Individual monitoring for internal exposure of workers; ICRP Publication 78, Oxford: Pergamon Press, 1997. Toohey, R.E., Cacic, C.G., Larsen, R.P., Oldham, R.D.: The concentration of plutonium in hair following intravenous injection; Health Phys. 40 (1981) 881. Gabelmann, H., Lerch, M., Kratz, K.-L., Rudolph, W.: Determination of uranium in urine samples of fuel element fabrication workers by beta-delayed neutron counting; Nucl. Instrum. Methods Phys. Res. 223 (1984) 544. Hallstadius, L.: A method for the electrodeposition of actinides; Nucl. Instrum. Methods Phys. Res. 223 (1984) 266. Holm, E.: Review of alpha-particle spectrometric measurements of actinides; Int. J. Appl. Radiat. Isot. 35 (4) (1984) 285. Gladney, E.S., Moss, W.D., Gautier, M.A., Bell, M.G.: Determination of U in urine: Comparison of ICP-mass spectrometry and delayed neutron assay; Health Phys. 57 (1) (1989) 171. International Commission on Radiological Protection: Basic anatomical and physiological data for the use in radiological protection: Reference values; ICRP Publication 89, Oxford: Pergamon Press, 2002. Riedel, W., Beyer, D., Dalheimer, A., Doerfel, H., Henrichs, H., Scheler, R.: Inkorporationsüberwachung auf Thorium; Reihe: Fortschritte im Strahlenschutz, FS-93-69AKI, 1993. Huber, G., Lenz, S., Pfeiffer, B., Kratz, K.-L.: Bestimmung von Uran und Thorium mittels Neutronenaktivierungsanalyse; Universität Mainz, Bericht IKMZ 94-5, 1994. Henshaw, D.L., Allen, J.E., Keitch, P.A., Salmon, P.L., Oyedepo, C.: The microdistribution of 210Po with respect to bone surfaces in adults, children and fetal tissues at natural exposure levels; in: Health effects of internally deposited radionuclides: Emphasis on radium and thorium, van Kaick, G., Karaoglou, A., Kellerer, A.M. (eds.) EUR 15877 EN, Singapore: World Scientific Publishing, 1995, p. 23-26. Harduin, J.C., Peleau, B., Levasseur, D.: Analytical determination of actinides in biological samples; Radioprotection 31 (2) (1996) 229. Environmental Measurements Laboratory: The procedure manual of the Environmental Measurements Laboratory; HASL-300, volume I + II, 28th edition (1997); and as CD-ROM, www.eml.doe.gov (2002). Roth, P., Werner, E., Wendler, I., Schramel, P.: Variation of natural 232Th excretion in nonexposed persons; J. Radioanal. Nucl. Chem. 226 (1-2) (1997) 285. Beyer, D., Dalheimer, A., Riedel, W., Neudert, N.: Die Bedeutung der natürlichen Ausscheidung bei der Inkorporationsüberwachung auf U-238, Th-232, Sr-90 und Ra-226; in: Radioaktivität in Mensch und Umwelt, Volume I, Winter, M., Henrichs, K., Doerfel, H. (eds.), Verlag TÜV Rheinland GmbH, 1998, p. 210-215. International Organization for Standardization: General criteria for the operation of various types of bodies performing inspection; ISO/IEC 17020, 1998, (Genève 1998). Krec, T., Neudert, N.: Eignung eines ICP-MS-Meßsystems für die Inkorporationsüberwachung auf Actiniden durch Ausscheidungsanalyse; in: Radioaktivität in Mensch und Umwelt, Volume I, Winter, M., Henrichs, K., Doerfel, H. (eds.), Verlag TÜV Rheinland GmbH, 1998, p. 51-55. Wihlidal, H., Sinojmeri, M., Lovranich, E., Steger, F.: Sequence analysis of actinides and Sr-90 in urine samples; in: Radioaktivität in Mensch und Umwelt, Volume I, Winter, M., Henrichs, K., Doerfel, H. (eds.), Verlag TÜV Rheinland GmbH, 1998, p. 222-226. International Atomic Energy Agency: Assessment of occupational exposure due to intakes of radionuclides; IAEA Safety Standards Series No. RS-G-1.2, Vienna, 1999.
