Radioactive Waste Management 2000: Challenges, Solutions and Opportunities (Imeche Event Publications)

Radioactive Waste Management 2000

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IMechE Conference Transactions

Radioactive Waste Management 2000 Challenges, Solutions, and Opportunities

Organized by The Nuclear Energy Committee of the Power Industries Division of the Institution of Mechanical Engineers (IMechE) The American Society of Mechanical Engineers The Japan Society of Mechanical Engineers

Co-sponsored by BNES - British Nuclear Energy Society ICE - The Institution of Civil Engineers IEE - Institution of Electrical Engineers IChemE - Institution of Chemical Engineers

IMechE Conference Transactions 2001-1

Published by Professional Engineering Publishing Limited for The Institution of Mechanical Engineers, Bury St Edmunds and London, UK.

First Published 2001 This publication is copyright under the Berne Convention and the International Copyright Convention. All rights reserved. Apart from any fair dealing for the purpose of private study, research, criticism or review, as permitted under the Copyright, Designs and Patents Act, 1988, no part may be reproduced, stored in a retrieval system, or transmitted in any form or by any means, electronic, electrical, chemical, mechanical, photocopying, recording or otherwise, without the prior permission of the copyright owners. Unlicensed multiple copying of the contents of this publication is illegal. Inquiries should be addressed to: The Publishing Editor, Professional Engineering Publishing Limited, Northgate Avenue, Bury St Edmunds, Suffolk, IP32 6BW, UK. Fax: +44 (0) 1284 705271.

© 2000 The Institution of Mechanical Engineers, unless otherwise stated.

ISSN 1356-1488 ISBN 1 86058 276 1

A CIP catalogue record for this book is available from the British Library.

Printed by The Cromwell Press, Trowbridge, Wiltshire, UK

The Publishers are not responsible for any statement made in this publication. Data, discussion, and conclusions developed by authors are for information only and are not intended for use without independent substantiating investigation on the part of potential users. Opinions expressed are those of the Author and are not necessarily those of the Institution of Mechanical Engineers or its Publishers.

Conference Organizing Committee D Bonser (Chairman) BNFL M Brewin AEA Technology Nuclear Engineering B Bryce Mitsui Babcock Energy I Critchley BNFL

A Goddard Imperial College of Science, Technology, and Medicine S Harnwell SDP Commissioning T Lawrence NNC C Waker Health and Safety Executive

C Ealing ALSTEC

Held 18-19 October 2000, at IMechE Headquarters, London, UK

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Contents Treatment Issues C584/004/2000

C584/006/2000

C584/015/2000

The new Dounreay low-level liquid effluent treatment plant P F G Thomson

3

The disposal of a radioactive cell M Harrison

21

Technical and operational risk management strategies for the Sellafield Drypac Plant (SDP) G McCracken

31

Waste Management Practice C584/026/2000

C584/001/2000

C584/022/2000

C584/025/2000

C584/009/2000

Radiation inheritance of Russian nuclear fleet and ecological safety problems relating to utilization of nuclear submarines and rehabilitation of other facilities in the Navy A P Vasiljev, V A Mazokin, M E Netecha, Yu V Orlov, and V A Shishkin

43

Decontamination and waste minimization techniques in nuclear decommissioning K F Langley and J Williams

47

Transuranic waste management at Los Alamos National Laboratory J J Balkcy and R E Wieneke

57

Disposition of Russian nuclear submarines - outlines of the concept and implementation problems B A Gabaraev, V A Shishkin, and V A Mazokin

67

Management of accumulated operational wastes at BNFL's decommissioning reactor sites A T Ellis, L McTagget, and R I Hey

73

Transport and Storage C584/013/2000

C584/032/2000

Transportation of spent fuel in Japan M Nakajima

85

Engineering considerations associated with plant used for storage of intermediate level waste - a regulator's view W Seddon

95

C584/018/2000

C584/014/2000

C584/016/2000

C584/002/2000

The packaging of waste for safe long-term management S V Barlow and J D Palmer

105

Independent monitoring of solid low-level radioactive waste disposals in the UK S Newstead, N A Leech, and S R Daish

117

Round robin test for the non-destructive assay of 220 litre radioactive waste packages L P M Van Velzen

129

The feasibility of surface storage for high-level waste LCave

141

Environmental and Regulation C584/012/2000

C584/017/2000

C584/020/2000

C584/027/2000

C584/010/2000

C584/019/2000

Authors' Index

Application of in-line monitoring to waste minimization during soil remediation T J Miller

153

Contained water management within the Chernobyl 'shelter object' A A Kornyeyev, C R Wilding, T H Green, and A P Krinitsyn

161

ALARP as applied to high-level waste - the regulatory approach at Sellafield C H Waker

171

Radiation safety problems arising with damaged nuclear submarines utilization V A Mazokin, M E Netecha, YU V Orlov, G A Stanislavski, G A Vasilicv, and V V Borisov

183

Experience in nuclear decommissioning and waste management G R Edler, D Bradbury, and C J Wood

191

Disposal of radioactive waste - a puzzle in four dimensions I J Duncan

201 211

Treatment Issues

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C584/004/2000 The new Dounreay low-level liquid effluent treatment plant P F G THOMSON NNC Limited, Knulsford, UK

SYNOPSIS In February 1997, NNC Ltd were awarded the contract by the UK Atomic Energy Authority (UKAEA) to design, build and carry out the inactive commissioning of a new Low Level Liquid Effluent Treatment Plant (LLLETP) at Dounreay. The purpose of the LLLETP is to collect and treat all the low level liquid effluent from the Dounreay site. The new plant will replace an existing facility which is nearing the end of its operational life. When operational, the LLLETP will enable the UKAEA to meet the more stringent sea discharge requirements specified by the regulatory authorities. This paper considers some of the significant aspects associated with the design and construction of the plant. This includes the key design issues and safety requirements associated with building a facility of this type at Dounreay.

1.

PURPOSE, LOCATION AND KEY FEATURES

The primary purpose of the LLLETP is to collect and treat low level liquid effluent from the Dounreay site. The design ensures that discharges of effluent to sea will be within the pH range of 5 to 9, as required by the Scottish Environmental Protection Agency, (SEPA), compared with the current limits of 2 to 11. In addition, the new plant meets modern standards to reduce radioactive doses to operators to As Low As Reasonably Practicable (ALARP). Prior to awarding the contract to NNC, a number of option studies had been commissioned by UKAEA to determine the best approach to meet the new pH requirement and to minimise the discharge of particulates and solvent to sea.

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The option studies identified a number of key requirements: two stage pH adjustment two large sea discharge tanks with solids settling capacity a sludge removal system a solvent removal capability. To achieve these requirements it was considered necessary to install a new plant which will replace the 40 year old existing facility on site. The new plant is located adjacent to the existing effluent treatment plant. A pictorial view is shown on Figure 1 and the location is shown on Figure 2. The main features of the plant are illustrated on Figure 3, which is a schematic diagram of the plant. Figure 4 shows the building and equipment arrangement. The main features of the LLLETP include: • A below-ground gravity fed receipt tank which collects the low level effluent from the site drains. The tank includes a solvent separation feature to remove solvent from the effluent stream and collect it for disposal. • A buffer tank, which provides sufficient volume to contain 15 hours of average effluent inflow from the gravity receipt tank prior to neutralisation. • Two staged neutralisation for continuous pH adjustment of the effluent stream by the addition of acid or alkali solutions. • Two sea discharge tanks, each able to hold in excess of the expected daily effluent flow to the plant. The tanks are designed to enable solids to precipitate out and for the accumulated sludge to be removed to a collection system. The design life for the plant, support structures and building housing is fifty years. Material selection, design detail and plant layout have addressed the requirement to provide a cost effective lifetime maintenance regime for the facility.

2.

DESIGN SAFETY PRINCIPLES

The design of the LLLETP satisfies the Design Safety Principles applicable to nuclear facilities. These are embodied in the HM Nuclear Installations Inspectorate (NII) Safety Assessment Principles (SAPs). The Safety Assessment Principles are addressed in the design as follows: • A mainly automated plant minimising the operator exposure time and capable of being operated remotely.

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Location of the control room outside the radiologically controlled area. Filtered vessel extract ventilation system to prevent releases to the building atmosphere. Filtered building extract ventilation system to continuously remove any potential airborne activity. A flushing water system to wash-down the inside of the tanks and pipework to reduce activity levels. An installed shield wall between the main tank area and the bulk of equipment requiring maintenance ensures reduced radiological exposure. Provision of uncomplicated equipment and controls. A combination of Hazard Assessment and HAZOP studies throughout the design process. A number of further specific design safety principles as identified in the Preliminary Safety Report (PSR), to assist in the detailed design of the new facility.

3.

WASTE ASSESSMENT

3.1 Incoming Effluent The LLLETP receives effluent from a number of diverse systems. Generally, all effluent is sampled and sentenced by the donor plants before authorisation is sought to discharge to the LLLETP. Each of the donor plants also generally includes buffer storage to enable scheduling of routine discharges. Solids will settle out of the effluent streams in the new plant. The mass of retained solids is expected to be considerably greater than at present as a result of precipitation resulting from pH adjustment. There is also a potential for solvent in the effluent. Solvent arisings are only anticipated to occur during fault conditions. If solvents were to arise, then they can be effectively removed in the gravity receipt tank. 3.2 Outgoing Wastes On average, the LLLETP discharges 450 m3 of liquid effluent daily through the sea discharge pipelines. In addition there are the following waste arisings: • Solids in the form of a sludge for transfer to an on-site sludge processing plant. • Small quantities of solvent to be transferred to an-site disposal facility. • Potentially contaminated ventilation filters • Miscellaneous clothing, wipes etc.

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4.

HAZARD ASSESSMENT

4.1 Introduction The purpose of the hazard assessment was to assess the hazards associated with both the construction and the subsequent operation of the new plant. As the construction involves new plant and the work was carried out in an essentially radiologically clean area, few hazards would have been expected other than conventional industrial hazards. 4.2 During Construction Prior to any construction activities taking place, procedures were developed to ensure that appropriate ground surveys and soil strategies were utilised should the ground on which the LLLETP was to be constructed be found to have contained any small areas of sub-surface contamination. 4.3 Normal Operation The hazards during normal operation could include: Radiation dose to the operator during normal operational tasks. Radiation dose to the operator and maintenance staff during programmed examination, maintenance, inspection and testing (EMIT). Radiation dose to other workers on the site not within the LLLETP building. Radiation dose to members of the public outside the plant. Industrial hazards to the operator or maintenance staff. Environmental effects. The design contains a number of features which ensures that in normal operations the doses are ALARP, these include: Automatic operation so the LLLETP is not normally manned. Routine manual activities take place in a controlled environment. Vessel and building ventilation with specific ventilation in sampling boxes. Health physics surveys to ensure maintenance activities are conducted under controlled and monitored conditions. The location of maintained equipment has been chosen to provide shielding from the main tanks even though the activity levels are low. As the LLLETP is not normally manned, duplicate instruments and controls are located both in the LLLETP control room and in an existing permanently manned control room approximately 700 m from the LLLETP. If a fault were to occur, the operator will determine the cause and the recovery procedure from the remote control room. The operator is thus protected from the initial consequences of a fault. At least one hour is available before operator intervention is essential in the case of the failure of the automatic control system or failure of power supplies. For all other faults several hours are available.

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5.

CONSTRUCTION AND OPERATION OF THE PLANT

5.1 Construction The LLLETP plant is housed in a building 35m x l6 m x l3m high with an annexe 25 m x 6 m x 4 m high. The building is a steel framed construction with part height brick walls and the remainder clad in profiled coated aluminium sheeting. The annexe contains the control room, change areas, switchgear and ventilation fan and filter rooms. The gravity receipt tank is located below the east end of the annexe within a concrete bunded pit. The building is designed to comply with the Scottish Building Regulations and relevant current British Standards. Foundation design is based on ground information obtained from a soil investigation. The building is supported on concrete piles and the foundations are arranged such that excavation and soil removal is minimised. Before construction commenced, the area was swept to check for any underground services and contamination in accordance with site procedures. Excavation was then carried out in pre-defined depths within strips with a radiation walk over taking place at each stage. Where contamination greater than background was found the material was segregated. Material at background or less was set aside and where suitable, was used as fill. Excavation for the receipt tank pit was been carried out within a sheet piled cofferdam following the same procedures. One of the most demanding elements of the construction programme was the fabrication of the Sea Discharge and Buffer Tanks. From the early conceptual stages of the plant it was recognised that it would be preferable to construct the tanks in the completed building. This would avoid chloride contamination of the stainless steel components and minimise disruption of fabrication activities due to the weather conditions at the exposed coastal location had the tanks been constructed outdoors. For construction of the tanks, a prepared access way was formed on the south side of the tank locations by placing fill material on the new slab up to the level of the nominally 300 mm high bunds. Access into the building was through a temporary door in the east gable. The plates from which the tanks were constructed were brought into the building piece-small and the tanks were fabricated in-situ. The stainless steel plates were welded to form strakes. When each strake was complete a series of jacks was used to raise the tank structure to allow another strake to be built beneath the first. This process was repeated until the tank structure was complete. A schematic of this procedure is shown in Figure 5. 5.2 Operations All the liquid effluent will reach the plant via the existing site drain system with the existing pipework extended to the new plant location. All these drain lines have been provided with secondary containment to prevent the possibility of uncontrolled releases to the environment.

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The incoming effluent volumes have been estimated at 150,000 m3 per year at an average flow rate of 20 m3 per hour with a maximum of 50 m3 per hour. The effluent is routed directly to an underground receipt tank with a volume of 35 m3. The tank contains a constant level solvent separator and a partitioned area which acts as a solvent collection tank. The purpose of the solvent separator is to enable the removal and recovery of light solvents from the site effluent. The collected solvent can then be pumped to a terminal point for disposal. The receipt tank is provided with a sludge removal pump and a flushing water system to mobilise the settled solids before they are pumped out. A level control system on the receipt tank starts the variable speed transfer pump as required to maintain a generally constant low level in the tank. This pump transfers the effluent to the 300 m3 buffer tank which has sufficient capacity to accommodate up to 15 hours of average effluent inflow. Because the role of the transfer pump is important in maintaining the availability of the LLLETP, a standby pump is provided. A diesel generator back-up power supply is also provided to each pump in case of loss of supplies. The buffer tank provides capacity to allow for downstream hold-ups and receipt fluctuations. The design aim in normal operation is to operate this tank with a constant low effluent level to maximise the buffer capacity. The tank is provided with sludge removal pumps and flushing water connections to mobilise settled solids if required before they are pumped out. The buffer tank pumps transfer the effluent to the first of two neutralisation tanks located at a high level in the building. The two neutralisation tanks are operated in series as a continuous process, the effluent gravitates from the first tank to the second tank. The pH of the effluent is measured and the control system adjusts the rate of effluent flow and alkali or acid addition to achieve the required pH. An agitator is provided in each tank to ensure thorough mixing of the contents. The alkali or acid is supplied to each tank using the chemical dosing pumps from bulk storage tanks; these are located outside the main building within their own bunded area. The effluent flows from the second neutralisation tank by gravity to the two sea discharge tanks. The effluent is collected in one of the sea discharge tanks. During the transfer to the sea discharge tanks, the effluent is sampled on a continuous basis to confirm that the effluent meets the discharge authorisation requirements. The tank contents are then recirculated using one of the two sea discharge pumps and a recirculation line to ensure homogeneous mixing of the tank contents. After a period, the recirculation of the tank is stopped and sludge allowed to settle to minimise the amount of solids discharged to sea.

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Following settling, and subject to sample analysis results, tidal conditions and the plant manager's authorisation, the key controlled bypass valve is closed, the sea discharge valve opened and the sea discharge pump stalled. During discharge, the effluent is sampled again and the discharge volume is recorded. Each tank is provided with connections to the sludge discharge pumps and water flushing connections to assist in sludge removal as required. In addition to handling effluent from the site drains, the LLLETP is also designed to handle effluent arising from the PFR Sodium Disposal Plant (SDP) located on the Dounreay site. Because of its high chloride content, the SDP effluent forms a separate stream within the LLLETP building. The SDP releases effluent in batches with a total daily release transferred to the 30 m3 SDP hold-up tank located in the LLLETP building. The total daily volume can be accommodated in the hold-up tank. The effluent is sampled at the SDP before transfer to the LLLETP and confirmed to meet the sea discharge requirements, the effluent can then be discharged directly to sea, subject to tidal conditions and the plant manager's authorisation. As with discharge from the sea discharge tanks, samples of effluent are taken prior to discharge. The LLLETP is generally automatically controlled by a Distributed Control System (DCS). Sufficient, independent, hard wired instruments and controls are also provided to ensure the plant can continue to be operated manually if the DCS were to fail. Each vessel is connected to a vessel ventilation system with a HEPA filtered extract to avoid active aerosol discharge to the building. A HEPA filtered building extract and a dedicated sample box extract system are also provided.

6.

DECOMMISSIONING DESIGN

The LLLETP is designed to have an operating life of fifty years. Decommissioning of the plant at the end of life has been considered during the design of the LLLETP. Equipment has been designed to avoid potential contamination traps and be readily decontaminable. This is to be achieved by the use of stainless steel for the majority of the equipment and the use of decontaminable finishes on surfaces. A flushing water system is provided to enable regular removal of sludge and contamination from within tanks and pumps and thus minimise activity build-up. Access is provided to the interior of all tanks to enable regular maintenance and inspection and thorough decontamination prior to dismantling. Tanks and equipment have been designed and located to ensure easy access whenever practical.

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7.

CONCLUSIONS

In the summer of 1999, NNC completed the final inactive commissioning and functional trials of the LLLETP. This has subsequently been followed by the construction of a new dedicated sludge handling facility to manage the solid waste arisings from the plant. Active commissioning trials will now be carried out together with the phased transition from the old facility to the new plant. On completion, the UKAEA will have a state-of-the-art facility with the capability of treating all active liquid effluent from the Dounreay site for the next fifty years.

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Dounreay low level liquid effluent treatment plant Figure 1 Low level liquid effluent treatment plant - pictorial representation

Figure 2

Location of Low Level Liquid Effluent Treatment Plant

Figure 3

Schematic Diagram - Low Level Liquid Effluent Treatment Plant

Figure 4

Low Level Liquid Effluent Treatment Plant - Building and Equipment Layout

Dounreay low level liquid effluent treatment plant Figure 5a Sea discharge and buffer tanks schematic construction sequence - sheet 1

Dounreay low level liquid effluent treatment plant Figure 5b Sea discharge and buffer tanks schematic construction sequence - sheet 2

Dounreay low level liquid effluent treatment plant Figure 5c Sea discharge and buffer tanks schematic construction sequence - sheet 3

Dounreay low level liquid effluent treatment plant Figure 5d Sea discharge and buffer tanks schematic construction sequence - sheet 4

Dounreay low level liquid effluent treatment plant Figure 5e Sea discharge and buffer tanks schematic construction sequence - sheet 5

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C584/006/2000 The disposal of a radioactive cell M HARRISON NUKEM Nuclear Limited, Dorchester, UK

SYNOPSIS Cave 9 in the Active Handling Building on the UKAEA site at Winfrith was an active cell used for experimental work on irradiated fuel during the 1970s and 1980s. During 1998 and 1999 the cave was stripped out, decontaminated and demolished. The challenge was to complete the decontamination of the structure to achieve free release disposal of the concrete and to minimise the volume of low level waste.

1.

CAVE 9 CONSTRUCTION

Cave 9 was the smallest and newest of the three cave lines in the Active Handling Building on the UKAEA site at Winfrith. It was designed and constructed over the period 1972-1974 as an active cell for inspection work on non-fissile items for UKAEA. Later a ventilation system was added to enable a broader range of studies to be carried out for the CEGB on more active fissile and non-fissile items. Figure 1 shows the cave in its operational state. The main cave structure, which was a free standing structure, was based on the availability of five large 40ton shielding blocks that had been displaced from use in the north cave line by an earlier modification. These five shielding blocks were approximately 2.7m wide, 1.5m deep and 3.7m high and were located within the structure, three forming part of the walls and two forming part of the roof. In addition thirteen blocks, each weighing approximately 2.5ton, were used in the construction of the cave supplemented by 36m3 of in-situ reinforced concrete. Figure 2 is a sectional elevation showing the arrangement of four of the 40ton blocks and nine of the 2.5ton blocks, the remaining blocks making up part of the front wall of the cave. The overall dimensions of the cave were 8.3 metres long, 4.8 metres deep and 5.6 metres high with walls 1.5 metres thick. The internal size provided by the construction was approximately

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5.2m in length, 1.8m deep and 4.0m high, and the internal walls were part lined with stainless steel. The operating face was fitted with two zinc bromide windows, one full size and one half size, and four master slave manipulators together with service plugs. Operation of the cave ceased in the early 1990s when plans were developed for its decommissioning and demolition.

2.

CAVE DECOMMISSIONING

Decommissioning of the cave began in 1998, at this time the contamination levels on the walls and on the bench surface were typically in the range 2-6mSv/hr, whilst at floor level contamination up to l00mSv/hr was recorded. The decommissioning of the cave was undertaken in two stages, the first by remote techniques to reduce the radiation levels within the cave, and the second by a series of man-entries. Use was made of the cave equipment, such as the manipulators and waste posting port, to remove contamination from the cave. The internal features of the cave were decontaminated using vacuum cleaners and swabs. Contaminated equipment left over from the last operations in the cave was size reduced in the cave using remote tooling before being posted out for disposal as intermediate or low level waste. This first stage of decommissioning was continued until man-entries to the cave could be justified on ALARP grounds. At this time the radiation levels in the cave were in the range 150 to 300 microsieverts per hour, though on the floor of the cave the levels were several times these figures. The strategy adopted for the second stage of decommissioning involving man-entries was to remove, from one end of the cave, the 40ton shielding block that formed the cave wall. A temporary enclosure was constructed at this end of the cave to allow this process to take place and to act as the controlled entry point into the cave. The enclosure provided a location with relatively low radiation levels from which further decontamination of the cave could be performed using extended tooling. In particular the higher levels of radiation on the floor including some very high point sources could be dealt with whilst incurring acceptably low dose uptake. The majority of the entries were made wearing air fed suits or half-suits as a precaution against the disturbance of contamination during the operations. In all 68 controlled entries were made, involving some 30 staff, with a total whole body dose of 7.24mSv being recorded, the highest individual dose being 0.95mSv. This stage of the process was completed over a four month period. During this phase the cave was stripped of all its equipment including its steel cladding, its windows and manipulators. Loose contamination was removed using HEPA filtered vacuum cleaners and damp swabbing with decontamination reagents. Sprayed on peelable coatings were also used to both temporarily fix the contamination and to remove it with the coating. It was found that care had to be taken with the application of the peelable coatings as they are difficult to remove from pitted surfaces such as bare concrete. A number of high radiation fixed contamination spots were found, these were dealt with by more vigorous mechanical techniques. The decontamination achieved a reduction in internal surface contact dose rates down to an average of less than 20uSv/hr with isolated spots up to l00uSv/hr. The surfaces of the cave were then coated with a water based masonry paint to temporarily fix any remaining loose contamination, achieving the required free breathing conditions within the

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cave. Once the levels in the cave had been satisfactorily reduced the two 40ton shielding blocks making up the majority of the roof were lifted down to ground level. A final decontamination exercise resulted in a cave carcase with low levels of fixed contamination ready for demolition, plus the three 40ton shielding blocks with similarly low fixed contamination levels. The project strategy was that the bulk of the cave structure should be decontaminated to free release levels rather than being despatched as low level waste. However, the project was also required to remove the structure from its location at the earliest opportunity. This drove the decision to demolish the structure in a controlled manner whilst it still had the low levels of fixed contamination, rather than continue with final decontamination in-situ. The demolition was achieved by a combination of dismantling the pre-cast sections (the two remaining 40ton shielding blocks and the thirteen 2.5ton blocks) and cutting the remaining walls using a variety of techniques. Much of the cutting was undertaken using diamond tipped barrel drills and a diamond wire saw, this technique proving to be the most efficient of those tried. Lesser use was made of a diamond tipped road saw and hydraulic bursting. The result was a collection of concrete blocks of between 40 and 2 tonnes in weight, the total weight being approximately 350 tonnes.

3.

DECONTAMINATION FOR FREE RELEASE

The challenge was to complete the decontamination of the concrete blocks to achieve free release of the concrete and to minimise the volume of low level waste. A number of abrasive techniques and equipment were used to remove contamination from all surfaces of each block. Containment tents, fitted with HEPA filtered extract, were erected within the Active Handling Building in which to undertake the decontamination. A shot blasting device supplied by USF Blastrac was effective on the larger plane surfaces of the blocks. This device uses 1mm diameter hardened steel shot that is contained by the blasting head and recovered for recycling. The device incorporates a vacuum system and a HEPA filtered exhaust to recover the debris. Two arrangements of the equipment's blasting head were utilised, the first uses a support beam and winch for blasting vertical surfaces in vertical passes by operation of the winch. The second arrangement is for blasting horizontal surfaces uses an electrically propelled carriage. More localised decontamination was undertaken using hand held scabblers and needle guns.

4.

FREE RELEASE PROTOCOL

The driving force for free release disposal of the cave structure arises from the large volume and weight of the material against the modest and essentially surface nature of the residual contamination. The alternative was to classify it all as low level waste and to transport it to the UK Low Level waste disposal site at Drigg in Cumbria.