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10-98 99Kau 99Neu 00Eis 00IAE 00ISO1 00ISO2 00Rob 01ISO 02Sch 03Bou 03IAE 03Nos
10 Measuring techniques Kaul, A.: Radiation dose assessment: Internal dose; Lecture notes, IAEA Regional Training Course in Radiation Protection, Johannesburg, S.A., 1999. Neudert, N., Roth, W.: Improvement of incorporation monitoring during decommissioning of the hot-cell facility at Karlstein; Proceedings of the 3rd European ALARA Network Workshop on Managing Internal Exposures, 15.-18.11.1999, Neuherberg, 1999. Eisenmenger, A.: Messung von Rn-220-Exhalation zur Ermittlung von strahlenschutzrelevanten Inkorporationen bei Beschäftigten der Thorium-verarbeitenden Industrie und Thorotrast-Patienten; Thesis, Freie Universität Berlin, 2000. International Atomic Energy Agency: Indirect methods for assessing intakes of radionuclides causing occupational exposure; IAEA Safety Reports Series No. 18, Vienna, 2000. International Organization for Standardization: Quality management systems - Requirements; EN ISO 9001: 2000, CEN European Committee for Standardization, Brüssel, 2000. International Organization for Standardization: General requirements for the competence of testing and calibration laboratories; EN ISO/IEC 17025: 2000, CEN European Committee for Standardization, Brüssel, 2000. Robredo, L.M., Navarro, T., Sierra, I.: Indirect monitoring of internal exposure in the decommissioning of a nuclear power plant in Spain; Appl. Radiat. Isot. 53 (2000) 345. International Organization for Standardization: Radiation protection - Performance criteria for radiobioassay -, part 1: General principles; ISO 12790-1, Genève, 2001. Schramel, P.: Determination of 235U and 238U in urine samples using sector field inductively coupled plasma mass spectrometry; J. Chromatogr. B 778 (2002) 275. Bouvier-Capely, C., Baglan, N., Montègue, A., Cossonet, C.: Validation of uranium determination in urine by ICP-MS; Health Phys. 85 (2) (2003) 216. International Atomic Energy Agency: Quality management systems for technical services in radiation safety; IAEA Working material, Vienna, 2003. Noßke, D., Dalheimer, A., Dettmann, K., Frasch, G., Hartmann, M., Karcher, K., König, K., Scheler, R., Strauch, H.: Retentions- und Ausscheidungsdaten sowie Dosiskoeffizienten für die Inkorporationsüberwachung; BfS-Bericht BfS-SG-02/03, 2003.
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11 Exposures from natural radiation sources
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11 Exposures from natural and man-made radiation sources
Natural background form the baseline upon which all man-made exposures are added and against which these exposures may be compared. Main contributors to the natural radiation exposure of man are radiations of cosmic origin or radionuclides present in the earth’s crust including the human body itself. Largely depending on altitudes above sea level, geomagnetic altitudes and geological external conditions worldwide and internal doses from natural radiological sources vary from 1 to 10 mSv annually with an average of 2.4 mSv a−1. Nearly 50 % is from indoor inhalation of 222Rn and its progeny.
11.1 Introduction The exposure of man to ionizing radiation from natural sources is a continuing and inescapable feature of live on earth. For most induviduals, this natural background exposures are much more significant than the exposures caused by man-made sources. Exceptions that apply to certain individuals are some exposures caused by medical radiation procedures, through mishandling of radiation sources, in accidents allowing radionuclides to be released to the environment, and at some workplaces. In all cases, the natural background source form the baseline upon which all man-made exposures are added, and it is a common level against which these exposures may be compared. Essentially there are two main contributors to the natural radiation exposure of man: charged and uncharged particles generated by high-engery particles of cosmic origin incident on the earth's atmosphere, and radioactive nuclides originating either by interaction of cosmic-ray particles in the earth's atmosphere or being naturally present in the earth's crust everywhere in the environment, including the human body itself. From these sources dose to man arise from both external and internal exposure. Exposure from extra-terrestrial sources (cosmic radiations and cosmogenic radionuclides) and of terrestial origin (e.g. 40K, and radionuclides of the uranium and thorium decay chains) contribute to the natural background at a comparatively constant level, although largely depending on geological conditions, altitudes above sea level, and geomagnetic latitudes. Origin and kinds of galactic and solar cosmic radiations and reference doses to members of the world population arising from these external sources as well as internally from ingestion of cosmogenic radionuclides are dealt with in Section 11.2. Exposure pathways and doses from terrestrial sources of natural origin are described in Section 11.3 with special emphasis to the inhalation of radon progeny of the uranium and thorium decay chains. In Section 11.4 exposures are summarized due to modification of natural sources by human activities, e.g. release of natural radionuclides to the environment in mineral processing and fossil fuel combustion. Finally, i.e. in Section 11.5, the worldwide average exposures from the single sources of natural origin and the total dose by adding up the various components of the effective dose are given together with the normal ranges of these exposures. These numbers serve as the basis for evaluating the present (year 2000) doses from man-made sources of artificial origin and of occupational radiation exposures.