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The protocol for defining the free release disposal requirements was established with the involvement of the contractor (NUKEM), the client (UKAEA) and the Environment Agency. The protocol for the disposal of the cave addressed a number of aspects; the measured surface contamination, the presence of surface coatings (paint) as a contamination trap, the activity within the bulk material and, penetrations within the blocks. The measurement of surface contamination was undertaken using standard health physics monitoring equipment by both smear and direct probe. The protocol required 100% surface monitoring of all surfaces by both methods with the limits set at 4Bq/cm2 beta-gamma and 0.4Bq/cm2 alpha, i.e. less than the levels required for removal from a designated area. Large area smears were taken for the detection of loose contamination. If any positive readings were obtained the area in question was subject to a more focused smear survey to find the source of the contamination. For the probe survey a grid was marked on the surface of the block in order to ensure that the whole surface was monitored. Although the contamination was expected to be mostly 137Cs and 60Co, the surveys also included monitoring for alpha activity. The presence of paint on the surfaces was considered as having the potential to mask contamination, the history of the cave and the blocks was uncertain therefore repainting to seal contamination may have occurred in their past. The protocol therefore required the removal of all surface coatings back to the original material surface. It was necessary to demonstrate that the bulk material, i.e. the concrete, was within the regulatory limits for free release disposal. The concern was that the bulk concrete had become radioactive by some means, be that migration into the concrete matrix or activation of the concrete by the high radiation levels that has existed in the cave during its operation. Whilst both of these processes were thought unlikely to have occurred, the protocol required that this be demonstrated. To provide this demonstration the chosen method was to remove core samples from the blocks for radiochemical analysis. Cores were removed by dry diamond drilling using a 50mm diameter coring bit, to a depth of 100mm. The core was then split along its length into four approximately equal 25mm long sections. Each of these sections was identified uniquely and analysed individually. The protocol aimed to limit the number of core samples taken for analysis by recognising the similarity in design and operational history of the blocks making up the structure. For this, the concrete blocks and cave structure were divided into four groups; the five 40ton blocks, the thirteen 2.5ton blocks, the in-situ east wall of the cave and, the in-situ west wall of the cave. The number of cores to be taken was then defined for each group. For the 40ton and 2.5ton blocks the sampling regime required that the first block have cores taken from approximately the centre of each vertical face, plus one from either the base or top face. If these cores proved to be within the free release criteria then the second block only required three of the vertical faces to be cored, though one of these had to be taken from what had been the inside face. If these cores proved to be within the free release criteria then only one core was taken from the inside face of the remaining blocks. The same regime was adopted for the 2.5ton blocks. The sections of the east and west walls of the cave were each sampled once, again at approximately the centre of the face that had been nearest to the inside face of the cave.

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The surface of the blocks at the location where the cores were to be taken was decontaminated and monitored prior to removing the cores. This ensured that contamination was not driven into the block by the drilling operation, and that the cores could be removed from the designated area and sent off the site for analysis.

5.

TREATMENT OF PENETRATIONS

The concrete blocks contained numerous penetrations ranging from small bolt holes just a few centimetres deep to en-cast liners through the full depth of the blocks and up to 300mm diameter. The monitoring of the smaller penetrations by smear or probe was clearly impractical and an alternative method was required. The method selected for the small penetrations was to remove any surfaces that had the potential to be contaminated, by over-drilling. The 'rule of thumb' adopted for this method was to drill out a core of twice the diameter of the penetration and one and a half times its depth. The oversize penetration was taped over to prevent recontamination. For the larger penetrations the original proposal was to decontaminate these using a grit blasting lance and then monitor them using extended reach tooling. However a detailed examination of the penetrations and the block revealed that many of them were constructed of concentric tubes where the inner tube did not extend the full depth of the block. To compound this problem it was also found that the tubes were not sealed together leaving an inaccessible gap between the two. The potential for this gap to be a contamination trap was considered to be too significant to ignore, surface decontamination and monitoring was therefore ruled out for these larger penetrations. In order to free release these blocks with the larger penetrations an alternative method was required. The method chosen was to seal up the penetrations at both ends by welding on metal caps, effectively sealing in any contamination. Once the remaining free release exercise was complete, the block was broken up using a hydraulic breaker and the sealed steel penetrations recovered for disposal as low level waste. In addition to the designed penetrations in the blocks there were also a variety of penetrations or gaps between the concrete mass and any steel features, tubes, plates etc. These small gaps were both potential contamination traps and were inaccessible for monitoring. In these cases the concrete was cut back until it was observed that the gap between the concrete and the steel had closed up. A further 10mm of concrete was then removed to ensure that any potentially contaminated material had been removed. Alternatively the steel plate was cut back to reveal a minimum of 100mm of the concrete face to allow this to be thoroughly monitored.

6.

CORE ANALYSIS

The core sample analysis was undertaken by Southampton Oceanography Centre. The objective was to determine the alpha and beta-gamma activity for each of the concrete samples and compare the results against the free release criteria of 0.4Bq/g. This, together with the surface monitoring, would determine whether the blocks could be free released for disposal or whether they would have to be disposed as low level waste.

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The concrete samples were ground, thoroughly homogenised and a sample was then removed for gamma spectroscopy. A further sample was removed and digested in aqua regia, the resulting leachate was measured for total alpha and beta activity. All anthropogenic radioisotopes identified were reported. In addition, limits of detection were calculated for the isotopes specified by NUKEM, namely 54Mn, 60Co, 137Cs and 241 Am. A typical result from the analysis of one sample was as follows: Sample reference:

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Mn I 60Co I 137Cs I 241Am I Total alpha I Total beta <0.007 | <0.016 | <0.007 | <0.004 | <0.04 | <0.08 No other anthropogenic radioisotopes were detected in the gamma spectrum. All results in Bq/g.

7.

TRITIUM ANALYSIS

The results of the radiochemical analysis of the core samples were presented to UKAEA who then requested that some analysis of Tritium levels be made. Two further cores were removed and the analysis was undertaken on these samples by RCD-RadioCarbon Dating. 50g of the concrete was mixed with 7g of anthracite known to contain no Tritium and combusted in pure oxygen in a high pressure combustion bomb to produce carbon dioxide and water. The water was collected and subject to direct liquid scintillation counting in association with background and standard samples. The two results obtained were 8.0 ± 0.8 Bq/kg and 7.8 ± 0.8 Bq/kg, significantly below levels of regulatory concern.

8.

LOW LEVEL WASTE ARISINGS

During the decontamination operations, any contaminated material removed from the surface of the concrete blocks, or any material that could not be monitored, was packaged as low level waste. No precise calculations were made of the amount of material removed and despatched as low level waste but by observation it is believed to be much less than 5% of the total weight of the cave structure.

9.

COST ANALYSIS

There were two main drivers for the free releasing of the structure of the cave. The first being the volume of the Drigg store that would be taken up by what was believed to be virtually all non-radioactive waste. It has been increasingly recognised over recent years that Drigg is a valuable resource for the nuclear industry. With the UK decommissioning programme just beginning it is clear that this resource needs to be used with care if it is to last.

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The second driver is the disposal cost of the waste at Drigg. At the start of the cave decommissioning project it was clear that the cost of disposal of the complete cave structure would be significant. What was less easy to calculate with any degree of accuracy was the likely cost of the alternative, i.e. the free release route. As previously indicated, Cave 9 was the smallest of the three cave lines within the Active Handling Building at Winfrith. The structure of the remaining two caves together is estimated to be eight times the volume of Cave 9. This early decommissioning of Cave 9 provided an opportunity to both refine decommissioning methods and to calculate decommissioning and disposal costs in order to apply these to the other two cave lines. The costs incurred in order to free release the dismantled structure were as follows: Labour Materials Equipment (depreciation only)

Decontamination

Analysis

Radiation and Contamination Monitoring Radiochemical Analysis Total

£45,000 £ 3,000 £ 6,000

£15,000 £15,000 £84,000

In comparison, the waste disposal charges at Drigg for the estimated 130 cubic metres of waste would have amounted to £208,000 based on a nominal £1,600 per cubic metre, a difference of £124,000. The strategy for the disposal of Cave 9 involved the dismantling of the structure whilst there remained low levels of fixed contamination on the exposed surfaces. From the experience gained in this work it is now believed that this is not the most efficient method. For dry cave structures, such as Cave 9, the more effective method would be to complete the decontamination and sampling operations whilst the structure remains essentially intact. Demolition would then take place as an inactive operation, avoiding the need for the additional handling operations during the decontamination work. A key factor in the time taken to decontaminate the concrete blocks was the detail in their construction. In particular the five 40ton blocks were better described as steel fabrications that had been in-filled with concrete, such was their complexity. It was the features in the fabrications such as the penetrations and the concrete to steel interfaces that added significantly to the time and therefore the cost of decontamination.

10.

SUMMARY AND CONCLUSIONS

The decommissioning, demolition and disposal of the Cave 9 structure was undertaken with the objective of free releasing the bulk of the structure for disposal, this was achieved. This project provided useful background data on all aspects of such decommissioning work. In particular the aspect of free releasing the structure for disposal as non-active waste provided specific knowledge and data. The involvement of the contractor, the client and the regulatory body allowed a pragmatic but justifiable protocol to be developed on which to undertake the

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monitoring, sampling and analysis. Lessons were learned on the most effective methods for the decontamination of such structures to achieve the free release criteria. Most significantly the project provided data on the costs involved in such operations and showed that these are significant even against what are considered to be the high costs for disposal as low level waste at Drigg. Overall, the project demonstrated that the free release disposal of such structures is both practical and cost effective.

Figure 1

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Figure 2. Sectional Elevation showing 40ton and 2.5ton blocks

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C584/015/2000 Technical and operational risk management strategies for the Sellafield Drypac Plant (SDP) G McCRACKEN Waste Management and Decommissioning, British Nuclear Fuels plc, Sellafield, UK

1

ABSTRACT

The development of a process to manage significant volumes of historic Intermediate Level Waste (ILW) has posed a number of challenges. Over a planned operating lifetime of 15 years the Sellafield Drypac Plant (SDP) will be required to process approximately 12,500m3 of waste, producing over 30,000 product drums. This paper describes the procurement, prototype and testing strategies used for the key SDP process elements to reduce the lifetime technical and operational risk. The methods used to validate the process and the benefits for commissioning and lifetime operations are explained as well as the key findings from the work to date.

2

INTRODUCTION

British Nuclear Fuels plc (BNFL) manages and operates facilities associated with all aspects of the Nuclear Fuel Cycle including the reprocessing of spent nuclear fuel at its Sellafield site in the North West of England. The Magnox fuel programme has involved reprocessing several million fuel elements at Sellafield from the late 1950's onwards. Fuel was processed by removing the Magnox Metal Cladding which was then stored under water in a large silo complex. In addition to Magnox cladding a large volume of technological scrap was stored in these silos. ILW pond sludges have also been stored in 1000m3 storage tanks which have arisen from reprocessing operations at Sellafield. These significant volumes of waste represent a historic liability to both BNFL and its customers. There is a requirement to safely discharge these liabilities whilst minimising lifetime costs. The effective management of technical and operational risks is therefore a key part of achieving this requirement. A

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number of major plants have been, and are being constructed to safely and efficiently condition these wastes into a form suitable for long term storage in surface stores and potentially in a future underground waste repository. The Sellafield Drypac Plant (SDP) is a key part of BNFL's. strategy for the treatment of ILW on the Sellafield site.

3

THE DRYPAC PROCESS

The Sellafield Drypac Plant is at the hub of BNFL's ILW Strategy. The plant has been designed to treat a range of ILW materials, including ion-exchange resins and technological scrap. The main feeds, however, are sludges formed as corrosion products form the long term underwater storage of the cladding from spent Magnox Fuel Rods. The plant is designed to significantly reduce repository storage volume for these sludges. Over a planned operating lifetime of 15 years, SDP will be required to process approximately 12,500m3 of waste, producing over 30,000 product drums and represents a significant investment for BNFL and its customers. The plant is currently under construction and is planned to be operational in 2004. A Drypac process schematic is shown in figure 1. Retrieved waste is delivered to SDP in 1.5m3 skips that are transported inside bottom opening flasks. The process consists of screening the waste and separating it into two streams, undersize and oversize.

Figure 1 - SDP process schematic

The undersize material is predominately sludge type material that can fall through a 60mm-mesh screen. All items remaining on the screen following vibration are treated as oversize waste.

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The undersize waste is transferred into sacrificial cans that are lidded and then dried to a predetermined moisture content before being compacted into pucks. The oversize waste, size reduced where necessary, is placed into an oversize sacrificial can which is lidded and then compacted into a single puck. The compacted height and weight of each puck is determined prior to it being transferred to a buffer store. Undersize pucks remain in the buffer store for a minimum period to allow cooling. Pucks for placement in a drum are selected such that the drum height utilisation is maximised without exceeding the drum weight and fissile limits. Undersize and oversize pucks are segregated in drums. The product drum is then transferred via an underground tunnel to the existing Waste Encapsulation Plant (WEP) for the addition of the final grout and drum lidding. The drums are then moved to an above ground store, where they can be safely stored for many years. A cross section of a filled and grouted SDP product drum is shown in figure 2.

4

PROJECT STRATEGY

An early decision was taken on the project to develop a risk reduction strategy as part of the design and procurement programmes. This required capital provisions and programme requirements to be reviewed and fully justified before implementation. BNFL already had considerable experience in developing, building and operating large nuclear facilities and therefore risk reduction measures focused primarily on the new, novel elements of the SDP process. These were: • Waste Separation and Can Filling • Size Reduction & Remote Maintenance • Drying • Compaction Two different strategies were employed for these process areas. An extended works test strategy was used for size reduction, remote maintenance and compaction and a full-scale prototype strategy for waste separation, can filling and drying.

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The Extended Works Testing Strategy involved building in the requirements for extended testing periods, managed by BNFL and its staff at contractors premises into the normal procurement contracts. Project programmes were reviewed to ensure sufficient time was allowed for these activities to be undertaken prior to delivering the equipment to Sellafield. The Prototype Strategy involved building fully engineered prototypes at manufacturers works and in development facilities at Sellafield. This strategy was chosen for those areas where there was considered to be significant technical and operational risk. Both strategies were developed to enable the robustness of the process to be validated at an early stage of the project lifecycle, but provided other benefits such as early validation of commissioning procedures and familiarisation for commissioning and operations teams. The intent was that design risks would be identified at an earlier stage and therefore provide more time to resolve these issues before the equipment was installed on site.

5

EXAMPLE 1 - SEPARATION CAVE PROTOTYPE TESTING

The separation cave is a major part of the SDP process. It is a large cell, with a high density of remotely operated mechanical equipment, a large proportion of which is new and of novel design. Using experienced gained from existing mechanical handling processes at Sellafield, an assessment was undertaken to establish which elements of this design were considered to have a high degree of technical and operational risk. The front end of the Separation process consists of two skip tipping and screening assemblies, 3 Undersize collection vessels and two measuring vessels/can filling machines. The frontend separation prototype was therefore designed to be a fully engineered representation of these major components. A schematic of the prototype is shown in figure 3.

Figure 3 - Separation cave prototype

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The actual test arrangements are shown in figures 4a and b. All the equipment will be operated and maintained remotely once active material has been introduced into the plant.

Figure 4a - Separation cave prototype can filling 6

Figure 4b Separation cave prototype skip tipping

OBJECTIVES

The main objectives of the testing strategy can be summarised as follows: • • • • • • • • •

Prove that the equipment and software can be commissioned, operated and maintained satisfactorily Prove there is adequate visibility and ergonomics. Prove plant throughput and overall equipment effectiveness (OEE). Prove interfaces between individual machines. Assess the potential for excessive build up of active material, contamination and splashing. To optimise dampening and wash sprays. Identify areas requiring design modifications prior to the commencement of equipment manufacture for the actual plant. Aid in the development of housekeeping, maintenance and failure recovery strategies. Verify the process robustness to waste feed variations.

All the above were considered to contribute significantly to the reduction of technical and operational risk before installation of the equipment on the actual plant.

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SUCCESS CRITERIA

The success of the strategy was dependent on two fundamental issues: The objectives were achieved such that there were no accidents or injuries to anyone. The objectives were achieved to the budget cost and programme.

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Key areas of risk would be verified and mitigation measures set in place by re-design, commissioning methodology or operational strategies.

8

PROCUREMENT STRATEGY

The separation front-end equipment was procured from a single contract supplier. The provision of a fully engineered prototype facility at the contractors' works was a key requirement during the tendering process for this equipment. The contract programme was also required to reflect the availability of the use of the prototype and subsequent test period's prior to final designs being agreed for manufacture of the main plant equipment.

9

ORGANISATION

The organisation arrangements for the prototype testing strategy are summarised in figure 5.

Figure 5 - Prototype testing strategy An integrated team approach was necessary to achieve the test programme and involved teams from within and out with BNFL. A project management function was required to ensure all the elements of the work were effectively co-ordinated and that the programme and budgetary elements were controlled. The use of experienced operators and maintenance personnel was a key element of the strategy. People were seconded from existing operational plants to provide practical feedback on all aspects of operating and maintaining the plant.

10

SUPPORTING EQUIPMENT & MATERIALS

In order to simulate the actual plant, wooden assembles were made to represent the actual cave dimensions and window views. Cranes, overhead and through-wall manipulators

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were used, along with camera viewing arrangements. These were used to simulate actual plant operations and particularly to validate remote maintenance tasks. The waste simulants used during the testing period were strictly controlled to ensure consistent quality. Simulants were carefully chosen to represent the full spectrum of likely waste feed scenarios to underpin the overall robustness of the process.

11

TEST FINDINGS

The test programme was developed to cover all the objectives identified. The testing revealed a number of design issues as well as providing an opportunity to optimise parts of the process. Design changes were identified to address the following: • Improve throughput, reliability and repeatability. • Prevent seizing of equipment due to build-up of material. • Prevent excessive splashing/contamination. • Hold up of sludge material within the undersize collection vessels. • Improve the remote replacement of cabling, plugs and sockets. • Improve remote maintenance features such as guides for replacement components, captive fasteners, accessibility for remote handling equipment and lifting features. • Improve equipment visibility using camera systems. • Optimisation of water sprays and water management systems. • Optimisation of remote handling equipment such as manipulators and cranes. The design changes identified as a result of the test programme were fed into the manufacturing programme for the main plant equipment. The manufacturing programme was developed to reflect hold points on certain items of equipment where there was considered to be a high risk that changes were necessary. This required a structured process for releasing equipment for final manufacture. Some of the design changes identified were considered to have remaining risk such as to merit backfitting the new design to the prototype for re-testing before manufacture. This process did have an impact on the manufacturing delivery programme, although this was not significant and did not compromise overall project timescales. The original test programme was successfully carried out to cost and programme. The integrated team approach used was considered to be of benefit, both to the robustness of the plant design and individuals involved. A significant amount of recorded and tacit knowledge was gained during the test programme which has given a much greater understanding of the process and has allowed commissioning documentation to be produced for the main plant as well as draft operating and maintenance instructions. Some remote maintenance tasks are complex and testing allowed key assumptions to be validated at a much earlier stage in the project lifecycle. The maintenance philosophy on a number of key items was changed as a result of the testing. The work provided a development opportunity for the operations personnel and allowed a number of new skills to be acquired, which was considered to be a significant benefit. The overall performance of the prototype was measured by using the Overall Equipment Effectiveness (OEE) technique; where:

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OEE = Availability x Quality x Performance Availability Breakdowns, set up and adjustment losses. Quality Start up and rework losses. Performance Idling, minor stoppage and reduced speed losses. This technique allows the actual performance of the process to be assessed against the theoretical operational research models used as part of the development of the plant design. Assumptions used in the model were validated and action plans put in place to address shortfalls in performance.

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EXAMPLE 2 - COMPACTION EXTENDED TESTING

Two high force compactors are used in the SDP process to reduce the volume of the waste by a factor of typically 3. The compactor test arrangement is shown figure 6. Existing facilities at Sellafield already had similar compaction operations but this was the first requirement to process intermediate level waste.

Figure 6 - Compactor 1 testing arrangement Existing compactors used glovebox techniques to perform maintenance and clean-up operations. The higher radiation levels in SDP meant that the compactor design needed to provide shielding to protect the operator and all items within the active cell needed to be remotely maintainable. Remote housekeeping and cleaning operations were also a concern. Limited compaction experience existed for this type of waste and the effects of compaction at up to 300°C were unquantified. The strategy chosen was to procure the compactors well in advance of the requirement for the overall project. This required significant design effort at the front end of the project programme. A nine month extended test period was included as part of the contract procurement programme. The strategy was to build the first compactor and perform extended testing using experienced

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BNFL commissioning, operations and maintenance personnel. The objectives, success criteria and organisation were similar to those described for the Separation front-end Prototype. In addition, there was a requirement to carry out a large number of compactions using various simulants to verify the robustness and product quality elements of the product wasteform. This included carrying out hot compactions with cans at temperatures up to 300 degrees C. On completion of the extended testing any designs changes necessary were to be carried out to the first compactor and included before final manufacture of the second.

13

TEST FINDINGS

The 9-month extended testing programme for compactor No 1 was completed successfully with over 250 compactions in 7 different campaigns. A total of 34 maintenance demonstrations were completed as well as regular housekeeping trials throughout the period. The knowledge of the compaction process and maintenance requirements gained by the project teams during this period was considered to be of great future benefit to the project. The main findings of the work can be summarised as follows: • The air cleaning system proved ineffective in removing dust from the • mould top requiring a different method to be devised. • Variable compaction force control was considered necessary to allow flexibility for varying waste characteristics. • Build up of simulant material was experienced on some areas of the compactor cell internals requiring remote housekeeping techniques to be developed. • A number of special items of equipment were required in order to undertake some maintenance tasks. • Some waste types containing high levels of plastics produced poor quality pucks, requiring a waste mixing strategy to be considered. • Window and camera viewing systems were found to be good. • Some maintenance procedures needed to be revised to reflect constraints identified. A number of changes were made to Compactor 1 as a result of the testing and these were also incorporated into Compactor 2 design. Both compactors were completed and shipped to the plant ahead of the programme schedule. The testing allowed commissioning preparations and maintenance methodologies to be developed and the compaction data allowed product quality assumptions to be verified. The consistency of the pucks produced by the compaction process was considered to be a significant project risk. The extended testing verified that successful compaction could be achieved and that the sacrificial can design concept was robust.

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CONCLUSIONS

The effective management of technical and operational risk on a major novel project needs to be effectively managed as an integral part of the project implementation strategy.

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Early identification of these key risks allowed a number of strategies to be adopted as part of the procurement process. Sufficient programme time needed to be allowed for testing to take place and funds need to be identified. The strategy adopted does incur increased costs at an early stage of the project. The work to date on SDP has provided significant benefits in reducing risks before plant and equipment has been delivered for installation on site. Many of the technical and operational shortfalls would not have become evident until the main commissioning programme was well advanced if detailed testing had not been undertaken leading to significant delays to the testing programme and much more difficult modifications in situ. This is typically 18-24 months later for SDP. The strategy allowed extra time to address these performance shortfalls. The cost of rectifying the design of the plant once installed on site is very much increased during the commissioning phase. The implications of delays when there is a large commissioning workforce is significant and the mechanisms for undertaking modifications are more complex. In addition to cost savings during commissioning it is anticipated that lifetime operational risks associated with plant performance have been mitigated or reduced as a result of the strategy.

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Waste Management Practice

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C584/026/2000 Keynote Radiation inheritance of Russian nuclear fleet and ecological safety problems relating to utilization of nuclear submarines and rehabilitation of other facilities in the Navy A P VASILJEV, V A MAZOKIN, M E NETECHA, Yu V ORLOV, and V A SHISHKIN SUE RDIPE, Russia

Intensive operation of national nuclear fleet in recent 25 - 30 years resulted in accumulation of considerable amounts of radioactive substances in the areas of the ships and vessels stationing. In the North of European part of Russia nuclear submarines (NS) decommissioned and subject to utilization, as well as their servicing infrastructure facilities, i.e.: coastal technical bases (CTB), radioactive waste and materials submerged to the sea bed, are major sources posing potential radiation and ecological hazards. By early 1999 about 180 NS were taken out of service in the Navy, more than 100 of them being removed from the Northern Fleet. Fuel was unloaded from 50% of the NS. Total radioactivity of reactors in NS taken out of service amounts to -500X106 Ci. Approximately 105 Ci remains in each reactor compartment after the fuel unloading. The coastal technical bases (CTB) of the Navy is another source of radiation and ecological hazard, as spent nuclear fuel (SNF) and radioactive waste (RW) have been accumulated in essential amounts. Ships providing technological services to nuclear fleet, namely: off-shore technical bases (OTB), special tankers for liquid radioactive waste (LRW) storage and floating towed tanks, are also radiation hazardous potential sources of environmental radioactive contamination. Submerged radioactive waste (including reactors with SNF) in the Barents and Kara seas cause anxiety of the environmental protection organizations and public in our country, as in well as some Western countries. It is worth mentioning, however, that currently the range of the submerged waste, places of submersion, radioactivity values are well-known. International Arctic Sea Assessment Project (IASAP) was undertaken to study possible impacts of RW dropped in the seas of the Arctic Ocean on the people and environment, its results suggesting no present or future risks stemming from the RW disposal.

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In the course of NS decommissioning, storage and utilization the ecological problems, which can arise from long-term anchorage of NS in the stationing area with unloaded SNF, and from management of the cut out reactor compartments (RC) containing conventional and damaged nuclear steam supply systems (NSSS), were clearly defined. The essence of ecological problems is potential impacts on the environment brought about by submersion in the coastal water of NS with unloaded SNF, which have not been prepared for long-term storage, and also from emergency situations involving RC in the period of their preparation, transportation and long-term mooring resulting in the contact of RC radioactive equipment and environment (sea water, ground, humid atmosphere on the coast, etc.). The ecological aspects concomitant to decommissioning of CTB and ships providing special technological services pose no less difficult problems. Solid and liquid radioactive waste is stored in CTB in unsatisfactory form violating the requirements. NS decommissioning predetermines the necessity of solving the problems relating to development and implementation of technologies for handling SNF, radioactive equipment of reactor compartments (RC), various types of solid and liquid radioactive waste, which are safe for the population and environment. A concept of NS utilization has been formulated and is being implemented in Russia, the concept envisaging unloading of SNF from NS reactors and removal (cutting out) of the reactor compartment (RC) from the submarine hull. Meanwhile, the NSSS radioactive equipment remains within RC in standard places. It is assumed that the reactor compartments after special preparation (taking a set of technical measures assuring radiation and ecological safety) will be arranged in long-term storage points (LSP) and kept there for -70 years, in the course of the period due to natural decay of radionuclides the radiation levels near the NSSS equipment will reduce essentially and that will permit subsequent dismantling of the equipment actually without any restrictions. Today feasibility studies are undertaken for the optimal choice of variants aimed at rehabilitation of the CTB buildings, structures and territories involving the elaboration of transport and process schemes of SNF and RW evacuation from the bases. The NS, which are currently in stationing areas (even with SNF) and three-compartment blocks, and, at a later time, the cut out reactor compartments, if emergency-free conditions of storage and handling are provided, produce actually no adverse effect on the ecological situation and population. Ecological problems may arise in case of emergency situations involving the NS, three-compartment blocks and single RC, when they are affected by natural or technogenic factors. The impacts mentioned can bring about: • submersion in coastal water of NS unprepared for long-term mooring, their SNF remaining on board; • submersion of the three-compartment blocks caused by loss of their tightness because of corrosion processes or as a result of man-made or natural calamities; • emergency situations involving RC in the period of preparation, transportation and longterm storage resulting in the contact of RC radioactive equipment and environment (sea water, ground, humid atmosphere on the coast, etc.).

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The coastal technical bases nowadays are largely equipped with special structures built in 1960-1965 and outdated (obsolete) physical and radiation monitoring equipment and instrumentation. Major part of the structures are in a state of disrepair. Accordingly, radionuclide release to the environment from the bases occurs periodically with that or any other intensity. In the context of actually taking place and potential ingresses of radioactivity to the environment the following becomes urgent: • study of the amount, composition and chemical forms of radioactive substances, contained in RC equipment (with/without SNF), radioactivity distribution over the structures, decline in the radioactivity in time, dose rate distribution near specific equipment and RC as such; • formulation of strategy aimed at construction of radiation-proof ecological barriers in RC in possible routes of radionuclide migration to the environment; selection, production and testing of protecting materials preventing radioactivity escape from RC; • comprehensive radiation inspection of the NS and three-compartment blocks in the stationing points; of cut out RC and RC prepared for long-term cooling; of the CTB territories, buildings and structures; of the off-shore technical bases under decommissioning; • study of radionuclides migration from reactor compartments to the environment in case of submersion of the NS, three-compartment blocks and RC brought about by the latter loss of tightness because of corrosion processes and in case of emergency situations; • study of radionuclides escape from RC, considering different methods of their long-term storage and possible loss of tightness stemming from RC shell corrosion or occurring as a result of emergencies; • study of radionuclides migration to the environment from leaky SNF and LRW storage places, coastal and off-shore technical bases, as well as from solid radioactive waste (SRW) storage places because of impacts of atmospheric precipitation and ground water; • study of the routes of the radionuclide intake by human organism from the environment and evaluation of radiation-induced risks of injury for people residing in the relevant region. The research trends enumerated are rather expense-consuming and they will necessitate involvement of essential material and research resources. Nonetheless, despite the grave economic situation in Russia, certain steps have been already taken and are being taken towards solution of the problems stated. A concept of NS comprehensive utilization has been elaborated and substantiated, approaches permitting rehabilitation of damaged NS have been developed. Foreign (USA) and certain domestic experience has been analyzed. Experimental and analytical studies have been made permitting acquisition of major portion of the necessary information about the amount of radioactivity contained in RC with and without SNF, radioactivity distribution over the structures, radionuclide composition of radioactivity, decline in the radioactivity in time, dose rate near specific equipment, etc. The

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first and second generation NS decommissioned, those containing damaged NSSS inclusively, have been inspected from radiation magnitude viewpoint. Finishing off of the experimental study of solidifying preservative technology, preventing the contact between reactor high-level radioactive structures and the environment, in case of emergencies involving RC, is in its terminal phase. Radiation situation at two CTB in the North-West and two CTB in the Far East of Russia has been examined. The inspection consisted in dosimetric measurements on the CTB territory, in its buildings and structures, radioactivity determination of SRW, LRW, ground, effluents and bottom deposits in the coastal water area. The relevant inspections of two off-shore technical bases under decommissioning in the Far East have been conducted. Specialists of SUE RDIPE, Russian Research Center «Kurchatov Institute*, Special Design Office of Machine Building, State Research Center» Institute of Physics and Power Engineering*, Special Design Office «Gidropress», Research and Production Enterprise «Ekoatom» and others took part in the measures mentioned. The study of radionuclides migration processes from reactor compartments to the environment in case of their submersion and in case of various emergencies has been started. Analytical research aimed at defining radionuclides release from CTB structures containing LRW to water area around the CTB has been made. In both cases radiation impacts on people via food chains have been determined. Institute of Biophysics, 23 GMPI of Ministry of Defence and others take part in the activities (besides the specialists mentioned above). Initial results of the NS utilization study described suggest certain recommendations and preliminary encouraging conclusions: • it is mandatory that spent fuel is unloaded from the NS and three-section blocks; RC prepared for long-term storage shall not contain SNF; • on removing SNF from the NS reactors, the internal cavities in the reactor vessels shall be filled with preservative; • storage of RC on coastal sites, their number being 50-100 pieces in one place, allowance made for the most severe emergencies involving one RC, does not pose ecological hazard for the population and environment in the region, where the RC long-term storage points are arranged; • submersion of a RC prepared for long-term storage in the open sea does not impose environmental hazard. Research and practical activities are also under way for CTB, and the next most important task is formulation of a concept for comprehensive rehabilitation of CTB with feasibility studies on measures aimed at bringing CTB facilities in nuclear, radiation and ecologically safe state. Thus, essential part of ecological problems arising from NS and other Navy facilities utilization are being solved successfully, though no less problems are awaiting their solving. © 2000 With Author

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C584/00172000 Decontamination and waste minimization techniques in nuclear decommissioning K F LANGLEY and J WILLIAMS United Kingdom Atomic Energy Authority, Didcot, UK

SYNOPSIS Decommissioning of redundant nuclear facilities requires judicious use of decontamination either to reduce the radiation levels or to minimise the volume of radioactive waste produced when a facility is dismantled. A range of decontamination techniques is available, such as scabbling or pressure jet washing. The choice of technique depends on individual circumstances. This paper describes the pros and cons of various decontamination techniques which are available, giving examples from a number of successfully completed UKAEA projects. It also considers instances of novel applications where such practices have not been so successful and the lessons that have been learned.

1.

INTRODUCTION

Since 1996, when the commercial arm of UKAEA was privatised as AEA Technology plc, most of the work of the UKAEA has been focussed on decommissioning redundant facilities remaining from over forty years of nuclear R&D. UKAEA is responsible for four nuclear licensed sites, Dounreay, Windscale, Harwell and Winfrith. In addition, UKAEA remains responsible for the UK's Fusion Programme, which is based at Culham (not a nuclear licensed site). Our long term aim is to restore our sites to a state as near to a 'green field' condition as practicable. In the case of Harwell and Winfrith this is a realistic objective over the next 30 - 50 years. Dounreay and Windscale will have their major decommissioning tasks completed on a similar timescale but may need to remain under institutional control for rather longer. Significant progress has already been made in decommissioning a number of redundant facilities (1). Of its original 18 reactors, UKAEA has completely dismantled five; a further nine have been decommissioned to at least Stage 1 and are under long term care and maintenance (2). Decommissioning work is currently proceeding on four. Other completed decommissioning projects include a fuel handling pond (3), a aplutonium handling facility and

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a chemical laboratory (4) at Winfrith; a variable energy cyclotron and suite of hot cells (5), a high activity handling cell (6), a large chemical engineering laboratory (7), a uranium fabrication facility (8) and a low level solid waste incinerator (9), all at Harwell. In addition, more than 150 free standing gloveboxes and a suite of large fixed gloveboxes have been dismantled. Elsewhere at Harwell a variety of other infrastructure facilities and equipment have been decommissioned, including liquid effluent tanks, drains and pumphouses with associated equipment and a several hundred redundant radioactive transport containers. All of the above decommissioning projects involved the application of a range of decontamination techniques. Decontamination may be defined as the removal of contaminants from surfaces of facilities or equipment by washing, heating, chemical or electrochemical action, mechanical cleaning or other techniques (10). It is a fundamental aspect of decommissioning nuclear plant and equipment. In this paper, we discuss: • • • • •

the objectives and constraints for decontamination the characteristics of decontamination techniques the implications for radioactive waste minimisation the process for selecting a technique for a given requirement experiences (both positive and negative) from completed projects.

2. OBJECTIVES AND CONSTRAINTS FOR DECONTAMINATION There are a number of reasons for wanting to decontaminate. Firstly, there is a need to reduce locally the inventory of radioactive material in a facility or an item of equipment, in order to reduce the radiation levels and minimise the potential for a release of radioactivity to the environment. For example, a shielded cell which has been used for post-irradiation examination of spent fuel will normally have high levels of radioactivity distributed widely across all the internal surfaces of the cell and on the equipment within it. This will include dust and larger fragments arising from the fuel itself. The radiation levels will typically need to be reduced sufficiently to allow man-access for further dismantling operations. In some cases, the facility will be kept in a safe state pending final dismantling at a later date in order to gain the benefit from radioactive decay. However, safety considerations for long term care and maintenance require loose contamination to be removed as far as practicable and any remaining radioactivity to be fixed or sealed-up in order to minimise the hazard. Secondly, there is a need to reduce the quantity of radioactive waste produced. Decontamination can be used to reduce the level of radioactivity on a contaminated surface so that the contaminated item can be categorised at a lower level (e.g. low-level waste rather than intermediate level). In some cases, it is possible to reduce contamination to the "free release" level (defined in the UK as <0.4 Bq g-1), which allows materials to be dispatched for recycling or disposal as non-radioactive waste. The radioactivity removed in the decontamination process is concentrated into a (usually much) smaller volume. Thirdly, there is a need to complete the final stages of decommissioning, which requires decontamination of buildings prior to demolition and remediation of the site to remove contamination from the foundations and surrounding land. However, remediation of ground

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contamination is a major topic in its own right (11) and is therefore beyond the scope of this paper. There are a number of constraints involved in decontamination. Decontamination necessarily involves the generation of secondary wastes. The operator must ensure that there is a suitable disposal route available for these secondary wastes and that their volume is small enough to justify the operation. The use of chemical techniques may create liquid effluents containing materials (such as chelating agents) which could interfere with down-stream processing or could be unacceptable for release to the environment (e.g. heavy metals such as lead). Decontamination operations may involve exposure of the operators to radiation dose. This needs to be justified in terms of the reduction of dose in subsequent care and maintenance and dismantling operations. Similarly, the financial cost of the decontamination operation needs to be justified in terms of the savings which will accrue from subsequent care and maintenance, dismantling and waste disposal. For example, the justification for decontaminating some steelwork to free release level should take account of the cost of doing so (including treatment and disposal of secondary waste) compared with both the scrap value of the steel and the avoided cost of disposal as radioactive waste. There is also a need to take care that a decontamination operation does not exacerbate the contamination either by spreading it more widely within the facility, or by converting it into a more intractable form. For example, mechanical abrasion techniques (e.g. scabbling) can create dust which, if it is not trapped in some way, can spread throughout a facility and increase the overall extent of contamination. Alternatively, pressure washing a concrete surface can have the effect of driving the contamination further into the concrete. Decontamination of building fabric by removing contaminated material can, if taken too far, affect the structural integrity of the building. In some cases, where complete walls or structural supports have to be removed, temporary alternative supporting members may need to be inserted. In some circumstances, it may be difficult to prove conclusively that all of the radioactivity has been removed from a contaminated item. This is particularly true of contaminated equipment such as pumps and motors with complicated internal structures. Painted surfaces can trap alpha activity which cannot be detected through the paint. Paintwork can be removed, but it may be difficult to remove all of it from crevices. In such circumstances there may be little benefit in trying to decontaminate the item as it will still need to be sentenced as radioactive waste because it cannot be shown to satisfy the free release criteria. The end-point for the decontamination operation should be clearly defined in terms of bulk activity (Bq g-1), surface activity (Bq cm-2) and radiation levels (typically uSv hour-1 at the surface). From a practical consideration, it should be noted that monitoring instruments are generally calibrated to give radiation dose rates, which then need to be interpreted to calculate surface activity or bulk activity. The free release level is defined only in terms of bulk activity. The extent to which the activity levels can be monitored in real time depends on the nature of the radioactivity present. Some forms of activity (alpha and soft beta) are difficult to detect. However, in many circumstances it is possible to determine a "fingerprint" that allows the total activity to be inferred from measurements of gamma activity.

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3. CHARACTERISTICS OF DECONTAMINATION TECHNIQUES Decontamination techniques can be classified as follows: Non-attritive methods of simple cleaning such as swabbing, sweeping and vacuuming, which leave the substrate surface essentially unchanged; Chemical (and electrochemical) treatment to remove a layer of the substrate surface, along with radioactivity. The depth of treatment depends on how far the radioactivity has penetrated beneath the surface; Physical attrition to remove a surface layer, such as the scabbling of concrete or milling the surface of lead bricks. Most techniques can be applied either in-situ or to material or components removed to a decontamination facility. The effectiveness of each technique will not be the same in all situations. The method of application must be considered for each technique in the context of the situation in which it is used. The application away from the facility is usually undertaken when the aim is to lower the waste category (for example to clean the material to allow its free release). The application to facilities being decommissioned is often somewhat different from the way they are applied to operational facilities. In the latter instance it is important not to damage the equipment, plant or facility whereas when decommissioning a plant the use of aggressive methods is acceptable. An example of this is that the methods used to decontaminate the primary circuit of a water-cooled reactor during operational shut downs must not affect the long-term integrity of the pressure circuit. When the reactor is being decommissioned more aggressive chemical reagents can be applied to remove more activity and produce a higher reduction in the radiation levels than can normally be achieved with the chemicals used when operational. 3.1 Non-Attritive Cleaning Non-attritive methods remove contamination from a surface without damaging the surface itself. They include simple cleaning techniques universally used in facilities under the heading of good housekeeping, as well as more sophisticated methods such as ultra-sonic cleaning. Inside hot-cells, caves, glove-boxes and any similar facilities it is good practice to keep the insides physically clean by the application of such methods as given in Table 1. 3.2 Chemical Decontamination With chemical decontamination the aim is the remove the radioactivity which has penetrated into the surface of the contaminated item. This is achieved by the dissolution of a layer of the substrate surface. The radioactive material will either end up dissolved in the chemical with a significant amount of the substrate or, where the radioactivity is not itself soluble in the chemical, it will be in suspended in the substrate solution. There are many types of chemical in regular use, the most common being simple mineral acids such as nitric acid. Details can be found in the many publications on decontamination (see, for example, references (10) and (12)). Table 2 lists various methods of applying chemical decontamination processes. Addition of a complexing agent such as citric acid helps to solubilise some radionuclides. In some cases an electrochemical rather than a simple chemical reaction is necessary. The choice of reagent to use will depend on the material

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being decontaminated and the form of the contamination itself. One variation of chemical decontamination is to use chemicals to remove contaminated layers of paint, thus removing the contamination at the same time. 3.3 Physical Attrition Chemical methods only work well when the contaminated surface is metallic. For other surfaces, such as concrete or plaster, methods which strip off layers of surface material by physical attrition may be required. Such methods can be applied either in situ or in a special facility away from the original location. Attritive methods are often used to decontaminate structural material. This then allows it to be released for unrestricted disposal or recycling, or allows a building to be demolished using conventional demolition.

4. WASTE MINIMISATION AND TREATMENT The Nuclear Installations Inspectorate require operators of a nuclear licensed site to minimise as far as is reasonably practicable the rate of production and total quantity of radioactive waste accumulated on their sites. Decontamination techniques can assist in minimising waste volumes by concentrating the radioactivity, leaving the bulk of the once contaminated material or item in an uncontaminated state, or reducing its waste category to a lower level, thereby making for easier storage or disposal. The extent to which this is possible depends on careful characterisation of the nature and extent of the contamination and the correct choice of decontamination strategy. However, the generation of secondary radioactive waste from decontamination processes is inevitable. The ease of dealing with secondary arisings depends usually on whether it is solid or liquid. If the waste is in a solid form, such as the arisings from scabbling or milling or that collected within a vacuum cleaner, then it can be dealt with as part of the normal solid waste route. Sometimes some pre-treatment may be required. Liquid wastes will require treatment in order to either convert them into a solid form or to remove the radioactivity to allow the disposal by a more conventional route. The treatment is often specific and must be considered at all stages. Care must be taken not to produce a liquid waste which when treated produces a solid waste which is unacceptable in a future waste repository (because for example it will change the chemistry in the repository or effect the integrity of the waste-form). It is more often the difficulties of dealing with the secondary waste which influences the decision whether or not to use a particular technique. In the past, UKAEA has tested a number of sophisticated methods for decontamination, but as experience has grown, the selection process tends to favour those which have been proven and for which the secondary wastes are most easily treatable. Melting of metallic components is a potential technique for decontamination and waste volume reduction. In some circumstances it may be possible to remove the contamination as a slag, allowing the metal to be released for recycling. However, assessments within UKAEA have consistently shown that melting is currently not an economic option, so it has not been used.

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5. SELECTING A DECONTAMINATION TECHNIQUE In order to determine whether to use a decontamination technique and select the most appropriate one, the following questions need to be addressed: What is the nature and extent of the contamination? What is the purpose of decontamination? What is the target end-point? What processes are available? What wastes will be generated? Is there a route to deal with these wastes? If so, what is it and is it acceptable? How effective is the process - will it satisfy the objective? How can the process be applied? Is the process ALARP (as low as reasonably practicable)? Does the benefit compare favourably with the cost in both financial and radiation dose terms? Can a safety case for its use be produced and approved? Before selecting a particular process, consideration should be given to the risk and consequences of failure. UKAEA prepares fall-back strategies to ensure a successful outcome. 6. UKAEA EXPERIENCE AND LESSONS LEARNED Table 4 summarises UKAEA experience on a selection of projects where decontamination was a significant issue. In general, the following are the key lessons which can be learned from this experience. Characterise the contamination and plan the work thoroughly. Planning should include a search through past operational records so that accidental contamination can be identified and potential problems anticipated. Keep it simple. After examining numerous options for decontaminating lead bricks using chemical techniques, it was decided to opt for a simple technique of shaving off the surfaces of the bricks with a standard planing tool. This has proved very successful. Use of liquids to wash down a contaminated building should be avoided unless the floor and other surfaces are impermeable. Decommissioning of the Hermes facility at Harwell was made more difficult by previous efforts to decontaminate by washing, which resulted in contamination penetrating into discontinuities in the floor, walls and windows (13). Avoid chemicals, such as chelating agents, which can compromise downstream processing. When possible agents for decontaminating the WAGR heat exchangers were investigated, nitric acid was chosen partly on the grounds that it would have least impact on the site effluent treatment system. A small amount of citric acid was also allowed. Trials involving the spraying of the acid into one section of a heat exchanger showed that this reagent (applied in this way to avoid producing large volumes of secondary waste) did not achieve the desired decontamination factor. A different disposal strategy, not involving decontamination, was adopted.

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When remote operations are required, test them in a mock-up facility. Before using a telerobot for decontaminating a High Activity Handling Cell, a mock-up facility was used to develop specific tooling and train staff (6). The problems ironed out at this stage would have been much more difficult to overcome if they had been encountered in the active environment. Avoid cross-contamination. Extensive use of spray-on strippable coatings can avoid surfaces being re-contaminated once they have been cleaned. This has proved particularly useful in decontaminating plutonium facilities. 7.

CONCLUSIONS

There is a wide range of decontamination techniques available which, if used judiciously, can be invaluable in reducing hazards in redundant nuclear facilities and minimising waste volumes during decommissioning operations. The choice of technique depends on circumstances and projects need to be carefully planned to determine the optimum decontamination strategy. 8.

ACKNOWLEDGMENT

The authors wish to acknowledge the support of the Department of Trade and Industry for funding the decommissioning work described in this paper.

Table 1 Non-abrasive methods of decontamination Technique Vacuum cleaning Sweeping/ brushing/ dusting Washing Swabbing Scrubbing Strippable coating Ultrasonic cleaning Freon cleaning Steam cleaning

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Typical Use Applied to clean up in all types of facility. Can be applied using remote equipment. Vacuum cleaner fitted with output filter. Conventional process — for large areas. Can be undertaken with remote handling equipment. Usually applied where the facility can deal with water. Surfactant can be added. Picks up particles well. Can use various liquids to wet the swabs. For smooth surfaces (but could wash contamination into cracks). Good method for sealing in the contamination. Reduces the likelihood of airborne suspension. Extensively used. Used principally for cleaning smaller components by immersing them in a tank of liquid agitated ultrasonically. Often used on contaminated items removed for repair. For small components intended for re-use which can be put into a special enclosure containing the freon cleaning equipment. No longer acceptable as freon is not environmentally friendly. Can be more effective than simple washing.

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Table 2 Chemical and electrochemical decontamination methods Method of application Circulation through chemical plant Spray reagent Foam reagent Gells Immersion in tank of reagent

Local use

Typical uses Applicable to chemical plant where reagents can be readily circulated. Can produce large volume of waste. Need a method of collecting the liquid reagent so use is limited. It can be difficult to reach all areas. Foam increases the reagent residence time. Foam can be readily collected using wet vacuum cleaning. The foam is then collapsed thus minimising the amount of reagent. Similar to foams in application but removal involves washing off rather than vacuum removal of the reagent. Mainly for components and not applicable for in-situ decontamination. Need to size reduce to fit into tank. Used for lowering the waste category, often to allow free release. Special devices developed to apply reagent to a surface locally then possibly wash after the reagent has done its job. Can be used/applied by a programmed robot.

Table 3 Attritive methods of decontamination Technique Scabbling Shaving/grinding/abrasive Milling Water jetting (with or without abrasive) Jackhammer (and similar devices) Microwaves Explosives Drilling/spalling/routing Sand blasting

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Typical Use For concrete - various tools available commercially For several types of material (concrete, masonry, etc) For metals such as lead bricks For concrete and other materials. This method might not physically remove the substrate so could be considered in Table 1 Concrete Concrete Concrete To remove persistent areas of contamination To clean the surface - paint removal

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TABLE 4 SUMMARY OF UKAEA EXPERIENCE AND LESSONS LEARNT PROJECT / OBJECTIVE

DECONTAMINATION TECHNIQUE

Trials for decontaminating WAGR Heat Exchangers in order to reduce dose levels for dismantling / size reduction operations. Site: Windscale

Washed with recirculating spray of water, then nitric acid (0.5M) and citric acid (0.0025M).

SGHWR Fuel Pond : removal of sludge and decontamination of pond walls.

Sludge was removed from the pond floor by vacuum cleaning. The pond walls were washed by applying a proprietary surfactant solution following low pressure water jetting, with operatives working off a floating pontoon. Lowering the pond water level as work proceeded gave access to successively lower parts of the pond structure while maintaining shielding and trapping contamination. Glovebox ventilation extract system and primary drain lines decontaminated using high pressure water jetting. Building fabric decontaminated using needle guns to remove paint from metal and scabbling to clean concrete surfaces Telerobot (NEATER) used to deploy a variety of tools for size reduction and decontamination of cell internals. Vacuum cleaning and foam washing used for decontamination.

Site: Winfrith

Decontamination of a Plutonium Fuel Manufacturing Facility Site: Winfrith High Activity Handling Cell: removal of cell internal equipment and decontamination to allow man-entry for final dismantling. Site: Harwell Removal of activation and contamination from LIDO concrete bioshield. Site: Harwell Decontamination of a Chemical Engineering Building prior to demolition. Site: Harwell Decontamination of Lead Bricks

Site: Harwell

Radioactivity carefully mapped by core sampling and surface monitoring. Trials funded by CEC on microwave spalling, explosive cutting and diamond cutting to remove active material. Methods used to decontaminate the building fabric include washing, concrete scabbling and paint removal. Hydraulic platform used to access high level surfaces. Following assessment of various options involving chemical and electrochemical techniques, the chosen method was to shave a thin layer off the surface of each brick using a simple planing tool.

OUTCOME / LESSONS LEARNT Trials gave a DF ~ 3. In some circumstances this might be useful but it was judged too little, so strategy changed. Heat exchangers removed and transported to Drigg LLW repository as single large items. Decontamination succeeded in reducing radiation levels to allow free access.

Decontamination successful, but progress slow due to difficulty of working in pressurised suit environment. No contamination above free release level was found during building demolition. Telerobot reliable and easy to use. Vacuum cleaning picked up fragments of 60Co which were the main source of background. Foam washing removed surface activity embedded in oil and grease. All methods worked to some extent, but stitch drilling using a diamond-toothed core drill was the simplest and most effective. 5% of the total mass was removed as LLW; the remainder was free release. The building was successfully decontaminated and subsequently demolished to time and budget. Planing method is quick and simple, with minimal secondary arisings. Several hundred tonnes of lead brick have been successfully decontaminated to free release level.

REFERENCES 1

J Williams, "Decommissioning Experience in the UKAEA." Proceedings of the Second EC Workshop on Decommissioning of Nuclear Installations Technical Aspects, Mol, Belgium, June 1999.

2

M S Barents, "Stage 1 Decommissioning of the Steam Generating Heavy Water Reactor." 5th IBC International Conference and Exhibition on Decommissioning of Nuclear Facilities, London, February 1997.

3

P Bartholomew. "The Decommissioning of the UKAEA's SGHWR Ponds." The Nuclear Engineer Vol. 38 No.l, p. 23-25, January 1998.

4

D Smith. "Decommissioning a Plutonium Fuel Processing Facility to a Green Field Site." The Nuclear Engineer Vol. 41 No.2, p. 60-65, February 2000.

5

D Loughborough et al. "Decommissioning of a Hot Laboratory and Cyclotron Complex." IMechE International Conference on Nuclear Decommissioning, London, November 1995. ISBN 0 85298 955 5.

6

T K Manners. "Decommissioning a High Activity Handling Cell to Stage 3." IMechE International Conference on Nuclear Decommissioning, London, November 1995. ISBN 0 85298 955 5.

7

E Abel and C Hamblin. "Seven Storeys of Decommissioning - The Complete Decommissioning of a Chemical Engineering Building." Nuclear Decommissioning 98 Conference, London, December 1998. ISBN 1 86058 151 X.

8

A J Inns and K F Langley. "Decommissioning Three Minor Nuclear Facilities at Harwell to Green Field Sites." 5th IBC International Conference and Exhibition on Decommissioning of Nuclear Facilities, London, February 1997.

9

M J Sanders and R A Peckitt. "Decommissioning of the Harwell Low Level Waste Incinerator." Waste Management 98 Conference, Tucson, Arizona, February/March 1998.

10

"Decontamination Techniques Used in Decommissioning Activities. Report by NEA Task Group on Decommissioning." Nuclear Energy Agency, Paris, 1999.

11

M Pearl. "Soil Washing Treatment Trials at UKAEA." Waste Management 2000 Conference, Tucson, Arizona, February/March 2000.