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[Ref. p. 11-12
Quantitative information on sources and doses from natural and enhanced natural exposures due to industrial activities as well as from exposures to the public from man-made sources of radiation such as medical and occupational exposures, and those from peaceful and military defence uses of nuclear energy is based on data recently published by the United Nations Scientific Committee on the Effects of Atomic Radiation UNSCEAR in its 2000 Report [00U].
11.2 Exposures by cosmic radiation and cosmogenic radionuclides 11.2.1 Origin and kinds of cosmic radiation Galactic cosmic rays incident on the top of the atmosphere consist of a nucleonic component (98 %), and electrons (2 %). The nucleonic component is primarily protons (88 %) and α-particles (11 %), with the remainder heavier nuclei. These primary cosmic particles have an energy spectrum that extends from 108 eV to over 1020 eV. Another component of cosmic rays is generated near the surface of the sun by magnetic disturbances. These solar particle events are comprised mostly of protons of energies generally below 108 eV and only rarely above 1010 eV. They can produce significant dose rates at high altitudes, but only the most energetic affect dose rates of background level. Solar particle events are highly variable in intensity and of short duration, typically a few hours. They have a negligible impact on long-term doses to the general population. The most significant long-term solar effect is the 11-year cycle in solar activity, which generates a corresponding cycle in total cosmic radiation intensity. The magnetic field of the earth partly reduces the intensity of cosmic radiation reaching the top of atmosphere. The form of the earth's field is such that only particles of higher energies can penetrate at lower geomagnetic latitudes. The high-energy particles incident on the atmosphere interact with atoms and molecules in the air and generate a complex set of secondary charged and uncharged particles, including protons, neutrons, pions and low-Z nuclei. The secondary nucleons in turn generate a cascade of more nucleons in the atmosphere. Because of their longer mean free path, neutrons dominate the nucleonic component at lower altitudes. The neutron energy distribution peaks between 50 and 500 MeV as well as around 1 MeV (produced by nuclear deexcitation) are important in dose assessment. The pions generated in nuclear interactions are the main source of the other components of the cosmic radiation field of the atmosphere. The neutral pions decay into high-energy photons, which produce highenergy electrons, which in turn produce photons (photon/electron cascade). Electrons and positrons dominate the charged particle fluence rate at middle altitudes. The charged pions decay into muons, the dominant component of the charged-particle flux at ground level.
11.2.2 Exposures by cosmic radiations At ground level, the muon component is the most important contributor to dose. At air craft altitudes, neutrons, electrons, positrons, and photons are the most significant components. At higher altitudes, the heavy nuclei components must by considered, too.
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11.2.2.1 World population external exposures at ground level At ground level, the dominant component of the cosmic-ray field is muons with energies mostly between 1 and 20 GeV. These contribute about 80 % of the absorbed dose rate in free air from the directly ionizing radiation. The remainder dose comes from electrons produced by the muons or present in the electromagnetic cascade. As altitude increases, the electrons become more important contributors to the dose rate. According to the UNSCEAR 1988 Report [88U] the world population dose rate from directly ionizing radiation is 31 nGy h−1 at sea level. The dose rate is to be considered as averages over the 11-year solar cycle, weighted by the fraction of 54 % of world population living in the northern and southern hemisphere at latitudes below 30°. Since mostly muons are involved, a radiation weighting factor of unity is appropriate yielding the same values for the effective dose rate, i.e. 31 nSv h−1. The world average effective dose rate at sea level from neutrons of isotropic incidence obtained by applying a neutron fluence energy distribution equally to 720 nSv h−1 per neutron cm−2 s−1 is 5.5 nSv h−1. For both the ionizing and neutron components, there is a substantial altitude effect (see Fig. 11.1 [93U] and Fig. 11.2 [93U]). For the directly ionizing and photon component the population-weighted average dose rate is 1.25 times that at sea level, for neutrons 2.5 times. Consequently the world effective dose rate from exposures outdoors is 39 nSv h−1 for the directly ionizing and photon component, and 14 nSv h−1 for the neutron component. Assuming a shielding effect of buildings of 20 % (shielding factor 0.8) and an indoor occupancy of 80 % of time for both cosmic radiation charged particles, photons and neutrons the world average effective dose rate from the directly ionizing and photon component of cosmic rays is about 25 nSv h−1 or 219 µSv a−1, the corresponding average values for the neutron component are 9 nSv h−1 or 78 µSv a−1. The total world average external annual effective dose is thus 297 µSv a−1. 10 4
2.0 La Paz
Equivalent dose rate [mSv /a]
Absorbed dose rate [nGy h− 1 ]
Solar minimum Solar maximum
10 3
10 2 Ionizing component
10
1.5
0.5
Neutron component
1 10-1
1
10
Quito
1.0
Mexico City Munich New York
10 2
Altitude [km] Fig. 11.1. Absorbed dose rate in air at 50° geomagnetic latitude from the ionizing and neutron components of cosmic rays as a function of altitude; [93U].