12

"State of the Art Technology for Decontamination and Dismantling of Nuclear Facilities." Technical Report Series No. 395, IAEA, Vienna, 1999. ISBN 92 0 1024991.

13

J Stiff. "The waste management implications of the decommissioning and refurbishing of active facilities." Paper Cl, 26th Meeting of the European Working Group on Hot Labs and Remote Handling, Ispra, Sept 1987.

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C584/022/2000 Transuranic waste management at Los Alamos National Laboratory J J BALKEY and R E WIENEKE Nuclear Materials Technology Division, Los Alamos National Laboratory, New Mexico, USA

ABSTRACT Los Alamos National Laboratory (LANL) was the first to ship transuranic (TRU) waste for deep geologic disposal in the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. This was the culmination of a tremendous effort on the part of personnel at the Plutonium Facility (TA-55) where the waste was generated and of Laboratory waste-management personnel. The Plutonium Facility provides a wide variety of actinide research and development capabilities in support of national defense needs. Many of these operations produce a wide variety of waste forms, including transuranic waste from activities in the glove-box lines throughout the facility. Critical to the success of this shipment was the integrity of the waste-management program at TA-55. A group of dedicated wastemanagement professionals worked intimately with actinide processing operations personnel to assure that all waste packages met the WIPP Waste Acceptance Criteria (WAC) and were documented and packaged in accordance with all applicable regulations. Records are largely computerized, eliminating paper forms and improving the integrity of data packages and expediting the review and approval process. Research into waste minimization has succeeded in identifying processes to reduce the volumes of transuranic waste produced and has resulted in the implementation of decontamination and waste-avoidance techniques in the facility. This paper provides a broad overview of current waste-management activities in the TA-55 Plutonium Facility. INTRODUCTION On March 26, 1999, the first transuranic waste to be shipped to WIPP for disposal left LANL by truck, enclosed in the Transuranic Package Transporter Model 2 (TRUPACT-II) designed to meet Department of Transportation (DOT) standards for over-the-road shipments of radioactive materials. This paper focuses on the management of transuranic waste by the Nuclear Materials Technology (NMT) Division at LANL that operates two nonreactor nuclear facilities; the TA-55 Plutonium Facility (PF-4) and the TA-3 Chemical and Metallurgical

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Research (CMR) Facility. Actinide-processing operations at these two facilities generate a substantial quantity of waste that must be segregated, assayed, documented, packaged, labeled, and staged before off-site shipment for disposal.

LOS ALAMOS NATIONAL LABORATORY Los Alamos is one of America's National Laboratories owned by the United States Department of Energy (DOE) and operated by the University of California (UC). The Laboratory is one of the original Nuclear Weapons Complex Laboratories dating back to Project Y of the Manhattan Engineering District during World War II. Consequently, research with radioactive materials has been conducted at Los Alamos for over half a century and remains one of the primary responsibilities of this institution. NMT's PF-4 and CMR are essential in the accomplishment of one of the Laboratory's core missions of Stockpile Stewardship of America's nuclear arsenal. The Laboratory occupies about 43 square miles in north-central New Mexico, about 25 miles northwest of Santa Fe. It is located at an elevation of approximately 7,200 feet on the Pajarito plateau on the east flank of the Jemez Mountains. NUCLEAR MATERIALS TECHNOLOGY DIVISION The Nuclear Materials Technology Division is one of the Laboratory's organizational units responsible for the operation of two nonreactor nuclear facilities at Technical Areas (TA) 3 and 55 that are about a mile and a half apart. These two nuclear facilities are the Plutonium Facility (PF-4) at TA-55 and the Chemistry and Metallurgy Research (CMR) Building at TA3. They are distinctly different facilities from the standpoint of age, design, size, and the nature of operations conducted within them, although both sites generate transuranic waste. The Plutonium Facility is a concrete-reinforced structure designed in accordance with DOE general design criteria for plutonium processing and handling facilities and was completed in 1978. It has a floor area of 150,000 square feet consisting of a service floor and operations floor that is divided into two independent halves and organized into four operating areas. The 100 and 200 areas contain plutonium research and development laboratories, reactor-fuels laboratories, plutonium-238 heat-source fabrication operations, analytical chemistry, and personnel decontamination areas. The 300 and 400 areas contain actinide processes (both wet chemistry and pyrochemistry), metallurgical operations, parts machining, waste operations, and nondestructive assay laboratories. Diverse activities and a rapidly changing project base present a challenge to waste operations. The Chemistry and Metallurgy Research Facility was completed in 1952 and designed in accordance with the 1949 Universal Building Code at that time. It has a floor area of 550,000 square feet and is divided into eight wings arranged around a spinal corridor. Each operations' wing has a basement, attic, operation floor, and filter tower at the end of the wing, opposite the spinal corridor. Utilities run though the attic and basement: the air intake, air conditioning equipment, supply fans, and ventilation distribution ductwork run throughout the attic; the exhaust ductwork, ventilation dampers, and duct wash-down system are in the basement; the filter plenums, exhaust fans, and exhaust stacks are in the filter towers. One wing (Wing 9) was added in 1959 and has two blocks of hot cells with remote manipulators to conduct high-

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radiation-dose-rate activities. The operations conducted in the CMR are basically support functions of radiochemistry and metallurgical analysis. Other activities include nuclear materials safeguards training, nondestructive assay, waste management, gas generation studies, and neutron source storage. The main challenges to waste operations in this facility revolve mainly around legacy waste and obsolete equipment and instrumentation.(1)

TRANSURANIC WASTE The Department of Energy defines transuranic waste as waste contaminated with greater that 100 nanocuries per gram of radioisotopes with atomic numbers greater that 92. TRU radioisotopes are further defined as alpha emitters with half-lives greater that 20 years. This waste is principally generated from defense activities and does not contain fission products. TRU waste is typically generated inside of glovebox lines from actinide processing operations. The Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico has been designated as the national deep geologic (2,250 feet underground in salt beds) repository for the disposal of waste of this type. The waste is emplaced in rooms mined in the salt beds that are sealed when full. The eventual collapse of the burial chambers completes the isolation process. There are two suitable containers for transuranic waste: the 55-gallon drum and the 82.5cubic-foot standard waste box (SWB). Both containers must be enclosed in the TRUPACT-II in order to be transported by truck to the repository. This packaging is in compliance with Department of Transportation regulations for the shipment of radioactive materials. Control must be exercised over the waste from the standpoint of Nuclear Material Control and Accountability (NMC&A) that is accomplished by nondestructive assay and the use of tamper-indication devices on the packages. Any waste containing hazardous materials as defined by the Resource Conservation and Recovery Act (RCRA) is also subject to regulation by the State of New Mexico. Radiological hazards, chemical hazards in the case of mixed waste (radioactive and RCRA hazardous), and mechanical and ergonomic hazards must be defined and mitigated in NMT waste operations. WASTE MANAGEMENT Waste management at LANL is conducted in the Facilities and Waste Operations Group (FWO). This paper presents waste management from the viewpoint of the waste generator. In this case, NMT Division recognized the impact of waste management and regulatory compliance on their operations and established a group within the organization to provide a responsive interface for both waste generators and the Laboratory waste-management organization. NMT waste management has the latitude to tailor their operations to provide optimum support for operations and to protest unreasonable or unrealistic regulation and propose compromises. The management of TRU waste is a cooperative effort by both the generator and wastemanagement personnel. This is necessary in order to optimize the process and minimize the costs associated with waste handling and disposal. The researcher whose process generates the waste is most qualified to provide characterization information on it. Basic information on

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waste composition and content from the generator is essential in determining the isotopic composition and to avoid comprehensive (and expensive) sampling and analysis. The role of the generator is also pivotal in maintaining proper segregation of waste and in working actively to reduce waste volumes. The role of waste-management personnel is to provide program definition and control and to furnish guidance to the generator. Waste-management technicians work with the generator to provide the proper documentation, to file and track records, to package, mark and label, and arrange transportation off-site for waste packages that are properly documented and meet all of the required criteria. Waste management is organized by function with two TRU waste-operations teams (solid and immobilization) at the Plutonium Facility and one composite team (TRU, low-level, recycle and salvage) at CMR. In addition to the waste-management technicians and their supervisors, there are various support personnel that provide specialized assistance in quality assurance, regulatory compliance, clerical, administrative, computer support, budgeting, and planning.

REQUIREMENTS A number of requirements, many carrying the force of law, control the way in which waste is managed. Of primary concern is the Waste Acceptance Criteria (WAC) provided by the repository that addresses many aspects of transportation, handling and disposal. Many requirements in the WAC are driven by the performance expectations of the repository, established during the performance assessment, which are required for licensing. Other mandates are based upon requirements set by the state for the disposal of RCRA hazardous (mixed transuranic) waste. These limitations ensure that the hazardous constituents of the waste will never migrate beyond the site boundary to contaminate the environment. Other requirements are driven by transportation regulations and various acts such as the Atomic Energy Act or DOE Orders. All of these requirement must be consolidated and tracked from their originating legislation, down through institutional documents to the procedures by which waste operations are conducted in NMT Division. DOE regulations, title 10 of the Code of Federal Regulations (CFR), part 835.120 "Quality Assurance," and the disposal facility quality requirements document, CAO-94-1012 "Quality Assurance Program Document," set the quality standards for the NMT transuranic wastemanagement program. These very prescriptive requirements make NMT waste operations a focus of auditors and assessors. Waste-management operations have an integral quality assurance program for this reason. Two full-time quality-assurance personnel are actively involved in the waste-management program. They continually review program documentation to assure that the requirements of the repository and regulators are being met. They regularly conduct assessments in the field, observing operations to ensure that the program is operating as designed. These internal assessments are documented and reported to management on a monthly basis. Key components of the program are monitored for indications of systemic weakness. Significant variances are investigated, the root causes are determined and any identified deficiencies corrected. Two, full-time regulatory compliance specialists also work in NMT waste operations. Their job is to ensure that state environmental regulations regarding the proper marking, labeling, and storage of hazardous and mixed waste are being met. They provide input to the

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Laboratory's operating permit, specifically the part that describes permitted activities at NMT Facilities. They also provide guidance to operations personnel in designing their activities in compliance with regulations and assist in applying for permitting of new activities when necessary. They conduct monthly inspections of waste-storage areas and report findings to the storage-area custodians for correction and to management.

WASTE DISPOSAL Waste-management work areas and personnel reside in the facilities alongside the operations and research activities that they support. This results in the expedient and cost-effective handling of transuranic waste. Waste-management personnel are able to establish close working relationships with operations personnel and become familiar with the processes that generate the waste. They visually inspect waste items and provide guidance to the generator in writing the originating documentation and preparing the items for transfer. The waste must be assayed for SNM content in balancing the process account from which they originate (standard NMC&A practice). This information is also used to verify the waste activity (transuranic or low-level) and to provide SNM content information to the disposal site. Items that meet the WIPP WAC are segregated by waste matrix and isotopic composition and placed directly into the waste container. Information on the waste item, its composition and characteristics, is entered into the Waste-management System (WMS), a computerized database that tracks waste items and records vital information. Items that do not meet the WIPP WAC (i.e., liquids) go to the immobilization process where they are fixed in gypsum cement. The WMS also tracks package limits; fissile content and weight are checked to make sure that they are not exceeded. Packages that are not full are kept locked to control waste that is placed in them. Once the packages are full, they are closed in accordance with DOT requirements and the data packages are completed on the WMS. The final data packages are then routed for the review and approval necessary to ship them to an on-site staging area. Any additional testing that is necessary; head space gas sampling for volatile organic compounds, real-time radiography to verify conformance with the WAC, and nondestructive analysis by tomographic segmented gamma scanning or passive active neutron interrogation, is conducted before final staging and packaging into the TRUPACT-II at another on-site waste facility. CHARACTERIZATION Both the repository WAC and state RCRA regulations require full characterization of transuranic waste. Sampling and analysis can accomplish the required characterization but is very expensive. An alternative to analytical methods is acceptable knowledge (AK). AK is built using existing documentation surrounding the process that generates the waste. These documents consist of process-design information, engineering flow diagrams, mass-balance equations and chemical reactions, chemical and material purchase orders, analysis of feed materials, operating procedures, and operating logs, verified by occasional sampling and analysis. Nuclear material control and accountability practices require that the radioisotopic compositions of nuclear materials be known accurately. These compositions are organized into standard material types (MT) that are used to track SNM in feed, product, and waste streams on the Materials Accounting and Safeguards System (MASS). Operations, generally,

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do not mix materials of differing isotopic compositions. NMC&A in this instance can be used to determine the isotopic composition of SNM in the waste items accurately without resorting to a full spectral analysis. Although all parties (LANL, DOE, WIPP, and NMED) agree upon the use of acceptable knowledge; there is disagreement over what constitutes adequate AK. The first waste that was shipped to WIPP from LANL was very well documented by virtue of the quality-assurance program inherent in the process. This was plutonium-238 waste from radioisotopic thermoelectric generator fuel production. Despite extensive records, the State of New Mexico required that sampling and analysis be done to verify the AK. The results proved, as expected, that the AK was correct but was very expensive and time consuming. The current approach to AK at the Plutonium Facility focuses on seven process streams: nitrate operations, chloride operations, metal preparation, pyrochemical operations, plutonium-238 operations, special isotope operations, and a miscellaneous category. TA-55 is fortunate that extensive waste tracking has been the practice for a number of years as part of the NMC&A System. Operations in the plutonium facility are organized into Process Status (PS) that are used as accounts in tracking SNM in the facility. By using this data, detailed AK is being assembled for waste generated in the Plutonium Facility back to the end of the 1980s. These process streams are cross-referenced to waste streams (waste with similar characteristics) that correspond with Transuranic Content (TRUCON) Codes set up to address transportation requirements. In this manner, characterization information is shown to track from the processes that produce the waste to the final waste package. Waste is certified as meeting the WAC by demonstrating compliance with and verifying the integrity of the wastemanagement program, not by the certification of each individual waste package. This AK methodology is for debris or nonhomogenous waste streams. Waste streams that are processed to meet the WAC (cemented or vitrified waste) must be randomly sampled to demonstrate compliance. This is most conveniently accomplished by sampling during the immobilization process. Legacy waste monoliths from these processes must be sampled by coring in a manner that guarantees a representative sample of waste from the package. This procedure is complicated and is currently in the development stage. The key to meeting the AK criteria is documentation and traceability of waste items back to the processes that resulted in their generation and in capturing as much information on the waste as possible. Meticulous archives of historical operation documentation have been invaluable. WASTE MINIMIZATION AND AVOIDANCE Effective waste management begins in the conceptual design phase of a project. Facility procedures require that waste issues be evaluated at the onset of activity planning. In this manner, problem waste types or expensive disposal requirements can be evaluated and steps taken in the design to eliminate or reduce the cost of waste management with minimal impact to the project. At the very least, design life-cycle costs associated with waste disposal and decommissioning can be identified and planned for at the onset. In some cases, substitution can be made for hazardous substances and the necessary changes made to the process so that quality or efficiency is not sacrificed. Where the use of hazardous substances cannot be

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avoided, the necessary permits must be procured, and steps may be taken to minimize the resulting waste. Several technologies have been developed at Los Alamos to reduce the generation of transuranic or mixed waste. One way of reducing costs is to reduce the activity of the waste. It is much less expensive to dispose of low-level waste on-site than to ship transuranic waste to Carlsbad for disposal. An Electrolytic Decontamination technique has been developed and implemented on a pilot basis for the successful decontamination of decommissioned gloveboxes.(2) Another method for reducing waste activity is the Salt Distillation Process by which chloride salts from pyrochemical processes are distilled to the point where they may be recycled back into the process.(3) This has the effect of reducing the waste volume from this process by an order of magnitude. Another area for both reducing the volume of waste sent for disposal and in eliminating the RCRA hazardous component of the waste is to address the combustible portion of the transuranic waste stream that consists of plastics from containers, packaging, and wipes used for cleaning in the glove-box lines. The techniques of Supercritical Waster Oxidation (4) and Pyroysis (5) have been developed to reduce these waste volumes. They have been demonstrated successfully, and funding is being sought to implement them in NMT Division. Recent reductions in allowable nitrates released to the environment in aqueous effluent from the TA-50 Radioactive Liquid Waste Treatment Facility (RLWTF) has required a dramatic reduction in spent nitrate solutions from anion-exchange operations at the Plutonium Facility and the curtailment of processing. A Nitric Acid Recycle System (6) was developed to reduce the quantity of these solutions that are sent to the evaporator and on to the RLWTF. This system recently completed cold testing and is ready for operation upon completion of its formal readiness review. The restrictive limitation in SNM loading of waste containers because of radiolytic hydrogengas generation and transportation limits has also been addressed.(7) This limit is especially important when dealing with plutonium-238 waste from heat-source fabrication. The higher specific activity of this isotope severely limits the efficiency of loading waste packages. A Vitrification System has been developed and is in the final stages of cold testing before installation in the Plutonium Facility.(8) This will significantly increase the waste-loading limit for waste solutions generated from nitrate operations (reduce the number of waste packages by an order of magnitude) and take the place of the transuranic waste immobilization process. COMPUTERIZED DATA MANAGEMENT The key to a successful waste-management program lies in the integrity of waste documentation and the ability to track waste back to the generating process. Concepts such as AK have placed an additional burden on systems that track and compile waste documentation. Directly charging programs for the waste that they generate is expected to promote the wasteminimization effort, but it requires an extensive accounting system and associated paperwork. All of these concepts have contributed to a proliferation in waste data requirements that have quickly overburdened paper systems.

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Two major projects were undertaken to take advantage of electronic data handling. The first was the development of an NMT computerized Waste-Management System (WMS).(9) This had the impact of eliminating the flow of paper waste documentation out of radiologicalcontrol areas and significantly reduced calculation and transcription errors, making the quality review process much easier. The process of review and approval was also improved, and data could be transferred electronically, instead of relying upon the mail system. This system was implemented for transuranic waste in 1995. The second system is an inventory and tracking system (Waste Inventory Tracking System; WITS) that employs bar codes and hand-held palm pilots with built-in scanners.(10) This takes the portability of using a computerized datamanagement system a step beyond notebook computers and severs the network connection by providing data storage in the hand-held unit. The palm pilot also has the ability to capture signatures, providing legal backing to the characterization process since the President of the United States signed into law legislation to make electronic signatures legally binding effective January 1, 2001. The WITS system is currently in the implementation phase with initial deployment of barcodes and palm pilots in waste operations. Data are being collected in databases set up for the room-trash operation. In this manner, the minor problems expected during implementation can be corrected on this pilot-scale activity before it is implemented division-wide at both nuclear facilities.

SYNOPSIS Waste management in NMT Division is a cooperative process requiring the partnership of both generators and waste-management personnel to be efficient and successful. Waste processes must be robust in order to withstand the rigors of audit, but flexible enough to change readily in response to regulatory drivers. Waste certification is demanding an increasing amount of information in order to demonstrate compliance with wastecharacterization requirements. This same information is also useful in assigning disposal costs and demonstrating waste minimization. The cost of acquiring and compiling this information (the development of new software applications is typically very expensive and carries a high element of risk) must be balanced with the benefits to be gained. Waste-minimization technologies can be easily evaluated for effectiveness but the challenge is in funding their implementation. Unless a comprehensive waste-management program is supported, both the quantity and cost of waste disposal will spiral out of control and can effectively cripple operations, as special interest and activist groups have discovered. The effective management of waste from actinide processing operations is a challenging activity and will continue to present complex issues into the foreseeable future.

REFERENCES Derr, Edward D. and Wieneke, Ronald E. "NMT-7 Approach To Waste Management At Los Alamos National Laboratory's Chemistry And Metallurgy Research (CMR) Facility." LA-UR-00-246. Proceedings of the Waste Management 2000 Conference. 27 February 2 March, 2000, Tucson, Arizona, USA Wedman, Douglas E. et al. "Electrolytic Glovebox Decontamination." LA-UR-97-2422. American Nuclear Society Winter Meeting. 16 - 20 November, 1997, Albuquerque, New Mexico, USA

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3. Hatty, David P. et al. "Pyrochemistry Salt Distillation - Performance Data and Upgrades." Nuclear Materials Technology Division Science and Technology Assessment. 11 - 14 May, 1999 4. Worl, Laura A. et al. "Hydrothermal Processing of Radioactive Combustible Waste." LAUR-98-1063. March 1998 5. Kathios, Daniel J. et al. "Enhanced Pyrolysis" LA-UR-99-2241. " Nuclear Materials Technology Division Science and Technology Assessment. 11 - 14 May, 1999 6. Mullins, Donald W. and Yarbro, Stephen L. "Nitric Acid Recycle by Distillation." Nuclear Materials Technology Division Science and Technology Assessment. 11 - 14 May, 1999 7. Bustos, Leah D. et al. "Hydrogen Gas Generation Measurement on Contact-Handled Transuranic Waste Drums." LA-UR-99-2311. Nuclear Materials Technology Division Science and Technology Assessment. 11 - 14 May, 1999 8. Nakaoka, Ronald K. et al. "Glovebox Vitrification System for TA-55 TRU Waste." LAUR-99-4260. 2 September, 1999 9. Wieneke, Ronald E. and Smith, Kathryn K. "Computerized Waste Documentation at the TA-55 Plutonium Facility." LA-CP-95-0027. 19th Annual Actinide Separations Conference. 12-15 June, 1995, Monterey, California, USA 10. Martinez, Bernadette T. et al. "Automated Inventory Tracking of Wastes At The Los Alamos National Laboratory Plutonium Processing Facilities." LA-UR-99-4324. Proceedings of the Waste Management 2000 Conference. 27 February - 2 March, 2000, Tucson, Arizona, USA © 2000 With Author

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C584/025/2000 Disposition of Russian nuclear submarines outlines of the concept and implementation problems B A GABARAEV, V A SHISHKIN, and V A MAZOKIN Research and Development Institute of Power Engineering, Moscow, Russia

Like any other facility, nuclear submarines have a certain service life at the end of which they have to be decommissioned. There may be also other reasons for decommissioning, such as economic expediency, operational considerations, an accident, or international commitments of Russia. As of today, about 180 nuclear submarines have been removed from service in the Russian navy; most of them are kept waterborne with nuclear fuel on board and are potential sources of radiation incidents in the region. The main problems related to disposition of nuclear submarines in Russia and to safety management in this process, arise from the fact that the existing regional infrastructure was not prepared in good time for mass decommissioning of submarines, as a result of which it is now impossible to provide the required rates of spent fuel unloading from the reactors, followed by its storage and transportation. No definitive decisions have been made in regard to handling large radioactive components of nuclear power systems, to storing and reprocessing radioactive waste. To meet this challenge in an optimal way, a national concept was developed with the primary objective of defining the fundamental principles of and approaches to assuring nuclear, radiation and environmental safety of handling radioactive components and materials from decommissioned nuclear submarines. The concept should also help optimise allocation of investments with regard to the financial and economic realities of the country. The concept of disposition of nuclear submarines in Russia is based on the following main principles (transparency 1): • priority of nuclear, radiation and environmental safety goals at all stages of the submarine disposition programme;

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unloading of spent nuclear fuel from submarine reactors as the first priority action; «closed» spent fuel handling cycle involving the reprocessing capabilities of the Industrial Association «Mayak»; «deferred» disposition of nuclear power system components - presenting a radiation hazard - after prolonged (for about 70 years) cooling inside the reactor compartment; optimal use of the available regional infrastructure for disposition of nuclear submarines; reliance on the protective properties of the reactor compartment hull to seal off radioactive components from the environment; placement of solid radwaste, produced during submarine operation and disposition, inside the reactor compartment; overall adherence to the nuclear nonproliferation and national security requirements. The above principles have been substantiated by research findings. For example: The principle of «deferred» disposition of radioactively hazardous components is based on the fact that most of the induced activity is concentrated inside the reactor and in the metal structures nearest to it (-99 %), with about 1 % found in the reactor compartment components and peripheral equipment (transparency 2). Although reduced - roughly by a factor of 10 - due to spent fuel unloading from reactors, the accumulated activity will still be as high as about 100 thousand Ci in one reactor (transparency 3). After a reactor is shut down, the total radioactivity will decline depending on the decay periods of the predominant radionuclides found in the structural materials (Fe , Co , Ni63 , Mo93). The table in transparency 4 suggests that activity of type A (hard gamma radiation), which governs the radiation field patterns and the personnel exposure, will not drop to the levels specified in the current norms (no more than 30 uSv/h for personnel) for at least 70 hours after the reactor shutdown. It is this fact that accounts for the cooling time recommended for reactor compartments with nuclear power systems inside. A shorter cooling period means a greater quantity of solid radwaste subject to reprocessing, conditioning and storage. The closed fuel cycle involves: • unloading of spent nuclear fuel from submarine reactors; • temporary storage of spent fuel in the region (in the storages of floating or onshore maintenance bases);

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• dispatching by special rail transport to a reprocessing plant. The radiation and environmental safety should be ensured at all stages of the submarine disposition process through the use of safe spent fuel and radwaste handling technologies based on the current norms and regulations, as well as by sealing off reactor compartments which house within their standard and additional protective barriers radioactively hazardous components of the nuclear system (reactors, pumps, heat exchange equipment, etc.). The main barriers to escape of radionuclides from the reactor compartment to the environment are provided by the high-strength hull of the compartment and its end bulkheads. Nuclear safety is ensured by a package of engineering and organisational measures in accordance with the current legislation as well as with the norms and regulations specific to each stage of the submarine disposition process: • the operating organisation bears full responsibility for the safety of the nuclear system, as well as for proper handling of nuclear and radioactive materials. To carry out the above activities, the operating organisation should have special permits (licenses) from the appropriate regulatory authorities; • in case of waterborne storage, nuclear safety is ensured by securing reactivity control components in the positions affording maximum reactor subcriticality, and by precluding any unauthorised movement of such components with the use of either normal drives or manual backup devices; • in the course of preparations for defuelling, nuclear safety is ensured in keeping with the regulatory and operational requirements pertaining to activities of potential nuclear hazard. Besides, before opening up the reactor lid, the reactor vessel is drained for added nuclear safety; • nuclear safety management during spent fuel unloading relies on the use of special equipment (OK-300PB, OK-300PBM) as well as on adherence to the technological and procedural requirements; • during spent fuel storage and transportation, nuclear safety is provided by the design of shipping casks and storage facilities, as well as by adherence to the technological and procedural requirements; • no nuclear hazard is associated with preparation, cutting and storage of reactor compartments after spent fuel has been unloaded. The submarine disposition procedure, schematically shown in transparency 5, amounts basically to the following: • removal from naval service and inactivation of a submarine; • temporary waterborne storage of the submarine (with spent nuclear fuel on board);

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submarine transfer to the disposition agent; preparation and unloading of nuclear fuel from reactors; spent nuclear fuel packaging and shipment to a reprocessing plant; packaging and preparation of the reactor compartment for extended storage; cooling (extended storage) of the reactor compartment in the prepared storage facility; disposition of the reactor compartment at the close of its storage - salvaging of «clean» metal, packaging and transfer of solid radioactive waste to storage (disposal). The problems of the Concept implementation lie mostly in the small capacity of the existing infrastructure and in its lacking some vital capabilities, such as reactor compartment storage facilities, handling devices (docks, cranes, vessels), free space in onshore spent fuel storage facilities, etc. Disposition of nuclear submarines being an unprofitable business, enhancement of the above capacity calls for substantial allocations from the state budget. Keeping decommissioned submarines on the float after they come to the end of their service life, is also very expensive and, more importantly, far from safe. For this reason, a makeshift procedure has been adopted in Russia for disposition of nuclear submarines, which forgoes the one-compartment configuration (transparency 6) in favour of a three-compartment option (transparency 7), i.e., the reactor compartment is cut out of the submarine hull together with the adjacent compartments which give buoyancy to the former and allow keeping it waterborne for about 10 years. Once the storage facility is in place, this procedure will be abolished and the 3-compartment sections will be altered to the 1compartment configuration. Experts from the RDIPE (NIKIET) in team with other Russian institutes carried out investigations on various design and site options for a storage facility, with the final choice to be followed by development on the basis of comparative technical and economic studies. The storage design options were considered for the following possible sites with regard to the climatic, geological and geographic conditions and to the location of facilities for construction of reactor sections: In the Northern Region (transparency 8): Kola Bay; Novaya Zemlya Archipelago; Barents Sea Coast; shipyard territory (water area)

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In the Eastern Region (transparency 9): Vladimir Bay; Strelok Bay; shipyard territory (water area). The capabilities of the following types of storage for one-compartment reactor sections (transparency 10) were examined as applied to the above sites: open shore sites; shore sites of industrial type; onshore trenches; underground tunnels; coastal waters, at a depth of 25 to 30 m. Analysis of the research findings shows that it is technically feasible to build storage facilities in the regions under consideration - but only of certain types for each of them. The technical and economic indicators of the considered options are compared in the table (transparency 11). Technical and economic indicators of storage options Storage site and type

1 . On an open shore site: • Kola Bay • shipyard • Vladimir Bay 2. On a shore site of industrial type: • Kola Bay • shipyard • Vladimir Bay 3. In underground tunnels: • Barents Sea coast • Vladimir Bay • Strelok Bay

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Relative cost Relative capital Relative annual Estimated time of cost per 1 reactor operation costs storage of capital construction construction, years compartment

1.0 0.48 1.45

1.0 0.5 2.1

1.0 0.48 2.1

4-5 4-5 5-6

1.4 0.76 1.8

1.4 0.9 2.7

1.34 1.0 2.5

5-6 4-5 5-6

0.25 1.8 0.2

0.26 2.76 0.25

0.36 2.5 0.43

3-4 6-7 3-4

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Storage site and type

4. In onshore trenches: • Kola Bay • Novaya Zemlya • Vladimir Bay 5. In coastal waters (under water): • Novaya Zemlya • shipyard • Vladimir Bay

Relative cost Relative capital Relative annual Estimated time of storage cost per 1 reactor operation costs of capital construction, years compartment construction

1.25 1.53 1.5

1.25 1.5 2.2

1.2 1.46 2.16

4-5 810 5-6

0.25 0.1 0.28

0.25 0.1 0.44

0.2 0.07 0.4

2-3 2-3 3-4

On the strength of the investigation results it was recommended to build storage facilities at Saida Firth (Northern Region) and Razboinik Cove (Eastern Region) as an open shore site. A few words about solid radwaste management: «deferred» disposition of reactor compartments makes it possible to minimise the quantities of solid radioactive wastes going into reprocessing and storage in the course of submarine disposition. In view of the fact that the reactor vessel is not subject to disposition even after the end of cooling, the internal space of the reactor may be used for accommodation of solid radwaste, to wit - hot lower parts of shells with absorber rods of the control and protection system. This technology is resorted to in unloading nuclear fuel from reactors. The components dismantled during preparatory operations are subsequently reinstalled in their normal positions in the reactor compartment; the radioactive components piled up during submarine repairs or upgrades, such as steam generators, pumps, heat exchangers, etc., may be placed in the free space of the reactor compartment in accordance with the pertinent regulatory document. With the aim of implementing the Concept, a Draft Federal Submarine Disposition Programme has been drawn up, which - upon appropriate finalisation - is going to be submitted to the Russian Government for approval. In conclusion, the following may be noted. The Concept implementation entails huge expenses. Bearing this in mind and recognising the potential radioactive contamination hazard to the seas washing the coasts of Russia and other countries (Barents Sea, Sea of Okhotsk) in the event of inordinate protraction of the submarine disposition process in Russia, the international community (Norway, Sweden, USA, UK, Japan and other countries) is rendering assistance to Russia in overcoming some of its problems, such as reprocessing of liquid and solid radioactive wastes, unloading and shipment of nuclear fuel. However, this subject falls beyond the scope of the current presentation. I would like to draw your attention to a list of the most important, top-priority activities which may significantly speed up submarine disposition as a normal process and minimise its duration. The activities listed in transparency 12 cover a broad field with ample opportunities for international collaboration both at the intergovernmental level and as joint efforts of companies, institutes and organisations. © 2000 With Author

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C584/009/2000 Management of accumulated operational wastes at BNFL's decommissioning reactor sites A T ELLIS, L McTAGGET, and R I HEY Reactor Decommissioning Unit, BNFL, Ffestiniog, UK

SYNOPSIS This paper describes the Intermediate Level Waste streams that have been accumulated during the operation of Berkeley, Hunterston A and Trawsfynydd. The existing storage arrangements are described, drawing attention to features that affect continued storage or make recovery challenging. BNFL's generic strategy for the long term treatment of the wastes are described, together with the work being carried out at each site.

1 INTRODUCTION Berkeley, Hunterston A and Trawsfynydd are three of the first generation of civil nuclear power stations built in the UK. The three stations were commissioned between 1961 and 1965 and were taken out of operational service between 1989 and 1993. The nuclear fuel has been removed from all three sites and dispatched to Sellafield for reprocessing. BNFL has announced that Hinkley Point A power station will not be re-started and that Bradwell will close in 2002. Latest operating dates for the other first generation stations have also been announced. These stations all have wastes which are similar to those described in this paper and which will be treated in similar ways. Detailed, site specific, plans for the later stations are still being prepared and they will not be discussed further. During operation of nuclear power stations, low level and intermediate level nuclear wastes are produced. (The third category of nuclear waste, using UK terminology, is high level, or heat generating, waste, which is produced during fuel reprocessing and therefore is not generated at power station sites.) In the UK, Low Level Waste (LLW) is defined as waste having specific activity levels of less than 12 GBq/Te beta/gamma emitting isotopes and less than 4 GBq/Te alpha emitters. Subject to various conditions on physical form, LLW can be disposed of at the BNFL operated national facility at Drigg in Cumbria. Therefore LLW is not accumulated at power station sites in any significant quantity.

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Operational ILWs can take a number of forms and can be stored wet or dry. A general distinction is made between slurry type wastes (eg sludges or ion exchange resins) and solid wastes (eg flux flattening bars). The strategy developed at each site to deal with ILW must address site specific issues (such as the design of existing storage facilities) and must also be consistent with the company wide (or generic) ILW and decommissioning strategies. There is a requirement for nuclear decommissioning and radioactive waste strategies to be reviewed every five years by the Health and Safety Executive. This review is based on a statement of the strategy prepared by the operator of the facility. The most recent review covering BNFL's reactor sites was submitted in April 2000. The main hazard associated with ILW processing is the potential radiation dose to plant operators. The existing storage locations are all adequately shielded, but care must be taken to ensure that the design of recovery and conditioning plant also provide adequate protection. In addition the waste packages must be properly shielded. During recovery of FED the remote possibility must be considered that some uranic material may have contaminated the waste and produced unstable uranium hydride during storage. The recovery equipment must be designed to deal with the possibility that uranium hydride could provide a source of heat during the recovery process. 2 BNFL's GENERIC RADIOACTIVE WASTE STRATEGY In general the strategy for accumulated operational ILW is to ensure passive safe storage as soon as possible after shutdown of the site and to retrieve and encapsulate those wastes not already in a safe passive state. Where waste is to be encapsulated for disposal, packaging arrangements are agreed in advance with UK Nirex Ltd, thus confirming that the waste package should be acceptable for disposal. 2.1 Slurry form wastes These consist of sludges and ion-exchange materials that result from the filtration and treatment of liquids, for example from fuel cooling ponds. The preferred option for these wastes is to retrieve them from existing storage tanks and to solidify them in cement within stainless steel containers. 2.2 Fuel Element Debris (FED) Following removal of used fuel from the reactors, external components of the fuel elements are removed and stored on site, prior to fuel shipment to Sellafield. This FED consists mainly of Magnox metal, although at Berkeley and Hunterston A it also includes graphite. FED will be retrieved from the storage vaults and could be processed in one of two ways, dissolution or encapsulation. Dissolution involves dissolving the Magnox metal in a weak carbonic acid solution retaining the bulk of the radioactivity as a residue, which is then solidified in cement. The dissolved metal is discharged to sea following any necessary treatment. Dissolution is not appropriate

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where the waste has a large graphite component (as at Berkeley and Hunterston) or where there is no direct disposal route to the sea (as at Trawsfynydd). Thus encapsulation is the preferred option for the three sites considered in this paper. 2.3 Miscellaneous Activated Components (MAC) MAC arises from the maintenance or replacement of items used in the reactors. These items are normally accumulated in concrete vaults and storage tubes built into the reactor biological shields. The majority of the radioactivity is fixed by being bound within the material rather than being on the surface. It will often be possible to show that the existing storage locations meet the requirements for passive safety, provided that water ingress is prevented. The generic strategy for MAC is, therefore, to retain it within existing storage locations and to deal with it at the same time as reactor dismantling. 2.4 Other items Most sites have miscellaneous contaminated items in stores. Some of this waste will be LLW, some of the remainder can be decontaminated to LLW levels. What remains will be encapsulated in a cement grout. Sites also have quantities of desiccant material used during the operational period to remove moisture from the reactor gas. The strategy for desiccant is to solidify it in a cement grout.

3 BERKELEY 3.1 Existing stores The majority of operational wastes at Berkeley are stored in four underground vaults which are approximately 18 metres long, 6 metres wide and 8 metres deep. Vaults 1, 2 & 3 contain wastes. Vault 4 is empty. There is also a chute silo which is slightly shorter than the vaults, narrower and extends above the ground so that it is 12 metres deep. The total volume of wastes stored in the vaults is approximately 1500 m3, which is almost equally split between the three vaults. The volume of MAC in the chute silo is comparatively small. Some of the waste has been corroded due to water ingress to the vaults. There is little segregation of waste streams and some is in cans will need to be opened. 3.2 Retrieval work A project has been set up to retrieve the wastes from the vaults and segregate mobile wastes (defined as all pumpable wastes with solids less than 5 mm diameter) from solid wastes. The solid wastes will be packaged to Nirex waste package requirements. The mobile wastes will be stored in tanks. To complete the project the facilities used to undertake the work will be decommissioned. The arrangements developed to implement the project fall into four separate areas, and are the subject of separate facilities constructed on the BPS site: The retrieval facility - to retrieve the wastes from the vaults The process facility - to separate solid and mobile wastes, characterise all wastes, and package solid wastes.

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The ILW store - to store the packaged wastes for up to 50 years. Approximately 400 3m3 boxes and one hundred 500 litre drums are expected to require storage. The mobile waste store - to store the mobile wastes 3.2.2 Retrieval facility The retrieval facility is situated over the vaults, and is contained in a large single skinned building. The waste vaults have been fitted with a new ventilation system and associated fire detection system, using optical smoke detectors. Fire suppression is by argon flooding of the vaults actuated by manual operation of valves on a fixed pipework system. Each vault has five penetrations created by stitch drilling. The largest penetration is in the centre of the vault, and lies within a reinforced concrete tunnel built along the centre line of all of the vaults. There are two circular penetrations on each side of the tunnel. A hydraulically operated shielded cover covers each penetration. A skip mast is located on rails on the top of the tunnel above the large central penetration. The function of this device is to lower a skip into the vault, which is loaded by two manipulators, one located on each side of the tunnel. The skip mast raises loaded skips from the vault into the tunnel, where each skip is placed on a bogie running on rails. The skip on its bogie travels through the tunnel and into the despatch area, which is over vault 4. A buffer store area has been constructed in vault four, and the skips can either be stored here or posted directly into a shielded overpack which transports them to the process facility. The chute silo is provided with two manipulator penetrations only. In addition there is a shielded horizontal opening leading into the tunnel, through which a reduced height skip will be lowered into the chute silo. The skip mast is not used on the chute silo. 3.2.3 Process facility The process facility is located in an existing building. The process plant comprises a shielded cell with walls approximately 900 mm thick. Within the process facility is a sorting area, functionally split into two. On the North side is the solids handling line, and on the South side is the can opening facility and the mobile waste processing line. Skips are posted into the process facility from the shielded overpack, and if the skip contains loose FED the contents are emptied by lifting and tipping on to a vibrating table along which they progress before entering a 3 m3 package. Instrumentation fitted above the vibrating table includes germanium gamma assay equipment, a gamma camera and dose rate measuring equipment such that high dose rate items can be detected. If the material in the skips is in containers, these are passed to the other side of the process cell where they are weighed, assayed by a Sodium Iodide system and opened The packages are grouted, cap grouted, lidded, weighed and swabbed. Dose rates are measured and the packages are posted out from the other end of the process cell in an overpack and they are taken to the ILW store.

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3.2.4 ILWstore The ILW store is designed to be a passive store and is situated in Blower hall 2A, which has been deplanted to foundation level and completely rebuilt as a 900 mm thick reinforced concrete structure. The ILW store is capable of accepting approximately 750 3 m3 boxes. A stillage containing four 500 litre drums can be substituted for one 3 m3 box. The boxes are stacked five high in the store. The majority of each stack of boxes will be below ground level. 3.2.5 Mobile waste store The mobile waste store, which will be used to store both intermediate level mobile waste and low level mobile waste, utilises existing tanks in the Caesium removal plant (CRP) including two tanks which were installed in the 1980s but which have not been used. One of these tanks will be used to store ion exchange material. Five other existing tanks within the CRP will be used for the ILW and LLW mobile wastes. A sixth existing tank will be used to permit supernatant water to be removed from the waste tanks and reused for resuspension in the process facility. Construction of the plant is complete and inactive commissioning is underway. Active commissioning is due to start in December 2000, and the vaults emptied by early 2004. 4 HUNTERSTON A 4.1 Solid Active Waste Building (SAWB) The SAWB is a large two-storey reinforced concrete building designed principally for the storage of FED. The main part of the ground floor consists of five in-line bunkers. Waste was discharged into the bunkers through plugged holes in the roof, which are accessed via a loading room on the upper floor. Each bunker has a capacity of 630 m3 with dimensions of 10.7m by 9m by 4m high The waste stored in the SAWB bunkers include FED, which at Hunterston A includes large quantities of graphite as well as Magnox splitters. A feature of the Hunterston fuel channel arrangement was the graphite sleeve that surrounded each fuel can During fuel discharge the graphite sleeves and bottom reflector were removed. Following separation from the fuel, the graphite sleeves and bottom reflector underwent a gross volume reduction process in a mechanical cracker unit. The resultant graphite pieces are up to 300mm in size. The SAWB also contains redundant fuel handling and waste processing equipment, filters and filter dust bags and redundant maintenance equipment. The total volume of wastes stored in the vaults is approximately 2407 m3. There is 628 m3 of Magnox in bunker 1, 1761 m3 of predominately graphite in bunkers 2-4, and 18 m3 of miscellaneous waste in bunker 5 Following careful examination of the option of leaving the SWAB contents in place, it has been decided that the waste should be retrieved and packaged. Retrieval and encapsulation options are being reviewed. The key issues are, the location of the encapsulation plant and the location for package storage.

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The location of encapsulation was considered with regard to the safety and suitability of conducting this operation on site versus shipping the waste to Sellafield to be processed in existing BNFL facilities. Issues surrounding the transport of raw wastes and double handling of wastes have led to on site processing being the preferred option. The location of interim package storage pending the availability of the national repository was assessed. In view of the large volumes of waste and transport issues it was considered that a purpose built storage facility would be provided on site. Planning consent for the new store has recently been obtained. It is anticipated that the retrieval and packaging process will be similar to that at Berkeley. Current plans are for technical options to be evaluated by 2001, the facilities built by 2003 and for recovery and packaging of the waste to be completed by 2010. 4.2 Slurry form wastes The majority of slurry form wastes are sludges stored in five reinforced concrete sludge tanks which are situated below ground with a concrete top slab. Three of the tanks have a capacity of 90m3 each while the other two have a capacity of 45m3 each. A much smaller quantity of ion exchange resin is stored in an above ground 50m3 stainless steel vessel within a shielded cell. The total volume of raw waste is approximately 260m3, 10 m3 being ion exchange resin. A Best Practicable Environmental Option study was conducted for the sludge and resin wastes, which are currently stored on site. The preferred option is encapsulation in cement and packaging in Nirex boxes. These boxes will be stored on site in the new store. It is anticipated that the encapsulation plant installed at Trawsfynydd will be employed to process these wastes. During the development work for dealing with these wastes it has been identified that approximately 20% of the sludges will meet LLW activity concentrations once encapsulated and would be suitable for disposal at Drigg. This option is actively being pursued. Recovery of these wastes has been programmed to take place between now and 2005. 4.3 ILW store Existing locations and redundant buildings on site were initially considered for potential storage locations for packaged wastes. However, the quantities of packages meant that no one location would be suitable. Planning consent was therefore obtained for a purpose built ILW store on site. The footprint of the store is expected to be approximately 206m x 22m and will provide storage for approximately 2700 Nirex packages. The optimum design being for unshielded packages in a shielded store.

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5 TRAWSFYNYDD 5.1 Miscellaneous Activated Components (MAC) During reactor operations neutron absorber bars were routinely discharged from the reactor cores into underground vaults, one beneath each reactor building. Other material such as fuel route components and thermocouple wires were also deposited in these vaults. There are about 6m3 of this waste on site with a total activity levels to be 6.7 x 1014 Bq. Each vault is about 3m deep and about 7.5m by 6.5m in plan. The vaults are equipped with sumps to collect any groundwater which seeps in due to their location below the level of the water table. Long term storage of the waste in these vaults would have required an engineered system to prevent the accumulation of water. Thus it would not be possible to demonstrate passive safe storage for the MAC vaults at Trawsfynydd. Therefore, the Trawsfynydd strategy is to recover and condition this waste. (There are smaller quantities of similar wastes stored in tubes built into the reactor biological shields - this waste will remain in place, subject to regulatory approval of the associated safety case). Access to the area immediately above the MAC vaults is possible. This has allowed access holes to be created in the roofs of the vaults which will allow remote controlled manipulators to be used to remove the waste. The waste will be placed in two types of containers to give good packing into 3m3 boxes, one type for flux flattening bars and one type for the other miscellaneous waste. The boxes will be filled with grout to form a monolithic waste package. After lidding the boxes are placed into a concrete overpack which provides sufficient shielding to allow the packages to be moved. Initially the packages will be stored in converted gas circulator hall basements (which are adjacent to the recovery and encapsulation plant). For the care and maintenance period the packages will be stored in a purpose built building. Non -active commissioning of the MAC facility has been completed. 5.2 Fuel Element Debris (FED) At Trawsfynydd, this waste stream also contains nimonic springs that became detached from the fuel elements during desplittering. The springs have a high Cobalt content and become highly activated resulting in significant doserates from the mixed waste. The mixed waste has specific activity levels between 1 and 10 TBq/Te, depending on the spring content. During the majority of the period of operation FED was placed into vaults, one per reactor each consisting of 16 cells roughly 2m x 2m x 4m high. Towards the end of operation these vaults became full and a purpose built Magnox Debris Handling and Sorting Facility (MDHSF) was built which allowed the waste to be sorted (to remove high doserate items such as nimonic springs) before being high force compacted into 500 1 stainless steel drums. These drums are currently in store - they will be grouted during the current programme of work. In line with the generic policy the FED will be retrieved and packaged, dissolution not being an option at Trawsfynydd. Consideration was given to using the MDHSF for the vault stored waste. However, the number of nimonic springs in the vaults and the possibility sludge contamination on some of the waste meant that major modifications would be needed (particularly to the ventilation system) and a system for dealing with high doserate items would be needed. (During

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operation these were dispatched to Sellafield). Therefore new plant is being installed to recover and sort the vault FED. Sorting is carried out to ensure that limits on the quantity of fissile material in each container can be met and to allow high doserate items to be placed in the centre of a container, thus minimising package doserates. The sorted waste will be placed into a 3m3 box, lightly compacted and grouted. After lidding the box will be placed into a concrete overpack and transported to an on-site store (initially a converted gas circulator hall basement, eventually to a purpose built ILW store). It is expected that the FED facility will be ready to start operation by the end of 2000. 5.3 Ion exchange resins The requirement to limit discharges of low levels of activity into Trawsfynydd Lake and fuel cladding corrosion problems in the fuel cooling ponds during the early years of operation have resulted in heavy use of ion exchange material at Trawsfynydd. When built a single storage tank of 100m3 capacity was installed. This coated mild steel tank is contained within a concrete bunker to provide shielding and secondary containment. The ion exchange material is stored under water. During the 1970s it became apparent that there was insufficient storage space in the original tank. Two additional, stainless steel, tanks were installed, each with a capacity of 70m3 and then a purpose built resin encapsulation plant was constructed to allow material to be recovered from the stainless steel tanks and encapsulated in a polymer resin. This plant was successfully operated during the 1980s and 394 sea dump type drums were filled with waste. These original drums continue to be stored on site (in a purpose built store associated with the encapsulation facility) as sea dumping is no longer available. At Trawsfynydd this waste stream has specific activities between 0.5 and 6 TBq/Te. A review of the options for implementing the strategy of recovery and encapsulation of this waste identified that the existing plant could be re-used or that a new facility could be installed to encapsulate the material in cement. It was decided that the installed plant should be re-used as this allowed an earlier start on the work and reduced capital expenditure. To minimise the modifications needed it was decided that sea dump drums would continue to be used. Although these might need to be placed in standard containers before transfer to a repository some way of dealing with this waste form would be needed because of the 394 drums already on site. Also, the sea dump drums are self shielded which simplifies movement on site. The encapsulation plant has been successfully re-commissioned and one of the stainless steel tanks has been emptied. The transfer system to allow the contents of the original tank to be recovered is being installed and the plant will be operating again later in 2000. The original drum store has been modified by installing additional restraints to allow the drums to be stored in a square rather than a triangular configuration. However, this does not provide sufficient capacity and part of the old mechanical workshop has been converted into a short term store to allow encapsulation to proceed. During the current phase of work the drums will all be transferred to a new storage building and the other stores dismantled.

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5.4 Sludges Trawsfynydd has about 40 m3 of sludges, the majority in a tank similar to the original ion exchange material tank and some in a vault which forms part of the fuel cooling ponds structure. The ponds sludges will be transferred to the sludge tank. The sludges have specific activities of about 0.3 TBq/Te and are stored under water. BNFL has procured a Transportable Intermediate Level Waste Solidification Plant which is designed to encapsulate sludges and similar waste. This plant has been installed at Trawsfynydd and will soon be commissioned to encapsulate the sludge in a cement matrix using the lost paddle technique. The waste will be packaged in Nirex large liners and stored in gas circulator hall basements before transfer to a purpose built store. 5.5 Storage of packages In order to allow an early start on ILW recovery and packaging, short term storage capacity has been provided as described above. The option of using gas circulator hall basements, and the fuel cooling ponds building for storage of packages until the national repository is available was considered. This would have reduced the need to construct new buildings, limiting the changes to the appearance of the site and minimising expenditure. However, these existing buildings do not have sufficient capacity and satisfying the requirement for passive safety (ie minimising the reliance on engineered systems) would have been difficult. Therefore, it has been decided that as new store (approximately 27m x 100m x 15m high) will be constructed. As this will take place fairly late in the work programme the store can be placed between the reactor buildings allowing the architectural design of the three buildings to be integrated. 6

CONCLUSIONS

BNFL has a well developed strategy for dealing with operational ILW at its reactor sites. At Berkeley, Hunterston and Trawsfynydd site strategies have been developed to take account of local circumstances. Good progress is being made with the implementation of these strategies. This will provide a good basis for the efficient management of wastes at those stations still operating.

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Transport and Storage

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C584/013/2000 Transportation of spent fuel in Japan M NAKAJIMA Nuclear Fuel Transport Co. Limited, Tokyo, Japan

ABSTRACT The present paper reports current status of spent fuel transport in Japan. Forty units of the NFT Type packagings were fabricated, a transport ship built to the requirements of the INF Code, 150-ton wharf crane and vehicles prepared for the transport. The first shipment of spent fuel to the JNFL reprocessing plant was carried out in October 1998 but subsequent disclosure of falsification of packaging's neutron shield resin delayed the second shipment to September 1999. Regular transport is expected to start in latter half of 2000.

1 INTRODUCTION With the 21st century stretching out before us, we feel strongly that the most realistic option for securing stable sources of energy for Japan's future is nuclear power, and in support of this vision, we must faithfully push forward with the establishment of the nuclear fuel cycle. Japan has 52 nuclear power plants generating 45GW of electricity at 17 sites. Nuclear energy supplies about 36% of Japan's electric power. Japan is making steady progress to establish the nuclear fuel cycle essential for the nation's long-term energy security. An essential component of this is the reprocessing of nuclear spent fuel where uranium and plutonium are recycled for further use. A commercial reprocessing facility of Japan Nuclear Fuel Limited (JNFL) in Rokkasho-mura, Aomori Prefecture (Fig.l) is now on-line. Nuclear Fuel Transport Company (NFT) is responsible for the domestic transportation of spent fuel from nuclear power plants to the reprocessing facilities. Since 1978, NFT has transported spent fuel from Japanese nuclear power plants to the Tokai Reprocessing Plant of the Power Reactor and Nuclear Fuel Development Corporation (PNC), now named the Japan Nuclear Cycle Development Institute (JNC). At present, the total volume of spent fuel transported by NFT has reached 860 tons.