0
1 2 3 Hight above sealevel [10 3 m]
4
Fig. 11.2. Equivalent dose rate of selected populations at various hights above sea level; data taken from [93U].
11.2.2.2 Exposures by cosmic radiations at aircraft altitudes Aircraft passengers and crew are subject to cosmic radiation exposure rates partly much higher than the rates at ground level, depending on the particular path taken through the atmosphere in terms of altitude above sea level (see Fig. 11.1) and geomagnetic latitude (see. Fig. 11.3 [00U]).
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12
Effective dose rate [nSv h −1]
10 8 6 4
Measurements Birattari et al. Nakamura et al. Fit to measurements
2
0
10
20
30 40 50 60 70 Geomagnetic latitude [deg]
80
100
Fig. 11.3. Geomagnetic latitude variation in effective dose rate from cosmic ray neutrons at sea level; data taken from [00U].
For altitudes of 9 - 12 km (commercial subsonic aircraft) at temperate latitudes, the effective dose rates are in the range of 5 - 8 µSv h−1, such that for a transatlantic flight of 6 hours from Europe to North America the route dose would be 30 - 50 µSv. At equatorial latitudes, the dose rates are lower and in the range of 2 - 4 µSv h−1, such that for a 10 hours flight from Europe to South Africa the route dose would be 20 - 40 µSv. For crew members the average annual flight duration, i.e. the time between leaving the terminal before take-off and returning after landing multiplied by the annual number of flights, is assumed to be 500 hours (300 - 900 h). For occasional flyers an average of 10 hours (3 - 50 h) and for frequent flyers (business flyers or couriers) of 100 hours (50 - 1 200 h) is assumed annually. A small portion of passengers and flight crews travel at higher altitudes of about 18 km on supersonic transports. Effective dose rates of 10 - 20 µSv h−1 are normally found with possible significant dose contributions from solar particle events. 11.2.2.3 Internal exposures by cosmogenic radionuclides The interactions of cosmic-ray particles in the atmosphere by high energy spallation interactions produce a number of radionuclides, including 3H, 7Be, 14C and 22Na at a global inventory of 1275, 413, 12750, and 0.44 PBq, respectively. Only for these elements, which are of metabolic importance in the human body, doses are worth mentioning. UNSCEAR [93U] previously assessed the annual effective doses from these cosmogenic radionuclides to be 12 µSv (14C), 0.15 µSv (22Na), 0.03 µSv (7Be), and 0.01 µSv (3H), respectively, or in total about 10 µSv.
11.3 Terrestrial radiation Naturally occuring radionuclides of terrestrial origin − primordial radionuclides − are present in various degrees in all media of the environment, including the human body itself. Only those radionuclides with half-lives comparable to the age of the earth, and their decay products, exist in significant quantities in these materials. Irradiation of the human body from external sources is mainly by γ-radiation from radionuclides in the 238 U and 232Th series and from 40K. These radionuclides are also present in the human body from ingestion and inhalation, and irradiate the various organs with α- and β-particles, as well as γ-rays. Some other terrestrial radionuclides, including those of the 235U series, 87Rb, 138La, 147Sm, and 176Lu, exist in nature but at such low levels that their contributions to the dose in humans are small.
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11.3.1 External exposures 11.3.1.1 Outdoors External exposure outdoors arise from terrestrial radionuclides present at trace levels in all soils depending on the types of rock from which the soils originate. Gamma-spectrometric measurements indicate that the three components of the external radiation field, i.e. from the γ-emitting radionuclides in the 238U and 232Th series and 40K, make approximately equal contributions to the externally incident γ-radiation dose to individuals in typical situations both outdoors and indoors. The activity concentration of 40K in soils is an order of magnitude higher than that of 238U or 232Th series. UNSCEAR [00U] suggested median values for 40K, 238U, and 232Th series of 400, 35, and 30 Bq kg−1, and population-weighted values of 420, 33, and 45 Bq kg−1. Based on corresponding dose coefficients of 0.0417, 0.462, and 0.604 nGy h−1 per Bq kg−1 the total median and population weighted absorbed dose rates in air were calculated to be 51 and 60 nGy h−1, respectively. The lowest values of the absorbed dose rate in air outdoors are in Cyprus, Iceland, Egypt, the Netherlands, Brunei, and the United Kingdom, all less than 40 nGy h−1, and the higher values are in Australia, Malaysia, and Portugal, all greater than 80 nGy h−1. In addition to variations from place to place, the ambient background γ-dose rate in air at any specific location is not constant in time. It is subject to considerable fluctuation, in particular from the removal of 222 Rn progeny in air by rainfall, soil moisture and snow cover. There are small areas of markedly high absorbed dose rates in air throughout the world that are associated with thorium-bearing and uranium-bearing minerals in soil, such as monazite sand deposits of high levels of thorium as in Guarapari in Brazil (beaches, population size: 73 000; from 90 - 90 000 nGy h−1), Yangiang in China (population size: 80 000; <370 nGy h−1 average), the states of Kerala and Madras in India (costal areas, population size: 100 000; 200 - 4 000 nGy h−1), and Ramsar and Mahallat in Iran (population size: 2 000; 70 - 17 000 nGy h−1), in the latter areas caused by 226Ra deposited from waters flowing from hot springs [00U]. In summary the population weighted absorbed dose rate in air outdoors from terrestrial γ-radiation is 60 nGy h−1. 11.3.1.2 Indoors Indoor exposure to γ-rays is mainly determined by the materials of construction and their surrounding configuration indoors, and inherently greater than outdoor exposure if earth materials have been used. When the duration of occupancy is taken into account, indoor exposure becomes even more significant. From surveys of absorbed dose rates in air inside dwellings the population-weighted average of the absorbed dose rate proved to be 84 nGy h−1 with national averages ranging from 20 - 200 nGy h−1 [00U]. The lowest values are in New Zealand, Iceland and the United States, all below 40 nGy h−1, which probably reflects the preponderance of wood-frame houses. The higher values (95 - 115 nGy h−1) are in Hungary, Malaysia, China, Albania, Portugal, Australia, Italy, Spain, Sweden, and Iran, which must reflect wide use of stone or masonry materials in buildings. 11.3.1.3 Effective dose from external exposures To estimate the annual effective doses, account must be taken of the conversion coefficient from absorbed dose in air to effective dose, and the outdoor and indoor occupancy factors. The averages of the numerical values of these parameters vary with the age of the population and the climate at the location considered. In the UNSCEAR 1993 Report [93U], the Committee used 0.7 Sv Gy−1 for the conversion coefficient from absorbed dose in air to the effective dose received by adults, and 0.8 for the indoor occupancy factor, i.e. the fraction of time spent indoors and outdoors is 0.8 and 0.2, respectively.
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From the data summarized in Sections 11.3.1.1 and 11.3.1.2 (outdoor population weighted absorbed dose rate in air: 60 nGy h−1; indoors: 84 nGy h−1) the worldwide average of the effective dose rate is 55 nGy h−1, and of the annual effective dose 486 µSv (outdoors: 74 µSv; indoors: 412 µSv). The latter is for induvidual countries generally within the 300 - 600 µSv range. For children and infants the values are about 10 % and 30 % higher depending on the conversion coefficient from absorbed dose in air to effective dose.
11.3.2 Internal exposures Internal exposures arise from the intake of terrestrial radionuclides by inhalation and ingestion. Doses by inhalation result from the presence in air of dust particles containing radionuclides of the 238U and 232Th decay chains. The dominant component of inhalation exposure is the short-lived decay products of radon, which is considered separately in Section 11.3.2.2. Doses by ingestion are mainly due to 40K and to 238U and 232Th series radionuclides present in foods and drinking water. The dose rate from 40K can be determined directly from external measurements in vivo of its concentration in the human body. The dose rate from uranium- and thorium-series radionuclides in the body is estimated either from measured activity concentrations after chemical analyses of tissues or from results of analyses of radionuclide contents of foods and drinking water, along with bioassay data and the knowledge of the metabolic and biokinetic behaviour of the radionuclides. 11.3.2.1 Radionuclides other than radon Intake by inhalation of natural radionuclides other than radon and its short-lived decay products makes only a minor contribution to internal exposure. They cover long-lived radon decay products due to disintegration of 222Rn in air and radionuclides of the 238U and 232Th series present in air because of resuspended soil particles [00U]: 210Pb (500 µBq m−3), 210Po (50 µBq m−3), 238U (1 µBq m−3), 235U (0.05 µBq m−3), 232Th and 230Th (each 0.5 µBq m−3), 228Th (1 µBq m−3), 228Ra and 226Ra (each 1 µBq m−3). That means the highest concentration to be for 210Pb, and those for the others to be lower by factors of 10, 500, 1 000, and 10 000. Ingestion intake of natural radionuclides depends on the consumption rates of food and drinking water and on the radionuclide concentrations. Based on reference consumption profiles, reference water balance information and reference values for concentrations of uranium- and thorium-series radionuclides in foods and drinking water UNSCEAR [00U] derived the follwing reference values for annual intakes: 210 Po (58 Bq), 210Pb (30 Bq), 228Ra (15 Bq), 226Ra (22 Bq), 238U (5.7 Bq), 230Th (3.0 Bq), 228Th (3.0 Bq), 232 Th (1.7 Bq), 235U (0.2 Bq). The age-weighted (age distribution: infants 0.05, children 0.3, adults 0.65) committed annual effective dose from inhalation of uranium- and thorium-series radionuclides in air is 5.8 µSv, that from ingestion is roughly 25 times higher, i.e. 140 µSv. The annual committed effective dose from the reference values of uranium- and thorium-series radionuclides in tissues evaluated in the UNSCEAR Report of 1988 [88U] and adjusted to revised tissue weighting factors [93U] is 130 µSv in close agreement with the estimate of 110 µSv derived for adults from the dietary consumption by adults of reference concentrations of radionuclides in foods and water. Potassium is more or less uniformly distributed in the human body following intake in foods, and its concentration in the body is under homeostatic control. The annual equivalent dose in tissues from 40K and hence the annual effective dose is 165 and 185 µSv for adults and children, respectively, i.e. 170 µSv weighted for age.