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For the transportation of spent fuel, NFT has constructed and/or prepared an exclusive-use vessel, a 150 ton wharf crane and specially designed overland carriers. The first transport of test spent fuel was carried out in October 1998. On this occasion, the transport took place to allow calibration tests of the burn up measurement equipment in JNFL's reprocessing plant in preparation for full-scale transportation of spent fuel. Just after the first transport of test spent fuel, falsified data was discovered on the neutron shield materials fabricated in NFT-type spent fuel transport casks. It has taken almost one year for NFT to completely solve the various issues caused by the data falsification problem, and to re-start the transportation of spent fuel in September, 1999. To start regular and full-scale transport of spent fuel, discussions to finalize the Safety Agreement are progressing among the local government, regional municipalities and the owners of the respective facilities. Regular and full-scale transport of spent fuel is expected to start again from the autumn of the year 2000.

Fig.l Locations of Japanese Nuclear Power Station

2 OUTLINE OF SF TRANSPORT PROCESS Transport casks containing spent fuel are carried from the nuclear power station to the nearest port and loaded on board the exclusive-use ship. The ship carries the casks to Mutsu-Ogawara (MO-port) in Rokkasho-mura. After arriving at MO-port, the casks are unloaded and stowed on special vehicles. The vehicles carry the casks to the reprocessing facility on a private JNFL road. Part of this road is public(l). MO-port is located on the Pacific coast of Aomori-pref.(Fig 1). This port is public and managed by local government. The "Rokuei-Maru" class ships come alongside Takahoko Quay at the MO-port. The length of the quay is about 260m and the depth of water is 7.5m. The front sea surface is wide enough for a 100m-long ship's maneuvering. Fig 2 shows location of JNFL facilities and MO-Port.

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Fig.2 JNFL Facilities and MO-Port

2.1 Transport Packagings(Casks) for Spent Fuel Spent fuel from nuclear power stations is contained in sturdy transport packagings that protect against radiation. These transport packagings(Fig.3) are made of steel and lead, are 6.3 meters long and 2.6meters in diameter, and weigh maximum 118 tons(2). They are designed to prevent leakage of radioactivity even during fires, after impact with other objects, or in the very unlikely event that the casks are dropped(3). They meet both International Atomic Energy Agency (IAEA) and Japanese standards and have passed the strict safety inspections of the Science and Technology Agency. The weight of the spent fuel carried in a transport packaging is roughly seven tons in maximum.

Fig.3 NFT Type Cask

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Dimension and Weight of NFT Type Cask Type of Cask NFT 10P NFT I4P

Dimension Limits 2.6m°x6.2mL 2.6m°X6.3mL

Weight L i m i t s 75 ton 115 ton

NTT 12B

2.3mDX6.4m0

73 ton

NFT 22B

2.6m°X6.3m"

97 ton

NFT 32B

2.4m°X6.4m

L

106 ton

NTT 38 B

2.6m°x6.4m L

118 ton

2.2 Main Particulars of Rokuei-maru Most Japanese nuclear power stations have their own ports. Depending on the port location, there are various restrictions on the type of ship which may call at the port. Factors include size and facilities of the port, weather and sea conditions, social requirements from local authorities and so on. In particular, front sea surface and water depth at the port limit the size of ships entering directly. This is the reason why "Rokuei-maru(Fig.4)" is approximately 100m in length with a maximum draft of 5.6m(4). 2.2.1 Ship Structure Considering unforeseen events such as collision and stranding, the ship hull and bottom have a duplex construction so that even if the ship were damaged, water would not penetrate the inside wall and the ship would withstand sinking. 2.2.2 Tie Down A strong tie-down device is provided to prevent the spent fuel package from shifting or falling down even if the ship is subjected to severe external force. 2.2.3 Fire Extinguishing Equipment The ships are equipped with fire fighting equipment in excess of INF Code requirements. Although spent fuel package is not practically inflammable, the holds are fitted with water flooding system in consideration of the worst credible accident scenario. 2.2.4 Spare Electric Generators Spare electric generators are additionally provided to prepare for accidental loss of power. 2.2.5 Automatic Collision Prevention Support Gear To avoid collision and stranding the ship is provided with safe navigation devices such as automatic collision prevention support gear and echo sounding equipment. Dimension Deadweight Capacity

88

Length overall Breadth

Approx. 100 meters Approx. 16.5 meters Approx.3.000 tons Spent fuel package 20 units, max.

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Fig.4 Rokuei-maru 2.3 150-Ton Wharf Crane A 150 ton wharf crane(Fig.6) at Mutsu-Ogawara Port unloads the SF packages. The crane boasts a number of unique features. 2.3.1 Inverter Control System The inverter-control system offers a wide range of speeds and provides smooth acceleration and deceleration, guaranteeing ease of operation. 2.3.2 Collision/Overload Prevention System Each crane is equipped with sensors that prevent collisions when it is on the move by detecting the proximity of the other crane. The cranes are also fitted with devices that halt operations in the event that the load is too heavy. 2.3.3 Backup Motors/Power Supplies In the unlikely event that the primary motor breaks down, the backup motor can raise or lower the suspended cargo, or move it laterally. In addition, an emergency power supply can safely lower a suspended load in the event of a power failure. 2.3.4 Drop-Prevention System If a load is left suspended by a power outage, double winding cables and a motor brake are activated to prevent it from dropping. 2.3.5 Curved Track System In order to maximize efficient use of the quay, when the cranes are not in use they can be moved to the port apron along a curved track, rotating 90°in the process. Their ability to travel on a curve in this way makes them the most unusual cranes in the world.

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Fig.6 150-ton Wharf Crane 2.4 Transport Vehicle for Spent Fuel Spent fuel is transported from Mutsu-Ogawara Port to the reprocessing plant of JNFL on specially designed overland carriers owned by NFT. Packages containing spent fuel unloaded at Mutsu-Ogawara port are placed on board specially designed overland carriers(Fig.7). These then travel approximately seven kilometers by private road to the reprocessing facility. The carriers are self-powered with no inter-vehicle links. They have 48 wheels to minimize the danger of skidding during snowy winters. In addition to ordinary braking systems, they have safety brakes to stop the vehicle reliably in emergencies. They also have rear obstruction detectors and rear-view TV cameras to allow the operator to check the vehicle's rear status when reversing. To enhance operational safety even further, the vehicles have been designed to carry ultra-heavy transport package loads of over 100 tons.

Fig.7 Transport Vehicle

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3 FALSIFIED DATA ON NEUTRON SHIELD MATERIALS USED IN NFT SPENT FUEL TRANSPORT CASKS In October 1998, it was discovered that employees of Genden Engineering Services and Construction Co. (Genden) had falsified data on neutron shield materials, which it procures and processes for use in NFT spent fuel transport casks. NFT set up an in-house task force to look into the facts of the matter. At the national level, the Investigation Committee on Spent Fuel Transport Casks Issue also initiated investigations with a view to clarifying the causes. The Committee issued its final report in December. The inquiry discovered that data on neutron shield materials had been doctored in connection with 39 out of a total of 43 casks of two types.. .31 NFT-type casks to be used by NFT for transporting spent fuel to Rokkasho, and 12 transport casks for use within power plants. The Committee also gave an account of the background to the data falsification, safety evaluations for cask shielding and measures to prevent the recurrence of such an incident. The data falsification occurred because of counter-measures taken by the fabricator of the neutron shield to rectify the delayed fabrication schedule. These counter-measures were of course inappropriate. Although the neutron shield material data was falsified, the Nuclear Safety Special Committee of the Science and Technology Agency (STA) confirmed that safety measures were clearly secured in the performance of the casks concerned. However, the scandal of the data falsification seriously damaged public trust in the nuclear industry. In December, the head of the Science and Technology Agency's (STA) Nuclear Safety Bureau directed NFT to rigorously exercise its control responsibilities as the issuer of the order for the transport cask. The STA also requested the return of the packaging certificate and ordered re-inspection of the casks. NFT has established Safety and Quality Assurance Department and, in conjunction with related and affiliated companies, it will make every effort to prevent a recurrence of this problem. In view of the seriousness of this problem NFT has also made appropriate personnel adjustments at the Directorate level. As stated before, NFT has established Safety and Quality Assurance Department, and directly and periodically checks the quality control in the fabricators and their sub-vendors of casks. As proof of our high level of quality assurance, NFT obtained the ISO9001 Certification in 1999 within only one year of the occurrence of the problem. During the transport of spent fuel for test purposes on October 2, it was confirmed that the transported materials' radiation dose equivalent rates and other measurements were well within the limits. In order to prevent recurrences of similar incidents, NFT established a "Charter for Corporate Activities", and has continuously developed an education program for employees to improve their levels.

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4 PUBLIC ACCEPTANCE ACTIVITIES NFT operations are carried out in the public domain (public roads, the open sea, and ports). This situation is very different from stationary power plants, because transport operation takes place in the vicinity of the general public. Transport regulations are formulated to ensure the safety of the public and environment. To the extent, therefore, that these regulations are fully observed, assurance of safety for transporting nuclear materials should be achieved. Transport safety will be further enhanced if the understanding and cooperation of the general public is secured. For this purpose, it is vital we take measures to obtain trust and support from the public. Appropriate measures to achieve such goals are as follows: 4.1 Disclosure of information to the general public. 4.2 Exchange of information by dialogue. As regards (4.1) this is performed by the competent authority, the Science and Technology Agency. Safety Analysis Reports of packages are open to the public at Public Information Centers. Protection of commercial secrecy is taken into consideration in releasing the information. As regards (4.2) this is performed between NFT, the local government and residents where the transport takes place from the port to the JNFL facilities. 4.3 Participation in NS Net On September 30, 1999, a Criticality Accident occurred at JCO-Tokai Plant. In order to prevent a recurrence of a similar serious accident, the Nuclear Safety Network was established with NFT among the participating companies. Comprising 35 corporations, including power companies, manufacturers, etc., and research institutions all involved in Japan's nuclear fuel cycle, this organization seeks to raise awareness of safety issues throughout Japan's nuclear power industry, improve ethics, and promote a common culture of nuclear power safety. Its primary activities are as follows: • Encourage the spread of a culture of safety in nuclear power. • Implement peer reviews among members. • Encourage the exchange and announcement of information concerning nuclear power safety.

5 CHALLENGES IN THE NEAR FUTURE In order to achieve economical and efficient transportation while securing improved levels of quality, safety and security under the circumstances of increasing total transport volume, Japan has identified the following challenges:

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5.1 Development of New Types of Transport Packagings and Baskets with Superior Characteristics: Transportation costs become more economical in proportion to the numbers of fuel assemblies loaded in each transport packaging. In this sense, the development of a new type of transport packaging, which is superior in subcriticality, heat removal and fire resistance characteristics, is essential. In addition to the above, the development of new types of fuel baskets, which are superior in heat conduction characteristics in terms both of material and construction, is also required to cope with the recent development of higher bum-up fuel. 5.2 Transportation under Public Acceptance : Nuclear fuel materials are transported not only within the battery limit of nuclear facilities, but also on public roads and on the open seas. In order to achieve safety and security in transportation as well as to minimize opposition and/or protests by the residents concerned, it is becoming more important for nuclear industries involved in the transportation of nuclear fuel materials to disclose sufficient and appropriate information. Under these circumstances, the World Nuclear Transport Institute (WNTI) was established in 1999 with the aim of developing Public Acceptance activities internationally. As a member company of the Institute, we, NFT, fully support the WNTI in order to smoothly implement positive activities in the near future. 5.3 Establishment of Enhanced Emergency Response Plan: As a lesson from the JCO Criticality Accident, NFT has reviewed the existing Emergency Response Plan. It is taking necessary actions to establish a substantial Emergency Response System, such as creating more realistic scenarios for emergencies and emergency exercises. It is also preparing the most advanced operation room and equipment to cope with any emergency, 24 hours a day. 5.4

Development of Transport Methods for Radioactive Wastes Generated through the Decommissioning of Shut Down Nuclear Power Plants For future business, transport methods need to be considered for radioactive wastes which emerge from the dismantling of nuclear reactors coming to the end of their working life.

6 CONCLUSION Once the reprocessing plant of JNFL enters into full operation, one ship will be insufficient. Electric power utilities have started to study the possibility of constructing another SF transport ship. At present, "Rokuei-maru" has not been evaluated according to the actual performance of SF transport. However the result of her sea trials were fairly good and she has won the respect of her crew members. In Japan, spent fuel has been transported from nuclear power plants to PNC's Tokai Reprocessing Plant for the last 19 years without a single incident. Based on this experience, the future transport of spent fuel to JNFL's Rokkasho Reprocessing Plant is expected to take

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place safely in cooperation with local governments and relevant entities and will contribute to the smooth operation of the reprocessing plant and the nuclear fuel cycle in general. In order to achieve the target for the reduction of greenhouse gas emissions as agreed at COP3 in Kyoto, it is inevitable that nuclear power will form a certain part of the total power supply. It is correspondingly vital for the nuclear fuel cycle operation to be supported by a secure transport system. For the success for our operation, we have strictly observed national and international regulations while establishing careful measures for transport. We have developed public acceptance activities to make local residents understand and accept our activities. With increasing international shipments of vitrified high-level waste and MOX fuel and with the commencement of regular domestic transport of spent fuel, it is very important to further develop public acceptance campaigns on a global scale. We would like to express our high hopes for the activities of the new World Nuclear Transport Institute. We will fully support the activities of the Institute, while further extending our record of safe transport of nuclear materials.

REFERENCES (1) Domestic Transport for the Nuclear Fuel Cycle in Japan, K.Nambu and K.Nakama, Proc 12th International Symposium on Packaging and Transportation of Radioactive Materials, 1998, pp. 1453-1460 (2) Development of NFT Type Spent Fuel Transport Cask, S.Shimura, Int J. Radioactive Material Transport,. Vol.8, Nos 3-4, pp.257-264 (3) Demonstration Test for a Shipping Cask Transporting High Burn up Spent Fuels-Thermal Tests and Analysis, H.Yamakawa et al Proc 12th International Symposium on Packaging and Transportation of Radioactive Materials, 1998, pp.659-666 (4) Basic Planning of Newly Built Exclusive Ship for Spent Fuel Transport, I.Obara et al, Proc 12th International Symposium on Packaging and Transportation of Radioactive Materials, 1998, pp.745-752

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C584/032/2000 Engineering considerations associated with plant used for storage of intermediate level waste - a regulator's view W SEDDON Nuclear Safety Directorate, Health and Safety Executive, Bootle, UK

ABSTRACT Everyone associated with the nuclear industry will know that the provision of a deep repository has receded by at least several decades. The question now is "How does the nuclear industry engineer a suitable storage regime for the foreseeable future?" This paper explores some of the engineering considerations that NII Specialist Inspectors need to address when considering the safety case for a long-term radioactive materials store. Any Licensee contemplating building a store should already know of such items. Integrity of waste form, storage container, handling equipment and store structure is addressed. Licensee's knowledge management is also considered.

1.

INTRODUCTION

Radioactive waste materials are stored on all nuclear licensed sites, and following the Rock Characterisation Facility decision, time-scales for this storage may well extend into the 22nd century at some sites. This paper is intended to summarise current thinking on the policy for long term storage of intermediate level radioactive waste materials on licensed sites. It is also intended to show some of Nil's thinking on the associated regulatory strategy and this is expanded into the specialist inspector area of engineering assessment associated with current waste storage proposals. It does not consider short-term storage activities during receipt, processing, stock holding on behalf of committed customers or dispatch of finished goods or materials. Some attempt is made to show that there will be differences between storage requirements for the various types of Intermediate Level Waste (ILW). Long-term considerations will apply to all forms of storage when sufficient time elapses such that physical or chemical degradation of the radioactive materials, of their containers, of the lifting equipment associated with the containers or of the storage structure, creates a significant hazard.

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There will always be a requirement for storage of 1LW on a nuclear site to be managed by the licensee and so preservation of the licensee's knowledge, control and competence should also be considered. Public confidence in the storage arrangements is of key importance and the actions and requirements of licensees, politicians and regulators will affect this aspect of prolonged storage.

2.

RISK AND HAZARD

During its routine operating life a nuclear plant, is regulated on the basis of enforcement being proportionate to RISK. This has been defined by HSE as " the chance that something adverse will happen" ( TOR Para. 11 and Appendix 1 (1)). It is usually taken as being the probability of occurrence of an event and consideration of the consequences of that event. Certainly one part of the NII document, Safety Assessment Principles for Nuclear Plants (2), covers the need for probabilistic safety analysis within a safety case for each nuclear plant. What may not be clear from that document is that NII places greater emphasis on the engineering principles that occupy approximately 75% of the SAPs. Arrangements to review the SAPs document have been made and will clarify this point. UK Nuclear Licence Conditions require the provision of a safety case, which among other things should demonstrate that all structures, systems and components with a significant safety function are adequate in engineering terms. It is Nil's view that any nuclear safety case must show how all reasonably foreseeable events have been determined and how the plant has been, or will be, engineered to reduce risk associated with the occurrence of these events, so far as is reasonably practicable. Hazard has now been defined by HSE in the document Reducing Risk Protecting People (R2P2) (3) as the potential to cause harm. Clarification has also been given in law that, when the Health and Safety at Work Act talks about "without harm to health" or "absence of risk to health or "exposure to risk to health", there is a requirement that hazards present are properly addressed. If the consequences are irreversible and deleterious there is an overriding need to introduce control measures for addressing the hazard even when there is considerable uncertainty about the nature of the hazard. This change of emphasis in the interpretation of UK Law is highly significant in considering the change from acute to chronic states. NII has found many limitations in attempting to regulate using risk in its probabilistic sense, such as when dealing with old plant, decaying organisations, "safe residues", wastes and decommissioning. For instance the extensive improvements now ongoing at UKAEA Dounraey and at BNFL, following last years team inspection, are not driven by risk based arguments. The requirements are based on comparison with modern standards of nuclear engineering and safety management. R2P2 also considers such issues as limitations of regulatory predictive power, capacity to convince the public about long term effects and public trust. This leads into national policy thinking.

3.

PUBLIC DEBATE

The main question here is how can nuclear wastes be managed safely when the time-scale over which they continue to have a harm potential can be extremely long, probably stretching

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through many generations? Traditionally the nuclear industry has favoured disposal with limited post closure control (the "dispose and walk away" option), where future generations are relieved of any obligation to manage the waste since it has been placed deep underground and sealed in by backfill. There appears to have been a change in thinking in the last few years, and there is now a reluctance to allow control of the waste to be relinquished in the short term. The concept of retrievable long-term storage has emerged, with the possibility that technological developments may lead to better solutions for final disposal. The concept of harm potential has already been mentioned. Part of the current debate is about what the harm potential of today's actions may be on future generations. These are not questions that can be addressed by the nuclear industry and/or regulators in isolation. However, Licensees do need to show that the chosen solutions for long term storage are well engineered and take account of known degradation mechanisms, such that both public and worker safety is maintained throughout storage life, and unauthorised discharges to the environment are avoided.

4.

GOVERNMENT POLICY

The primary responsibility for drafting civil radioactive waste management policy in the UK lies with the Department of the Environment, Transport and the Regions. The current policy was presented in 1995 in Cm2919 and a review is to be completed after the issue of a consultation paper this year. The policy is still for ultimate disposal via a deep repository, however, only time will tell if this will continue. Until such time as a repository becomes available, there remains the need for long-term storage of all forms of nuclear material, both for radioactive waste and for materials which are not declared as waste, but for which future use is not specified. Again Cm2919 makes mention of this with storage being achieved using "passive" means. Unfortunately there is no clear definition of what is meant by passive safe storage. Hence HSE and particularly NII has to look towards safety legislation, discussion with those currently designing, building and operating long term stores, and any guidance in the SAPs, in order to advise government, DETR and the nuclear industry on the current regulatory strategy and possible changes in strategy for such long term storage.

5.

REGULATORY STRATEGY

In generating a strategy for the regulation of radioactive waste management, NII aims to take due notice of best international practice. There is a series of technical reports on radioactive waste management that have been produced by the IAEA, with European Law having an increasing impact on all that is proposed. NII fundamental expectations are consistent with the principles published by such sources. These expectations are: (i)

Production of radioactive waste should be avoided and if that is not practicable then minimised

(ii)

Radioactive waste should be managed and stored safely in a responsible manner

(iii)

Full use should be made of existing routes for disposal of radioactive waste

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(iv)

Remaining radioactive material should be processed promptly for passive safety and stored pending future disposal or other long-term solutions.

6.

ENGINEERING CONSIDERATIONS

The main thrust of this paper is intended to come from a few of the phrases already mentioned, namely: A safety case for each nuclear plant Control of the waste in the short term The concept of retrievable long term storage Physical or chemical degradation of the radioactive material Physical or chemical degradation of the radioactive material container Physical or chemical degradation of the in-store handling equipment Maintenance of a suitable storage environment Ageing of the storage structure Preservation of the Licensee's knowledge, control and competence 6.1 Safety Cases The need for a safety case has already been raised. It is a requirement of each Nuclear Site Licence granted in the UK by HM Chief Inspector of Nuclear Installations. Each licence has a common schedule of 36 Licence Conditions appended to it. Within these LC Number 14 on Safety Documentation requires arrangements to be in place for the production of safety cases "to justify safety during the design, construction, manufacture, commissioning, operation and decommissioning phases of the installation". Additionally LC Number 23 requires that a Safety Case is in place for an operational nuclear plant. Other Licence Conditions will be mentioned later. 6.2 Control of the waste in the short term The safety case must demonstrate that, in moving from a mobile waste form to a stable waste form held within an appropriate primary containment, there is control and containment of the radioactive material. Similarly when the contained material is positioned in a suitable store, there is a Licence Condition requirement (LC34) for "the licensee to ensure, so far as is reasonably practicable, that radioactive material and radioactive waste on the site is at all times adequately controlled or contained, so that it cannot leak or otherwise escape from such control or containment. Additionally there is a requirement that "no such leak can occur without being detected" and having been detected any leakage then has to be "notified, recorded, investigated and reported".

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6.3 The concept of retrievable long term storage Inevitably when retrieval has to be considered after a considerable time of storage, there will be a need to demonstrate that the storage regime will permit this. Even before the store is used a safety case should be presented to show that the environment within it, coupled with the suitability of primary container will be conducive to maintaining waste form integrity or at least prevents leakage from the container for the stated lifetime of the plant. Additionally effective storage should be such that the container will remain of sufficient integrity such that it matches conditions for acceptance in the next storage place, whether that is in a further long-term store or in a repository. 6.4 Physical or chemical degradation of the radioactive materials ILW can range from heat generating radioactive materials through to wastes that are only slightly above activity levels that would permit placement in low level waste stores. The overarching requirement is to immobilise the waste within a suitable container. The specifications that exist for immobilising the materials require that the waste form retains its integrity for a defined period, which may be many decades. For some of the heat generating wastes this is likely to result in the need for in-store cooling, thus preventing or controlling corrosion within the waste form. Again the safety case for the storage facility should clearly state if there is such a need. It must also demonstrate that there has been sufficient research and development to show that the possible extent of corrosion is understood, that suitable operating parameters have been derived, and that there is an adequacy of engineering provision to adequately control these parameters. By so doing the safety case will allow justification to be made in respect of satisfying Nil's Engineering SAPs. It is particularly important here to recognise the need for validation of analytical models 'against experiments which replicate as closely as possible the expected plant conditions (SAPs Principle 87). It is unlikely that experiments or development trials will be able to replicate the range of anticipated plant conditions. For example there will never be sufficient lead in times to run representative ageing tests. SAP Principle 88 would then come into the frame: 'The data used in the design and fault analysis of safety related aspects of the plant performance should be shown to be valid for the circumstances by reference to established physical data, experiment or other appropriate means'. In their engineering assessments NII Specialist Inspectors take this to mean that claims against extrapolated data need to be carefully examined to seek adequacy of justification. It may well be, that in the case of a store, the environment is controlled so that degradation is minimised. However, the specialist inspector is likely to look for assurance that there is monitoring of both the store environment and the stored items most at risk. 6.5 Physical or chemical degradation of the radioactive waste container The previous section noted the possibility of the waste form being degraded by corrosion. This invariably leads to volumetric growth of the waste form and, where it is a close fit within the primary container, problems will ensue. The primary container in this case will thus be subjected to internal pressure that could ultimately result in failure of the containment. Additionally the environment within the store will be important as has previously been mentioned in connection with retrievability. It is possible to envisage condensation inside a store where no heat is generated, or little air movement takes place. The possibility of stagnant water being in contact with the containers, coupled with the air in the store possibly being salt laden, as many stores are likely to be near the sea, could lead to a recipe for

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corrosion leading to failure of the lifting integrity even of stainless steel containers. What will this do for the integrity of the waste package? The safety case will need to explicitly deal with such matters. 6.6 Physical degradation of the in store handling equipment The possibility of degradation of handling equipment needs to be considered during the design and safety case preparation stage for a store. The use of the term handling equipment in this paper is intended to cover both lifting equipment and any lifting accessories. It may be possible to remove the crane used during store filling from the storage environment, both when it is between store loading campaigns and at end of filling. This will assist in being able to comply with the thorough inspection requirements of the Lifting Operations and Lifting Equipment Regulations 1998. However, it is also necessary to consider any stillage that may be used for transfer and storing of containers. The Approved Code of Practice, Safe Use of Lifting Equipment (4) gives guidance on this: Para. 244 'You should ensure that any lifting accessories used for securing the load are compatible with the load, taking into account any attachment points on the load, the environmental conditions in which the accessories will be used and their configuration of use'. Para. 245 'You should ensure that appropriate measures are taken to prevent the load, or part of the load disintegrating while being lifted'. It is important to consider this from the perspective of harm potential as mentioned previously. Any store containing radioactive materials packaged for external transportation will be accessible to workers, hence degradation can be noted by inspection and the harm potential from load disintegration causing a dropped load can be avoided by implementing suitable measures. In a non accessible store the harm potential comes from any radiation dose received by workers during the retrieval of failed lifting equipment and accessories, and during the recovery of waste containers, failed or otherwise. Suitable and sufficient risk assessment and inspection regimes will be necessary. 6.7 Maintenance of a suitable storage environment Most of the sections already covered have mentioned the need for some form of control of the storage environment. Each safety case for an ILW store will have to demonstrate that an appropriate environment will be created consistent with the needs for control of waste form corrosion, container integrity and handling equipment integrity. Inevitably there will be a need once the store is in place to demonstrate that the storage environment is being maintained to the safety case specified conditions, both by initial commissioning tests and by suitable operating rules. 6.8 Ageing of the storage structure Throughout this paper there has been a common thread of the need to consider at the outset the long-term effects of extended storage of wastes. How will this impact on determining the life of each store? The nuclear industry is little more than 50 years old, yet there are already unsuitable buildings being used for storage of waste that require replacement. If current needs are already showing that stores are needed into the 22 nd century, can this be achieved using the building techniques that are now available? At best there can be a qualified maybe. Although there are buildings around that have survived for centuries would it be appropriate to store radioactive materials in them? The current type of building being considered for ILW stores is usually reinforced concrete or block with steel framed clad upper structure and roof

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For such structures to continue to provide containment there will be a need to have an ongoing inspection, maintenance and refurbishment regime and to achieve this safety cases should now consider how containment may be maintained whilst refurbishment is in progress. If novel approaches are to be considered for store structures, there has to be a demonstration that the effects of ageing have been considered. An example of this may be consideration of the use of stainless steel as a building cladding. With only approximately 50 years experience of stainless steel, how can any claim be made say for 100 years cladding life. An inspection regime with expectation of major refurbishment is likely to be the only way to show that such an idea is appropriate. 6.9 Preservation of the Licensee's knowledge, control and competence It is possible to draw out from the preceding text a number of items that have to be retained within the Licensee's knowledge: Licensees do need to show, by provision of a safety case that the chosen solutions for long term storage are well engineered and take account of known degradation mechanisms, such that both public and worker safety is maintained throughout storage life. This case will be required to undergo periodic review, hence it will require to be covered by a document control system throughout the life of any stores. The safety case needs "to justify safety during the design, construction, manufacture, commissioning, operation and decommissioning phases of the installation" to show that the store's environment, coupled with a suitable primary container will be conducive to maintaining waste form integrity. It also must demonstrate that, as a waste is conditioned from a mobile form to a stable form held within an appropriate primary containment, there is control and containment of the radioactive material. There is also a requirement that "no such leak can occur without being detected" and having been detected any leakage then has to be "notified, recorded, investigated and reported". This leads to a requirement for monitoring of the store environment by the Licensee and the provision of records. The safety case for the storage facility should clearly state if there is a need for instore cooling, thus preventing or controlling corrosion within the waste form. It may well be that in the case of a store the environment is controlled so that degradation is minimised. However, there must be assurance that there is a monitoring regime for both the store environment and the stored items most at risk. Records of such monitoring will be required to comply with Licence Conditions. For the safety case to explicitly deal with matters such as degradation of the radioactive waste container, there will be a need to monitor the containers for signs of degradation. Typically continuing measurement of the circumference of an appropriate sample of containers could be necessary. Records of all such measurements need to be kept and in order to be of use the individual containers need to be given unique identifiers that will remain discernible throughout long-term storage. Implicit in interpretation of any dimensional change is provision of competent persons with knowledge of acceptance standards provided within the safety case.