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The total committed annual effective dose from inhalation and ingestion of terrestrial radionuclides weighted for age is 316 µSv of which 170 µSv is from 40K and 146 µSv is from the long-lived radionuclides in the uranium and thorium series. 11.3.2.2 Radon and decay products 11.3.2.2.1 Sources, health risks and exposure-to-dose conversion Inhalation of radon and its short-lived decay products in the atmosphere are the most important contributors to human exposure from natural sources. 222Rn and 220Rn are the gaseous radioactive products of the decay of the radium isotopes 226Ra and 224Ra, which are present in all terrestrial materials. Some of the atoms of these radon isotopes are released from the solid matrix of the material by recoil when radium decays, and escape from the mineral grain into the pore space. Radon atoms entering the pore space are then transported by diffusion and advection through this space until they in turn decay or are released into the atmosphere outdoors or indoors, i.e. into buildings (see Fig. 11.4 and 11.5 [01B]). α Long-lived decay products
β
214 84 Po
−
214 83 Bi
β
−
218 84 Po
214 82 Pb
α
α 222 86 Rn
Diffusion
Convection Diffusion
Aerosols
Rain trops
Sedimentation Convection
Precipitation
Surface of the earth Earth cracks 238 92 U
α 234 90 Th
β
−
234 91 Pa
β
−
234 92 U
α
α
α 230 90 Th
226 88 Ra
15% 222 86 Rn
85%
Decay products in geological formations
Fig. 11.4. Formation of 222Rn by decay of 226Ra in uranium-bearing minerals, release of radon atoms from mineral grain into pore space and partial transportation by diffusion into the atmosphere outdoors or indoors; redrawn from [01B].
Inhalation of the short-lived decay products of 222Rn, and to a lesser extent of the decay products of Rn (thoron), and their subsequent deposition along the walls of the various airways of the bronchial tree provide the main pathway for radiation exposure of the lungs, predominantly by α-particles. From miners studies it is known that the α-particle irradiation of the secretory and basal cells of the upper airways is responsible for the lung cancer risk. Thus, the damage to these critical target cells of the respiratory tract depends in a sensitive manner on the source/target geometry. The dose that is relevant to the amount of lung-cancer risk depends critically on those environmental factors that affect the probability of the radon decay products to be deposited near the critical target cells after inhalation, i.e. the fraction of radon decay products which is attached to aerosols, the size distribution of the aerosols, and the unattached fraction of radon decay products, as well as the radon activity concentration, the equilibrium between radon and its decay products, the subject’s inhalation rate, and the time of exposure. 220
11-8
11 Exposures from natural radiation sources
e.g.15 Bq /m 3
[Ref. p. 11-12
e.g.50 Bq /m 3
e.g.120 Bq /m 3
e.g.2000 Bq /m
3
e.g.300 Bq /m 3
Fig. 11.5. Indoor diffusion of Rn and its distribution in a building; redrawn from [01B]. 222
Absorbed doses to the critical cells and effective doses are determined by applying exposure-to-dose conversion factors. For 222Rn the range of these values derived from epidemiological studies and physical dosimetry varies from 6 to 15 nSv (Bq h m−3)−1. As in its 1993 Report [93U] UNSCEAR applies in its 2000 Report [00U] the dose conversion factor of 9 nSv (Bq h m−3)−1. Since there are no epidemiological data for lung cancer risk following 220Rn exposure from which to derive a conversion convention for thoron decay products the value of 40 nSv (Bq h m−3)−1 derived from the ICRP Human Respiratory Tract Model [94I] (see Chapter 7) was used for the estimation of equilibrium equivalent thoron doses for indoor and outdoor exposures. 11.3.2.2.2 Air concentrations outdoors and indoors Recent results of radon measurements outdoors tend to confirm the estimates of typical outdoor 222Rn and 220 Rn concentrations made in the UNSCEAR 1993 Report [93U] of 10 Bq m−3 for each radon isotope (range: for 222Rn 1 to >100 Bq m−3; for 220Rn much smaller due to the short half-live). The equilibrium factor, defined as the ratio of the actual potential α-energy concentration to that, if all the decay products in each series were in equilibrium with the parent radon, is suggested for radon in the outdoor environment to be 0.6 for 222Rn, and 0.01 for 220Rn, respectively [00U]. For radon indoor concentrations the corresponding representative data are [00U]: 222 Rn activity concentration 40 Bq m−3, equilibrium factor 0.4; 220 Rn activity concentration, 10 Bq m−3, equilibrium factor 0.03. 11.3.2.2.3 Effective doses For the above worldwide arithmetic outdoor and indoor radon gas concentrations, representative equilibrium factors for the actual potential α-energy concentration for outdoor and indoor occupancy factors of 0.2 and 0.8 corresponding to annually 1760 and 7000 h, and the dose conversion factors of 9 and 40 nSv (Bq h m−3)−1 for 222Rn and 220Rn, respectively, the following annual effective doses are derived [00U]: 222
Rn
Outdoors: Indoors:
10 Bq m−3 × 0.6 × 1 760 h × 9 nSv (Bq h m−3)−1 = 95 µSv 40 Bq m−3 × 0.4 × 7 000 h × 9 nSv (Bq h m−3)−1 = 1 000 µSv
Ref. p. 11-12] 220
11 Exposures from natural radiation sources
11-9
Rn
Outdoors: Indoors:
10 Bq m−3 × 0.01 × 1 760 h × 40 nSv (Bq h m−3)−1 = 7 µSv 10 Bq m−3 × 0.03 × 7 000 h × 40 nSv (Bq h m−3)−1 = 84 µSv
For completeness, the contribution to the annual effective dose from dissolution of the radon gases in blood with distribution throughout the body is [00U]: 222
Rn
Outdoors: Indoors: 220
3 µSv 48 µSv
Rn
Outdoors: Indoors:
2 µSv 8 µSv
The total global annual average of the effective dose from inhalation of 222Rn and its decay products present in air, from dissolution of radon gas in blood and ingestion of radon gas with tap water (2 µSv) [93U] is 1148 µSv with fractions of about 95 % from inhalation outdoors and indoors and 5 % from dissolved radon gas in blood and from ingestion of tap water. The annual effective dose from 220Rn is 101 µSv with fractions of about 90 % from inhalation and 10 % from thoron dissolution in blood. These estimates of the global averages of the annual effective doses for radon only define the normal radon and thoron exposures. One may expect to find many large populations around the world in the order of 106 individuals, whose average exposures differ from the above global averages by a factor of more than 2, and up to a factor of more than 10 for many smaller populations in the order of 104 individuals.
11.4 Enhanced exposures form industrial activities There are numerous circumstances in which materials containing natural radionuclides are recovered, processed and used, causing extra or enhanced population exposures. These exposures are those arising from the mineral processing industries and from fossil fuel combustion by emission of radionuclides by fly ash to air and water, and subsequent eventual intake by humans. Landfills after dredging or wastes disposed on land may also provide pathways of exposure. Main industries are: • • • • •
Phosphate processing Metal ore processing Uranium mining Fossil fuels for electric power production Oil and gas extraction.
Estimated maximum exposures are greatest for phosphoric acid production and for the mineral-sandsprocessing industries. Although effective dose rates of the order of 100 µSv a−1 could be received by a few local residents, levels of the annual per caput effective doses of 1 - 10 µSv would be more common [00U]. These exposure rates constitute a negligible component of the total annual effective doses from all natural sources of radiation.