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Degradation of handling equipment needs to be considered in order to comply with the thorough inspection requirements of the Lifting Operations and Lifting Equipment Regulations 1998. The degradation can be noted by inspection and the harm potential from load disintegration causing a dropped load can be avoided by implementation of a suitable inspection regime determined by a competent person appointed by the Licensee. There is a requirement to keep records of such inspections. There is also a need to assess the continuing fitness for purpose against any changing requirements. There has been mention of the need for some form of control of the storage environment. Hence there will be a need once the store is in place to demonstrate that the storage environment is being maintained to the safety case specified conditions. Retention of monitoring records will generally satisfy this requirement, but there will be a need for a competent person to periodically review such records to confirm that the storage environment remains adequate to minimise waste form and container degradation. For store structures to continue to provide containment there will be a need to have an ongoing inspection, maintenance and refurbishment regime. Again the consideration of such inspections will be covered in periodic reviews. All of the above should be covered in a part of the safety case that deals with the arrangements that are put in place, at the inception of an 1LW storage project, to manage safety.

7.

CONCLUSION

This paper has explored a number of the engineering considerations that NII Specialist Inspectors will address when considering the safety case for a long-term radioactive materials store. Any Licensee contemplating building a store should already know of such items. Part of the function of each and every Directorate in the Health and Safety Executives is to give advice when requested. Hence many persons present will know of ongoing discussions on provision of appropriate storage facilities. Public opinion and debate will no doubt determine some of the options that are followed, however it is up to the industry to show to Government, regulators and the public that it has provided stores that are safe so far as is reasonably practicable.

8.

ACKNOWLEDGEMENT

The author wishes to thanks fellow NII Principle Inspectors Dr D Turton and Eur Ing I McNair for providing text from which the early sections of this paper was prepared. The author also wishes to thank HM Chief Inspector of Nuclear Installations/ Director of Nuclear Safety Directorate of HSE, for permission to publish this paper. The views expressed are those of the author and do not necessarily represent those of the inspectorate/HSE.

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REFERENCES 1. Tolerability of Risk from Nuclear Power Stations , Health and Safety Executive, 1992 ISBN 0-11-886368-1 2. Safety Assessment Principles for Nuclear Plant, Health and Safety Executive, 1992, ISBN 0-11-882043-5 2. Reducing Risk Protecting People (R2P2), A Health and Safety Executive Discussion Document, 1999, DDE 11 C150 5/99 4. The Approved Code of Practice, Safe Use of Lifting Equipment, Health and Safety Commission, 1998, ISBN 0-7176-1628-2

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C584/018/2000 The packaging of waste for safe long-term management S V BARLOW and J D PALMER United Kingdom Nirex Limited, Harwell, UK

SYNOPSIS/ABSTRACT The role of Nirex is to provide the United Kingdom with safe and environmentally sound options for the long-term management of radioactive waste generated by the UK's commercial, medical, research and defence activities. This includes all intermediate level waste and some low level waste (ILW and LLW). One of the key objectives of Nirex over the past 10 years or so, has been to ensure that when waste is packaged, it is in a form suitable for its future safe management including storage, transport, handling and potential disposal. To provide a basis for developing waste packages that are compatible with future waste management, Nirex has developed standards and performance specifications for waste packages that include wasteform and container design, quality assurance and data recording requirements. In addition to the specifications, Nirex also provides detailed advice on the suitability of specific packaging proposals and plant designs against the foreseen requirements for future transport, handling, storage and potential disposal. Where packaging proposals meet these requirements, Nirex is prepared to endorse the proposed approach through the issue of a 'Letter of Comfort'. This approach has enabled the commencement of waste packaging operations with a high degree of confidence that the waste product will meet future waste management requirements, including potential disposal requirements. This paper provides a summary of the standards and specifications developed by Nirex for waste packages, and of the assessment process applied by Nirex in providing advice and endorsement of specific packaging proposals.

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1.

INTRODUCTION

The role of Nirex is to provide the United Kingdom with safe and environmentally sound options for the long-term management of radioactive waste generated by the UK's commercial, medical, research and defence activities. This includes all intermediate level waste and some low level waste (ILW and LLW). One of the key objectives of Nirex over the past 10 years or so, has been to ensure that when waste is packaged by waste producers, it is in a form which is suitable for its future safe management including storage, transport, handling and potential disposal. To provide a basis for developing waste packages that are compatible with future phases of waste management, Nirex has developed standards and performance specifications for waste packages that include wasteform and container design, quality assurance and data recording requirements. The Waste Package Specification (WPS) has been, and is being, used by waste producers in the development of their waste management strategies. It has been derived using knowledge from national and international regulations, wide-ranging research programmes, and experience gained from developing concepts, designs and safety assessments for facilities for the disposal of the UK's intermediate level and some low level wastes. Where the knowledge was developed from site-specific studies and evaluations, the specifications are based on a generic interpretation of handling, transport and potential disposal requirements. This paper describes the standards and Specifications established by Nirex and the manner by which Nirex assists the industry to produce waste packages in a form suitable for safe longterm management including storage, transport, handling and potential disposal.

2.

DEFINITION OF STANDARD CONTAINERS

Based on the information available to Nirex from the National Radioactive Waste Inventory, Nirex has been able to determine the range of containers required to meet the needs of the UK waste producers. Standardisation is important as it enables handling and transport operations to be optimised around a limited number of variants, with consequent benefits in safety, logistics and cost throughout all phases of waste management. The standard range comprises four standard waste containers for ILW and two containers for LLW. The Nirex interest in LLW is restricted to that fraction which is not suitable for existing disposal facilities. The six standard containers currently defined by Nirex for packaging ILW and LLW are listed in Table 1.

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Table 1: Nirex Standard Containers Waste Container

Typical Contents

Overall Dimensions

.Gross1 Mass

500 litre Drum

The normal container for most operational ILW

800 mm diameter x 1200 mm height

2,000 kg

3m3 Box

A larger container for solid wastes

1720 mm x 1720mm plan x 1225 mm height

12,000kg

3m3 Drum

A larger container for in-drum mixing and solidification of liquid and sludge type wastes

1720 mm diameter x 1225 mm height

12,000kg

4 metre Box

For large items of waste, especially from decommissioning operations

4013 mm x 2438 mm plan x 2200 mm high

65,000 kg

4 metre LLW Box

For LLW

4013 mm x 2438 mm plan x 2200 mm high

30,000 kg

2 metre LLW Box

For LLW

1960 mm x 243 8 mm plan x 2200 mm high

30,000 kg

When a waste container is filled with conditioned waste, the complete assembly is denoted a waste package. Nirex has defined two generic types of waste package: "

unshielded packages, which owing either to radiation levels or containment requirements, require remote handling and must be transported in a reusable shielded transport container; shielded packages, which have built-in shielding (where appropriate) and contain lowactivity materials, so that the packages can be handled using conventional techniques and are transport packages in their own right.

•

2.1.

Standard Packages

2.1.1. Unshielded packages The 500 litre Drum, 3m3 Box and 3m3 Drum are typically manufactured from stainless steel. They are not designed to provide any radiation shielding in themselves hence when filled with waste these are known as unshielded packages. Handling and storage of these packages require remote handling facilities. For transport through the public domain, the unshielded waste packages will generally be placed within a shielded reusable transport container to form a Type B package as defined in the Transport Regulations of the IAEA [3, 4]. The shielded reusable transport containers are designed to contain either four 500 litre Drums within a transport stillage, a single 3m3 Box or a single 3m3 Drum.

1

maximum mass when filled with conditioned waste

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2.1.2. Shielded packages The 4 metre Box is designed to meet the requirements for an Industrial Package freight container as specified in the IAEA Transport Regulations. The box is manufactured from stainless steel and is provided with a concrete lining that can be varied in thickness to suit the radioactivity of the contents. The radioactivity content is restricted to that which can be classed as Low Specific Activity (LSA) or Surface Contaminated Objects (SCO) [as defined in 3, 4] at the time of transport in the public domain. No additional transport container is required for the 4 metre Box because shielding is provided by the package itself in the form of the concrete liner. The box when filled with waste is described as a shielded package since it can be handled by conventional means; the dose rate external to the box is restricted consistent with its classification as an Industrial Package. Similar to the ILW box, the 4m and 2m LLW boxes are also designed to meet the requirements for an Industrial Package freight container. The difference between the ILW and LLW versions being that internal shielding will not be required for LLW material on account of the low activity and subsequent low external dose rate. The 2m LLW box is a halflength version suited for dense materials i.e. concrete or steel, that would otherwise cause the filled box to exceed its weight rating of 30 tonnes. The LLW boxes are similar to the containers currently in use for the disposal of LLW to the Drigg site in Cumbria. 2.2.

Non-standard Packages

A limited number of non-standard packages will also require transport, handling and potential disposal. These non-standard packages are specifically designed by the waste producers, for wastes that cannot be packaged into standard waste containers. These designs may pre-date the establishment of Nirex standards or may be developed to fulfil a specialised role. Nonstandard packages will have to satisfy the safety requirements to an equivalent level to that provided by the standard packages.

A description of the standard and non-standard waste radioactive waste packages being considered by Nirex are given in Report N/012 [5].

3.

SPECIFICATION OF WASTE PACKAGES

The packaging standards for intermediate level waste are set-down and defined within the Waste Package Specification for Intermediate Level Waste [1]. The Specification is comprehensive and covers key aspects of the waste package including dimensions, handling and other features, performance requirements, wasteform characteristics, quality assurance and data recording requirements. The Waste Package Specification therefore addresses both the waste container and the conditioned waste within: the wasteform. The Waste Package Specification is not specific to a particular waste management option, but is derived from considerations of generic requirements for safe storage, transport, handling and potential disposal. The Waste Package Specification therefore facilitates the production of waste packages that are suitable for all phases of waste management.

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Following its production, a waste package may be expected to undergo some, or all of the following phases: • • • •

Interim storage, usually at the site of arising. Transport to a disposal or storage facility. Handling and emplacement at the storage or disposal facility. Disposal

Each phase will place its own demands on the waste package. These can be summarised as follows.

3.1.

Waste Management Phases

Interim storage Interim storage of packaged waste is the responsibility of the waste producer. During this phase, packages are maintained within appropriate storage arrangements to ensure compliance with site licence conditions and associated safety requirements. Packages are kept in a manner so that future transport, handling and potential disposal requirements will not be compromised. Transport Although the siting and design of future long-term waste management facilities have yet to be finalised, the requirements and design of the transport system needed to transfer packaged waste between stores or to any future facilities are well understood. The transport system, which may involve both road and rail transportation (and potentially sea transport should such a facility be located on an off-shore island), is required to be able to accommodate both unshielded and shielded waste package types. The transport system will need to meet all UK regulations for transport of radioactive materials, which like almost every other country are based on International Atomic Energy Agency (IAEA) Transport Regulations.

Handling and Emplacement Although the design of any future storage and/or disposal facility will not be completed until the preferred waste management option is identified, such a facility can however be described in terms of the following generic elements and operations: • •

• •

receipt of transport packages, which will either be the re-usable shielded transport container type or the shielded package type; handling of the transport packages at surface facilities, which will comprise facilities for receiving and checking of transport packages as well as management, administrative and other support functions; transfer of transport packages to underground facilities via shaft or drift tunnel (assuming an underground facility is being considered); in the case of packages transported inside re-usable shielded transport containers, the unloading of waste packages in a shielded inlet cell;

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•

emplacement of waste packages in shielded vaults, where operations will permit monitoring and retrievability as necessary.

Nirex has recently undertaken a "peer preview" exercise seeking public involvement in the development of proposals for monitoring and retrievability. A workshop to discuss feedback received and take forward the issue is planned.

Disposal Nirex recognises that deep geological disposal is only one possible solution for the long-term management of the UK's radioactive waste and is the option Nirex has most experience of to date. This concept is based on isolating packaged wastes in vaults excavated at depth in a stable geological environment. The concept makes use of engineered and natural barriers, working in conjunction, to achieve the necessary degree of long-term waste isolation and containment. This multiple-barrier approach to containment takes credit for the fact that wastes are immobilised in an appropriate matrix, the immobilised wastes are packaged in metallic or concrete containers and are emplaced in underground vaults. The vaults will be backfilled with a cementitious grout material at an appropriate stage before sealing and backfilling accesses from the surface. In addition to these engineered barriers, the host rock will provide additional natural barriers, both physically and chemically, to the transport of radioactive material back to the human environment. Following cessation of repository operations, it is expected that the waste packages will experience; • •

a period of care and maintenance under institutional control, prior to final closure; backfilling of the disposal vaults with a specially formulated cementitious grout mixture to provide chemical conditioning and sorption of key radionuclides.

The requirements for all the above phases have been incorporated when developing the Waste Package Specification.

3.2.

Waste Package Specification

The Waste Package Specification is the means by which Nirex defines each of the standard waste packages. The specification incorporates the requirements placed on waste packages from the potential waste management phases identified above and also takes account of relevant international and national legislation, regulatory advice and other guidance and industry best practice as appropriate. The specification is applicable to the waste package as a whole: that is the waste container plus its contents, the wasteform. The specification defines dimensions and key features of the container and sets down minimum performance standards for the complete waste package. Criteria are specified for activity content, dose rate, heat output, surface contamination, dimensions, lifting feature, mass, venting, integrity, properties of the wasteform, impact and fire performance, stackability and identification. In the case of the shielded packages (those that are also transport packages), additional criteria are specified for applicable Transport Regulations. The Waste Package Specification also defines the quality assurance controls that

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will be applied to waste package production and the data measurements and records that are required for each waste package. The Waste Package Specification will be subject to periodic review as scientific knowledge of package performance increases, or when the generic bounding envelopes which support the specifications become superseded by specific site and design information for the final facility. It is envisaged that the act of becoming site and design specific may lead to the removal of some of the conservatism from specification criteria, as margins arising from current uncertainties are removed from the bounding conditions.

4.

OBTAINING PACKAGING ADVICE FROM NIREX

4.1.

Advice and Letters of Comfort

Waste producers are encouraged to discuss their detailed waste packaging plans with Nirex at an early stage, in order to obtain independent advice on particular packaging proposals. Nirex is prepared to give advice on specific applications based on its knowledge of waste package behaviour and performance requirements under storage, transport, handling and disposal conditions and from its experience obtained during the research and development of systems for the transportation, handling and disposal of radioactive waste. This advice normally takes the form of one or more Letters of Advice, which are reports issued by Nirex following assessment of a waste packaging proposal. The advice identifies further information requirements, or may highlight issues that need further development before an assurance can be given. Nirex is also prepared to provide assurances that the proposed waste packages are consistent with the envisaged transport and disposal system. This assurance is provided in the form of an endorsement known as a Letter of Comfort. A Letter of Comfort may be sought for management purposes (for example, before making capital expenditure commitments) or for regulatory purposes before commencing active operations on a packaging plant. The Letter of Comfort system has been in existence for more than 10 years and was established in the 1980s when the industry first started to condition and package wastes in a form ready for ultimate disposal. Following the rejection of the Nirex proposal to develop a Rock Characterisation Facility at the Longlands Farm site in Cumbria in 1997, Nirex has worked closely with the regulators and the nuclear industry, to ensure that the packaging advice process meets the needs of all parties and provides an appropriate level of transparency to decision making. The advice and Letter of Comfort process is being widely used by all the major waste producers as a component part of their waste management strategies and as is described shortly, has been strengthened by the provision of more formalised procedures which are based upon the established good practice from the nuclear industry.

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4.2.

Waste Packaging Assessments

When a waste producer wishes to receive advice or is seeking a Letter of Comfort for particular waste packaging proposals, the waste producer will make a submission to Nirex providing details of the wastes, proposals for retrieval and conditioning, supporting information regarding research and development and providing assurances regarding the provision of data records and the application of quality assurance. The submission will be subject to a formal assessment in a number of areas using in-house experts and external specialists as appropriate. Assessment will cover such aspects as wasteform stability, container design, waste package performance under normal storage and under fire and impact accident conditions, transportation, quality assurance and data recording proposals.

4.3.

Provision of Packaging Endorsement

Nirex is prepared to provide formal endorsement of specific proposals only when it can be demonstrated that the proposed waste packages will be compatible with the Waste Package Specifications, the Nirex transport and disposal concept and Nirex Packaging Principles. The Packaging Principles have been set down by Nirex in order to facilitate a transparent understanding of the principles against which packaging proposals are judged. The principles are published in the Nirex report Packaging of Waste for Safe Long-term Waste Management [6]. In cases where Nirex decides that a waste packaging proposal is consistent with the Waste Package Specification, the transport and disposal concept and the principles outlined above, Nirex will provide the waste producer with a Letter of Comfort. This approach is illustrated in Figure 1. The arrangements for assessing waste packaging proposals, for the provision of advice and endorsement by the issue of Letter of Comfort, are subject to formal procedures within Nirex and are governed by a Waste Management Advisory Committee. The Waste Management Advisory Committee has been established to reflect the good practice on nuclear licensed sites and provides a mechanism for regulatory overview of waste packaging and safety related matters. A detailed account of the waste packaging assessment process and the operation of the Waste Management Advisory Committee has been presented to a recent conference [7].

5.

ASSESSMENT OF QUALITY ASSURANCE ARRANGEMENTS

It is Nirex policy and a regulatory requirement that quality assurance (QA) be applied to all activities related to the disposal of radioactive waste. A key component of safe disposal is the quality of the waste package. However, it is the waste producer and not the disposer who has responsibility for packaging wastes. Nirex as the potential eventual receiver of the waste product, therefore needs assurance from those packaging the waste that all aspects that could affect the quality of the product are carried out under an appropriate quality assurance regime.

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A quality assurance specification has been established as an integral part of the Waste Package Specification [1]. It requires that all activities that affect the relevant safety cases, including the packaging of wastes, are carried out under an appropriate Quality Management System. The purpose of the QA system is to provide the evidence to demonstrate that waste packages comply with a defined product specification, which can be used to inform any future waste management decision. The Quality Assurance Specification is mandatory and describes the QA programmes and plans, the Waste Product Specification and the system of verification required to be put in place by the waste packager. The quality management system is a key component of the strategy to minimise the risk that packages are found not to be acceptable for transport and disposal at some stage in the future. As currently implemented assurance that packaging is being conducted in accordance with the Waste Packaging Specification, is provided by independent assessment either conducted by Nirex or its agents or by virtue of Third-party Certification issued by an appropriate Accreditation Body. Historically assessments have been focused on demonstrating that packagers have an appropriate quality system but now greater emphasis is being placed on the need to demonstrate that the waste packages being produced are consistent with the Letter of Comfort and basis upon which it was originally issued.

6.

OVERVIEW

Nirex provides assistance and advice to the UK nuclear industry in a number of areas. This includes compilation of the national radioactive waste inventory, by which Nirex is able to specify the range of standard containers that best meet the needs of the UK waste producers. Also by using its knowledge of waste package behaviour and performance requirements under storage, transport, handling and disposal conditions, Nirex is able to provide the industry with a generically based Waste Package Specification and detailed advice on specific waste packaging proposals. Where a proposal is consistent with Nirex specifications, the safety requirements for the transport and disposal concept and the Packaging Principles, endorsement is provided in the form of a Letter of Comfort. The Letter of Comfort signifies that the proposal has been subject to a systematic evaluation to confirm that it is consistent with foreseen requirements for interim storage, transport, handling and potential disposal. The issuance of a Letter of Comfort minimises the risk of inappropriate packaging and the subsequent need for future reworking or over-packing of waste packages. The advice and Letter of Comfort process is being widely used by all the major waste producers as a component part of their waste management strategies and has been strengthened by the provision of more formalised procedures which are based upon the established good practice from the nuclear industry. Nirex has developed and continues to maintain generic concepts for the transport, handling and potential disposal of wastes, based on the nature and volume of wastes defined by the national radioactive waste inventory that fall within the Nirex remit. Feedback between the packaging advice given at present and the development of future management concepts is a key synergy of the process. It is expected that the Waste Package Specifications and the Letter of Comfort process will continue to provide a firm basis for packaging wastes in the future, despite remaining uncertainties as to the final solution for the long-term management of the wastes. This

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approach has enabled the commencement of waste packaging operations with a high degree of confidence that the waste product will meet future waste management requirements, including potential disposal requirements. The continued application of an approved QMS, and the provision and retention of data records, is necessary to allow future generations to take waste management decisions with confidence, without incurring significant additional dose or financial burdens.

7.

REFERENCES

1.

United Kingdom Nirex Limited. Waste Package Specification for Intermediate Level Waste. Nirex Report N/007.

2.

United Kingdom Nirex Limited. 1998 UK radioactive Waste Inventory: Main Report. Nirex Report N3/99/01.

3.

International Atomic Energy Agency. Regulations for the Safe Transport of Radioactive Material. Safety Series No. 6, 1985 Edition (as amended 1990).

4.

International Atomic Energy Agency. Regulations for the Safe Transport of Radioactive Material. Safety Standards Series No. TS-R-1 (ST-1, Revised), 1996 Edition (Revised).

5.

United Kingdom Nirex Limited. Data Sheets for Standard and Non-standard Containers for Radioactive Waste. Nirex Report N/012.

6.

United Kingdom Nirex Limited. The Packaging of Waste for Safe Management. Nirex Report N/006.

7.

Packaging of Wastes: Nirex Assessment and Endorsement Process, J D Palmer and S J Wisbey. Proceedings of Containment 2000, International Conference on Nuclear materials - Containment and Disposal, 9-11 May 2000, Institution of Nuclear Engineers.

Long-Term

© 2000 UK Nirex Limited

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Figure 1: The Provision of Packaging Advice and Endorsement

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C584/014/2000 Independent monitoring of solid low-level radioactive waste disposals in the UK S NEWSTEAD and N A LEECH The Environmental Agency, Lancaster, UK S R DAISH Waste Quality Checking Laboratory, NNC Limited, Dorchester, UK

ABSTRACT The independent monitoring of solid low level radioactive waste (LLW) in the United Kingdom is undertaken by NNC Limited on behalf of The Environment Agency to ensure that disposals are within the authorised limits. Waste consignments are seized by the Agency prior to disposal and are transported to the Waste Quality Checking Laboratory (WQCL) at Winfrith, where the contents are analysed and assessed by destructive and non-destructive testing. The work performed at the laboratory is supported by a Quality Assurance System and specified tests are accredited by the United Kingdom Accreditation Service (UKAS). This paper outlines the regulatory framework for control of LLW disposals in the UK and describes the techniques used at WQCL for radioactive waste assessment.