11-10
11 Exposures from natural radiation sources
[Ref. p. 11-12
11.5 Worldwide average exposure from natural and man-made sources Worldwide average annual exposure by adding the various components described in Sections 11.2 to 11.4 proves to be about 2400 µSv effective dose. The sources of exposure and the single values of the worldwide annual effective doses are summarized in Table 11.1 in µSv and in % of the total effective dose, respectively. It should be stated that this average annual effective dose does not pertain to any one individual, since there are wide distributions of exposures from each source, and the exposures combine in various ways at each location, depending on the specific concentrations of radionuclides in the environment and in the human body, the latitude and altitude of the location and many other factors such as living habits. Table 11.1. Average worldwide exposure to natural radiation sources Source of exposure Worldwide average of the annual effective dose
Cosmic radiation Directly ionizing and photon component Neutron component Cosmogenic radionuclides Total exposure from cosmic and cosmogenic sources External terrestrial radiation Outdoors Indoors Total external terrestrial exposure Internal exposure: Inhalation Uranium and thorium series 222 Rn + daughters 220 Rn + daughters Total inhalation exposure Internal exposure : Ingestion 40 K Uranium and thorium series Total ingestion exposure Internal exposure: from blood 222 Rn, 220Rn + daughters Enhanced exposures: from industrial activities Total (rounded)
[µSv a−1]
[% of total]
219 78 10 307
9.1 3.2 0.4 12.7
300 - 2000
74 412 486
3.1 17.1 20.2
300 - 600
6 1148 101 1255
0.2 47.5 4.2 51.9
200 - 10000
170 140 310
7.1 5.8 12.9
200 - 800
45
1.9
10 - 400 (?)
<10 2400
<0.4 100.0
1 - 10 1000 - 10000
Typical range [µSv a−1]
The normal ranges of exposure to the various components of natural radiation are also indicated in Table 11.1. This accounts for common variations in exposures, but excludes those individuals at extreme ends of the distributions. On this basis, the worldwide average annual exposure to natural radiation sources of 2400 µSv being the present estimate of the central value would generally be expected to be in the range of 1000 - 10000 µSv. About 15 % of the worldwide average exposure is due to cosmic and cosmogenic sources, about 20 % to external terrestrial exposure, in the order of 10 % is from ingestion of natural radionuclides and about half of the total annual effective dose is due to the inhalation of both radon isotopes 222Rn and 220Rn together with their radioactive decay products.
Ref. p. 11-12]
11 Exposures from natural radiation sources
11-11
The present (year 2000) worldwide total annual per caput effective dose from man-made sources is about 410 µSv (400 µSv from diagnostic medical examinations, 5 µSv from atmospheric nuclear weapons testing, 0.2 µSv from nuclear power production, and 2 µSv from the Tschernobyl nuclear reactor accident [00U]), i.e. only about 15 % of the dose from natural sources. The largest contribution to exposures of individuals worldwide is from medical diagnostic procedures (about 98 %), only 0.05 % from nuclear power production. There are a number of occupations, in which workers are exposed to enhanced natural and man-made sources of radiation, i.e. to doses that are directly due to the work. Enhanced natural sources are air travel (crew), mining (other than coal), coal mining, mineral processing, and radon at above ground work places. Man-made sources are the nuclear fuel cycle (including uranium mining), industrial uses of radiation, military defence activities, medical uses of radiation, education, and veterinary. The present (year 2000) average annual effective dose from occupational exposure to enhanced natural sources is 1800 µSv, due to man-made sources is 600 µSv [00U], i.e. about 75 % and 25 % , respectively, of the worldwide average annual per caput effective dose from natural background of 2400 µSv.
11-12
11 Exposures from natural radiation sources
11.6 References 87N 88U 93U 94I 96B 00U 01B
Nakamura, T., Y. Uwamino, T. Ohkubo: Altitude variation of cosmic-ray neutrons; Health Physics 53(5) (1987) 509-517. United Nations. Sources, Effects and Risks of Ionizing Radiation. United Nations Scientific Committee on the Effects of Atomic Radiation, 1988 Report to the General Assembly, with Annexes. United Nations, New York 1988. United Nations. Sources and Effects of Ionizing Radiation. United Nations Scientific Committee on the Effects of Atomic Radiation, 1993 Report to the General Assembly, with Annexes. United Nations, New York 1993. International Commission on Radiological Protection. Human Respiratory Tract Model for Radiological Protection. Annals of the ICRP 24 (1 - 3). ICRP Publication 66. Pergamon Press, Oxford, 1994. Birattari, C., B. Moy, T. Rancati et al.: Neutron measurements at some environmental monitoring stations; Internal Report. CERN, TIS-RP/IR/96-13 (1996). United Nations. Sources and Effects of Ionizing Radiation. United Nations Scientific Committee on the Effects of Atomic Radiation, 2000 Report to the General Assembly, with Annexes. United Nations, New York 2000. Bundesumweltministerium, Bundesamt für Strahlenschutz: Radon-Handbuch Deutschland 2001, BMU/BfS.