1

INTRODUCTION

1.1 UK Legislation and Authorisations The disposal of radioactive waste to the environment is subject Radioactive Substances Act 1993 (RSA'93)(1). Although a relatively to consolidate an earlier one, the Radioactive Substances Act amendments introduced by subsequent legislation including Part Protection Act 1990 (3).

to the provisions of the recent Act its purpose was 1960 (RSA'60) (2) with V of the Environmental

Limits and conditions on the disposal of radioactive wastes are detailed in site specific Authorisation and Transfer Certificates. Over 900 premises in England and Wales are authorised. The majority of these consist of hospitals, universities and industrial research or manufacturing centres. The more significant radioactive discharges however are from a relatively small number of sites licensed under the Nuclear Installations Act 1965 (4). These are generally referred to as "nuclear sites" and are also authorised under RSA'93 to discharge radioactive wastes. These nuclear sites include nuclear fuel fabrication and reprocessing plants, nuclear power plants, atomic research establishments and isotope production centres. 1.2 Regulatory Authorities The Environment Agency (the Agency) is responsible for administration and enforcement of RSA'93 in England and Wales. Separate but similar arrangements exist in Scotland and Northern

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Ireland where the Scottish Environment Protection Agency (SEPA) and the Environment and Heritage Service are the respective regulatory authorities. The Agency has widespread responsibilities under environment legislation for management and regulation of the water environment, and for controlling industrial pollution and wastes including those from the nuclear industry. 1.3 Independent Monitoring Operators are required to determine and record the radioactive content of waste disposals in accordance with conditions specified in Authorisations. In support of its regulatory function the Agency's National Compliance Assessment Service (NCAS) provides a range of monitoring support services that includes managing a comprehensive programme of independent radioactivity monitoring. This programme allows checks to be made on the monitoring and discharge data provided by the site operators to the Agency's Radioactive Substances Regulation (RSR) site Inspectors and also provides data to support independent assessment of the exposure of the public to radioactivity from non-food pathways. These results are published annually by the Agency (5). As part of this programme consignments of solid low level radioactive waste (LLW) are sent to the Agency's Waste Quality Checking Laboratory (WQCL) for independent monitoring. This paper focuses on this process and the subsequent checking procedure and describes how the Agency uses the results to obtain added assurance that disposals and transfers are in compliance with Authorisations.

2

WASTE DISPOSAL

2.1 Low Level Radioactive Waste Solid radioactive waste is classified under broad categories, according to its heat-generating capacity and activity content. Low level waste is waste containing radioactive materials other than those acceptable for disposal with ordinary refuse and with activity contents not exceeding 4 GBq/t of alpha emitting radionuclides or 12 GBq/t of beta/gamma emitting radionuclides. The largest volumes of solid LLW originate from the following sources: • Nuclear fuel cycle plants operated by British Nuclear Fuels plc (BNFL) • Magnox nuclear power stations operated by BNFL Magnox Generation • AGR and PWR power stations operated by British Energy plc • Research establishments of the UK Atomic Energy Authority • Ministry of Defence facilities

Several landfill sites receive solid LLW and very low level wastes (VLLW) for controlled burial. These are wastes which can be disposed of safely with special precautions. The landfill site to be used, radioactivity limits and burial arrangements are specified in Authorisations issued to the waste producer.

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2.2 Drigg The primary disposal route for solid LLW however, is to the repository operated by BNFL at Drigg in West Cumbria about 6km south-east of BNFL's reprocessing facility at Sellafield. The site started operations in 1959 and receives waste mainly from Sellafield but also from other nuclear and non-nuclear radioactive waste producers elsewhere in the UK. The site occupies about 120 hectares (300 acres) close to the Cumbrian coast. Wastes were historically deposited by tumble tipping into trenches cut into clay to a depth of about 8 metres. This method of disposal ceased in 1994. Suitable wastes are now compacted and placed into half height ISO-containers at the Waste Monitoring and Compaction (WAMAC) facility at Sellafield. After transport to Drigg the wastes are fixed in a concrete grout prior to their orderly emplacement in a concrete lined vault. The majority of the waste typically comprises discarded protective clothing (overalls, overshoes, gloves, paper hats etc.) and general trash from areas of low contamination. The waste is generally accumulated in 200 litre drums and the total activity of such a drum is typically 1 to 2 MBq beta/gamma but can vary between 1 kBq and 20 MBq due to the inherent inhomogeneity of this type of waste.

3

WASTE QUALITY CHECKING LABORATORY

3.1 History The independent monitoring or quality checking of LLW is carried out at a laboratory established by the Agency for this purpose at the Winfrith Technology Centre in Dorset. The Laboratory was first postulated in 1983 and a contract to build the laboratory was awarded in 1985. During this first contract the design and construction of the facility were completed and the laboratory was equipped and staffed by scientists and technicians. The work took approximately three years to complete with the laboratory being commissioned in 1988. From 1988 to 1991, the laboratory was contracted to perform research into the analytical methods required to identify and quantify the wide variety of radioisotopes which can potentially be found in LLW. From 1991 to 1997 the laboratory was staffed and operated by Taywood Environmental Consultancy and performed routine quality checking of solid low level radioactive waste. In 1997 the present contract for managing WQCL was awarded to NNC Limited. This contract continues and expands on this work for the Agency with the development of neutron assay as an aditional non-destructive testing (NDT) tool and inclusion of site-based reference drum tests on operators' LLW drum gamma scanners. 3.2 Laboratory Description The WQCL monitoring facility is located on the Winfrith nuclear licensed site operated by the UKAEA. This provides secure facilities for consignments of radioactive waste to be received at

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the laboratory and secondary waste arisings to be disposed of via site services. The laboratory also makes use of other site facilities such as emergency services, site security and the library. The facility is housed in part of a large building, once used for an experimental reactor and consists of a suite of offices, laboratories, a workshop and waste receipt and storage areas. Waste consignments can be transported to the laboratory in a variety of containers, these include full height and half height ISO freight containers, individual drums and loose or packaged waste in skips. Following acceptance of the waste at Winfrith and receipt of the consignment at WQCL, the transport container undergoes a series of checks prior to opening and unloading. These are described in more detail in Section 5. For drummed waste received in ISO freight containers the drums are unloaded and stored pending examination. For loose or packaged waste received in skips, a tented enclosure can be erected for repacking the waste into drums in preparation for NDT. In addition to the waste receipt area, the ground floor of the facility also houses the gamma spectrometry laboratories and a permanent X-radiography facility which was installed and commissioned in 1996. The X-ray equipment was upgraded in 1998 to perform Real-Time X-radiography (RTX). The upper level of the facility comprises office accommodation and a suite of radiochemistry laboratories where the destructive testing (DT) and waste sampling is performed.

4

QUALITY ASSURANCE

Quality checking operations undertaken at the laboratory are carried out within a quality assurance system which was developed to ensure that all the work is performed to recognised and acceptable standards and that the results reported to the Agency are accurate and reliable. The quality system together with a number of key test methods were assessed by the National Measurement and Accreditation Service (NAMAS) in November 1993 and accreditation formally awarded to the laboratory in January 1994. Since then further test methods have been assessed and accredited as part of an ongoing programme to expand the laboratory's scope of accreditation. NAMAS became the United Kingdom Accreditation Service (UKAS) in 1997. Formal accreditation provides assurance that the measurements made on the waste are accurate and traceable to national or international standards.

5

NON-DESTRUCTIVE TESTING (NDT)

5.1 Waste Receipt Checks Upon receipt of a waste container at the laboratory, the consignment is given a unique identification number and each transport container is examined for evidence of damage or loss of integrity, any such findings are photographed and recorded. The labels attached to the transport container are photographed and all information recorded. Seals placed on the container by the Agency at the point of seizure are also examined and photographed. The container is checked for

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non fixed external contamination and radiation dose rate and finally the gross weight and external dimensions of the container are measured and recorded. Following completion of the transport container checks, the waste consignment is opened and the contents unloaded. For drummed waste received in ISO containers, the drums are unloaded directly into the waste receipt area of the laboratory and are logged into the QA system. Further checks are carried out on the waste drums at this stage. These include, radiation dose rate measurements, contamination checks, drum weight measurement and a note of the condition and integrity of the drum. Since the NDT procedures performed at the laboratory require the waste to be contained within 200 litre steel drums, any waste which is not received in this format, such as loose waste in skips, must first be repacked into drums. This requires direct handling of the waste by operators who must be suitably protected with the appropriate personal protective equipment. Repacking operations may also require a certain amount of size reduction in order to fit bulky items, such as lengths of wood or pipe, into the 200 litre drums. Each drum received or packed at WQCL is given a unique identification number and a seal is placed on the lid to provide proof of sample integrity whilst at the laboratory. The nondestructive testing is carried out on the whole of the waste consignment and involves two tests, these are X-radiography and Segmented Gamma Scanning (SGS). 5.2 X-Radiography Firstly each drum is examined by X-radiography to visually determine its contents without the need to open each drum. The Real-Time X-radiography (RTX) system consists of an X-ray generator and an image intensifier unit linked via a CCD camera to video equipment. The drum is placed on a motorised turntable and scanned at each of five vertical positions. Video recordings are made and used to produce and store still images on a PC. Each recording is examined by trained staff to determine the contents of the drum. This is important for the identification of any prohibited items as defined in Authorisation Certificates and BNFL's Conditions for Acceptance of wastes for disposal at Drigg. These include, free liquids, aerosol canisters, materials that are likely to cause fire or explosion hazards and large amounts of putrescible or rotable materials. The identity of any drum containing prohibited items is noted for opening and more detailed examination, at which point these items would be removed from the waste. 5.3 Segmented Gamma Scanning The most important non-destructive technique used in waste quality checking is Segmented Gamma Scanning (SGS) - see Figure 1. Using this technique the gamma emitting radioisotopes within each drum can be identified and quantified. Each drum is placed on a turntable within the instrument which allows the radioactivity within defined segments or slices of the drum to be determined. Up to 40 segments can be defined within a single drum, although, more typically, eight segments are used. The drum is assayed by a single high purity germanium detector which is aligned with each segment in turn.

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Individual isotopes are listed and quantified and the total radioactivity within the drum is then calculated by adding the results from each segment. A correction for the attenuating effect of the drum's waste content is made by use of an external Eu-152 gamma emitting transmission source. The instrument is routinely calibrated and checked using reference radioactive sources traceable to national standards.

Figure 1. The Segmented Gamma Scanner The wide range of waste material densities together with the large number of gamma emitting radioisotopes found in LLW can give rise to significant uncertainties in the radioactivities determined by the SGS. In an effort to reduce these uncertainties and achieve UKAS accreditation for SGS measurements, an extensive research programme has been undertaken. Experimental measurements have been made using known reference sources with a wide range of gamma ray energies placed at different positions within a series of drums filled with materials of differing densities. This work, together with projects undertaken in collaboration with European partners (see Section 8) has led to significant improvements in the use of SGS and interpretation of the results for waste monitoring purposes. A third NDT technique, neutron monitoring, is currently being considered by the Agency for implementation at WQCL. This will allow the presence of nuclear materials such as plutonium in LLW to be determined non-destructively.

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On completion of the NDT campaign the gamma emitting radioisotopes identified are listed and the total gamma emitting radioactivity for the waste consignment is calculated for comparison with the waste producer's declaration.

6

DESTRUCTIVE TESTING (DT)

6.1 Drum Selection In order to determine the alpha and beta emitting radioisotopes within a consignment of LLW, destructive testing must be performed on a representative portion of the waste. In general, approximately 5% of the packages or drums within a consignment are selected and sampled for radiochemical analysis. The criteria used for the selection of drums are dependent on the Agency's requirements and the nature of the waste being assessed. Examination of the RTX images, for example, may reveal prohibited items such as aerosol canisters or free liquids which must be removed. The presence of dense objects seen during X-radiography may conceal sources of radioactivity which may not have been revealed by SGS monitoring. Drums may also be selected from examination of the gamma emitting radioisotope content as found by the SGS and by specific request from the Agency e.g. based on the origin of the waste within the producer's site. 6.2 Sampling Once a drum has been selected for destructive testing it is transferred to the radiochemistry laboratory and attached to the waste receipt glove box. The lid of the drum is then removed from inside the glove box and the contents of the drum are examined. The waste receipt glove box is fitted with a fixed video camera and all drum opening operations are recorded on video tape. Any prohibited items found in the drum are photographed to provide evidence of the finding and segregated from the remainder of the waste which is then transferred to a second glove box. Here the waste is packaged, if necessary and the contact radiation dose rate and weight of the package are measured and recorded. Representative sub-samples are then taken and transferred to a fumehood for radiochemical analysis. Destructive testing begins with the preparation of an aqueous solution of the solid sample taken from the waste. This can be accomplished in a variety of ways depending on the type of waste material found. Methods such as acid dissolution, leaching or fusion are commonly used, the principal objective being the extraction of all the radioactive species into solution. Once the primary solution has been prepared, aliquots are first taken for the determination of total alpha, total beta, total and individual gamma emitting radioisotopes. 6.3 Analysis The total alpha measurements are made by preparing a counting disc from the primary sample solution by evaporation onto a planchette. This is then analysed in one of twelve alpha spectrometer cells, counting times being calculated from count rate measurements. The results are used to identify the component alpha emitters and quantify the total alpha radioactivity of the sample.

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Total beta determinations are made using Liquid Scintillation Counting (LSC). An aliquot of the sample solution is added to a vial containing a scintillation medium and is then analysed using a liquid scintillation counter. As for the alpha measurements the results are used to identify component beta emitters wherever possible and calculate the total beta radioactivity of the sample. Component gamma emitting radioisotopes within the sample are determined by analysing a 50mLaliquot of the primary solution in a fixed geometry on a gamma spectrometer. A separate low energy gamma spectrometer is used to determine low energy gamma and X-ray emitting radioisotopes such as Fe-55, I-125 and I-129. For all three of these techniques the chemical and radiochemical concentration of the solution must be controlled to optimise the counting characteristics and reduce interferences. All the analysis instruments used are regularly calibrated and checked using reference sources traceable to national or international standards. To ensure that the methods used and the results obtained from destructive testing are acceptable, the laboratory participates in regular inter-laboratory comparison exercises organised by the Agency as well as other bodies. In addition to the total alpha, beta and gamma techniques described, the laboratory has a number of other specific radioisotopic methods which can be used for destructive testing. The determination of specific radioisotopes by destructive testing first requires radiochemical separation from the other species found in the primary sample solution. The method adopted will depend on the chemistry of the element being isolated and may involve solvent extraction, distillation or ion-exchange chromatography. The laboratory has accredited analysis methods for most of the radioisotopes commonly found in low level radioactive waste in the UK. These include alpha emitters e.g the radioisotopes of U, Pu, Th and Cm and specific beta emitters; H-3, C-14, Sr-90, Ca-45, S-35, Cl-36 and Tc-99.

7

REPORTING AND APPLICATION OF RESULTS

All the results produced by the laboratory from quality checking operations on waste consignments are reported to the Agency. Written reports are produced on the findings of the non-destructive and destructive testing campaigns and these are forwarded to the Agency for review. Some typical results obtained from the quality checking of a waste consignment are shown in Table I. The tables compare the results obtained by NDT and DT for four different drums taken from three separate waste streams within a single waste consignment. It can be seen that in general there is very good agreement between the two techniques. From the regulatory point of view the results of the checking process may be considered as being in two distinct categories. Firstly, there are qualitative issues such as whether there was free

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liquid in the waste or whether the waste contained prohibited materials or items. Secondly, quantitative results can be compared with the activity as declared by the waste producer. Results reports are always issued to the relevant Agency RSR site Inspector who makes the judgement as to what action, if any, should be taken against the waste producer. Qualitative issues are usually an indication that either the operator's procedures are deficient in some way, or that the procedures have not been complied with. These are concrete issues which the Inspector would take up formally with the operator and would ensure by subsequent site inspections that adequate corrective actions had been taken. Quantitative issues can be much more complex, particularly where the results are from destructive testing and analysis. Results from SGS analyses are not dependent on sampling as all drums in the consignment are analysed whereas for chemical analyses a proportion (typically 5%) of the drums are selected and a further selection of material within a drum is sampled. Nevertheless, the correlation of total activities calculated by the two methods is generally much better than might be expected. This increases confidence in the results. If the check analyses indicated that authorised or declared activities had been exceeded further samples or analyses would be carried out to confirm the results. In all cases the follow up action is taken by the RSR Inspector for the appropriate site and in severe cases an operator would be liable to prosecution. Independent monitoring has so far given confidence that operators have taken a responsible and thorough approach to complying with disposal Authorisations. Non-compliances found have been in the nature of qualitative breaches as described above and appropriate corrective actions have been undertaken by operators.

8

EUROPEAN NETWORK ACTIVITIES

Since October 1992 the laboratory has participated in the European Network of Testing Facilities for the Quality Checking of Radioactive Waste Packages (ENTRAP). This Network was formed to promote co-operation between laboratories within the European Union who are involved in quality checking activities. The countries represented in this Network are: Belgium, Germany, France, Spain, Italy, The Netherlands, Austria, Finland and The United Kingdom. Each country has laboratory and regulatory participants represented on the Steering Committee of the Network and a number of Working Groups have been established to focus on important aspects of quality checking. There are currently three Working Groups whose remits are: (a) gamma and neutron measurements, (b) chemical and radiochemical testing and (c) quality assurance/quality control. The Steering Committee and Working Groups meet twice a year to discuss technical issues and matters of mutual interest. The Network is currently involved in jointly submitting a number of research proposals to the European Commission as part of the fifth framework programme on nuclear fission safety, having successfully completed a number of projects under the fourth framework programme. These research programmes attracted joint funding from the European Commission and included an inter-laboratory comparison test for gamma measurements on 220 litre waste drums, the optimisation of gamma assay techniques, which included improved

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transmission correction methods for SGS and the control of alpha emitting radionuclides in radioactive waste.

9

CONCLUSIONS

The Environment Agency exercises regulatory control over the discharge and disposal of radioactive waste under the Radioactive Substances Act 1993. Independent monitoring is a key element of this control and, in the context of solid LLW, the Waste Quality Checking Laboratory (WQCL) fulfills an important role. The quality and reliability of the laboratory's work is underpinned by formal accreditation of its test methods by UKAS and its participation in ENTRAP. The results from this work confirm that, in general, waste consignors have appropriate systems in place to ensure compliance with Authorisations. Identification of nonconformances continues to be a useful feedback to Agency Site Inspectors who use such information for identifying where operator's procedures and application of procedures can be further improved.

REFERENCES [1]

Radioactive Substances Act 1993, Chapter 12, HMSO London (1993).

[2]

Radioactive Substances Act 1960, 8 & 9 Eliz. 2, Chapter 34, HMSO London (1960).

[3]

Environmental Protection Act 1990, Chapter 43, HMSO London (1990).

[4]

Nuclear Installations Act 1965, HMSO London (1965).

[5]

Radioactivity in the Environment, A summary and radiological assessment of the Environment Agency's Monitoring Programmes, Report for 1998.

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TABLE I COMPARISON OF ACTIVITY DETERMINATIONS BY DIFFERENT METHODS Waste Stream 1, Drum 080201 Radioisotope NDT Results (kBq/Drum) Mn-54 12.4±5.1 Co-60 954 ± 90 Zn-65 52.3 ±22.5 Cs-137 N/D

NDT % of Total 1.2 93.6 5.1 0

DT Results (kBq/Drum) 2.4 ±0.4 1005 ±13 12.8 ±1.5 65 ±10

DT% of Total 0.2 92.6 1.2 6

Waste Stream 1, Drum 080223 NDT Results Radioisotope (kBq/Drum) Mn-54 1033 ±322 Co-60 18910 ±2470 Zn-65 1543 ±498 Cs-134 150 ±68 N/D Cs-137

NDT % of Total 4.8 87.4 7.1 0.7 0

DT Results (kBq/Drum) 1600 ±100 26000 ± 300 1900 ±400 128 ± 29 1.9 ±0.2

DT% of Total 5.4 87.7 6.4 0.4 0

Waste Stream 2, Drum 080228 Radioisotope NDT Results (kBq/Drum) Mn-54 53.5 ±19.2 Co-60 2278 ± 226 Zn-65 197 ±63 Cs-137 3.04 ±1.07 Eu-155 N/D

NDT % of Total 2.1 90 7.8 0.1 0

DT Results (kBq/Drum) 69.8 ±9.7 2691 ±24 275 ± 37 0.011 ±0.008 12.6 ±2.4

DT% of Total 2.3 88.3 9 0 0.4

Waste Stream 3, Drum 080237 Radioisotope NDT Results (kBq/Drum) Co-60 2840 ± 348 Cs-137 56.9 ±12.6

NDT % of Total 98 2

DT Results (kBq/Drum) 1772 ±42 82 ±7

DT% of Total 95.6 4.4

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C584/016/2000 Round robin test for the non-destructive assay of 220 litre radioactive waste packages L P M VAN VELZEN NRG Arnhem, The Netherlands

Summary A Round Robin Test for Non Destructive Analysis of 220 litre drums containing fissile and non-fissile radioactive material has been performed to validate and to improve the QA & QC of monitoring procedures. Various national laboratories involved in the independent checking of waste packages (i.e. members of ENTRAP) agreed that such a Round Robin Test would be a beneficial method for the validation of their procedures and results. The main objective of the Round Robin Test was to obtain comparable and representative data from various nondestructive examinations to improve the accuracy of present NDA techniques for 220 litre radioactive waste packages. The Round Robin Test involved fourteen 220 litre drums of nonfissile waste and three drums containing fissile material. These reference standards were prepared in the participating laboratories and transported between them in a sequence that enabled their simultaneous measurement at a given laboratory. All data collected during the tests has been collated. The overall conclusions that can be made of the Round Robin are that the non-fissile testing produced a good inter-comparison and achieved the project objectives. However, the fissile tests were not as satisfactory due to the lower number of drums involved, the widely varying activity inventories and differences in the nature of testing. The good results from the non-fissile Round Robin enabled state of the art performance assessments of the participating scanning devices to be carried out. This gained experience enables the setting up of a clear set of recommendations for best practice. 1.

Introduction

Various independent radioactive waste quality checking laboratories of various member states of the European Union are united in ENTRAP ( acronym for European Network of Testing facilities for the quality checking of RAdioactive waste Packages). These laboratories (see Figure 1.1) have developed NDA equipment based on various techniques and procedures to perform independent regulatory monitoring and control of radioactive waste packages which

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Figure 2.1 Map of Europe Showing the Participating Laboratories and Number of Standard Drums

are destined for interim storage or final disposal. It is evident that these laboratories must have reliable and validated NDA waste characterisation procedures in order to gain general acceptance by the public of the apparent problem of radioactive waste. Therefore these ENTRAP laboratories proposed a Round Robin (inter-comparison) test between themselves to benchmark their NDA characterisation techniques for fissile and nonfissile containing waste. Such Round Robin testing is an accepted method of validating results and systems and a way to improve the quality assurance and quality control. Hence the main objectives of the Round Robin Test (Ref 3) were: • To obtain comparable and representative data from various non-destructive examinations to improve the accuracy of present techniques for 220 litre radioactive waste packages but not to pass judgement on any one system • To combine pre-existing knowledge of the various techniques with the new experience gained from the Round Robin to produce a set of recommendations. 2.

Methodology

2.1

Selection of 220 litre standards

Each laboratory was allowed to supply up to two 220 litre standardised packages containing non-fissile material with the proviso that at least one of these was a good representative example of the 220 litre drums routinely controlled by that laboratory. Each laboratory had to

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assign a transport classification for their standards, i.e. LSA I, LSA II or LSA III according to the definitions of the transport classifications in the ADR (Ref. 1). The NRG Test Vessel was selected as the reference standard due to the fact that this standard had been previously proven to be completely homogeneous and known to contain only natural radionuclides. In total, fourteen non-fissile 220 litre standards were specially prepared and selected by the laboratories. The three standards for the fissile Round Robin Test were made of fissile reference material available at ENEA, FZJ Jiilich and TUM/RCM. Based on their experience and knowledge, the experts involved with NDA of fissile material decided to restrict the amount of fissile material to 3 gram 239Pu or 235U per 100 kg instead of 15 gram of fissile material per 100 kg as permitted in the ADR. An overview of the characteristics of the 220 litre standards is given in Table 2.1. Table 2.1 Classification Scheme for the 220 litre Standards in the Round Robin Test.

Matrix Distribution

Number of Standards

2.2

Uniform Radioactivity Distribution

Homogeneous

Shielding

Density

1

Yes

No

Heavy

Yes (Reference)

3 1

Yes No

No No

Light Light

No No

1

Yes

No

Bitumen

No

2 1 3 2 3

Yes Yes No No No

Yes No Yes No No

Heavy Heavy Heavy Heavy Heavy

Yes No No No No (Fissile)

Equipment and procedures

At the start of this Round Robin Test not all NDA gamma scanning systems were able to measure the final selected 220 litre standards. Therefore participants took care that necessary modifications were performed at systems and procedures were updated before the first transport took place. Modifications have been dealing with upgrading of the detector, mechanical components, electrical components, nuclear instrumentation, gamma analysis software, calibration and correction procedures. The following NDA techniques were applied during the Round Robin Tests: • Neutron Assay; Active and passive methods

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•

Gamma Scanning; Angular scanning, multi-rotational scanning, spiral scanning, swivelling scanning, dose rate, point measurement, emission- and transmission computerised tomography and radiography. The most frequently applied correction formulae were based on: • Calibration with the aid of a series of standard sources. In general sets of calibration curves will be recorded and used for interpolation between individual curves. • Numeric correction functions. These functions calculate in general photon energy dependent correction factors, based on the measured absolute efficiency of a calibration standard and an actual sample. • Monte Carlo calculations. These methods are able to calculate directly the absolute efficiency of detectors with or without collimators e.g. like the MCNP-code At this moment also "semi-empirical Monte Carlo codes" are existing. These codes are calculating a ratio between measured absolute efficiencies of calibration standards and actual sample-detector configurations.

However no single participating laboratory had access to all the above techniques. A detailed description of the systems used in this Round Robin Test can be found elsewhere (Ref. 2). To simplify the evaluation of the Round Robin Test results a generic classification of participating gamma scanning methods was set-up to take into account the scanning principle used, together with the gamma spectroscopy system, calibration and correction procedures, i.e. Table 2.2

Overview of the Definitions of the NDA Scanning Methods Used and NDA systems involved.

NDA waste scanning method Type Number 4 Cl

Homogeneous

3

•

C3

1

•

2

C5

3

Radial

Axial

Angular

Radioactivity distribution Non-uniform Uniform Radial Axial Angular •

•

C2

C4

Matrix Non-homogeneous

• • •

• •

•

•

•

•

•

Cl assumes a homogeneous matrix and uniform activity distribution for the whole drum. C2 assumes a homogeneous matrix for the whole drum and a uniform activity distribution within (height) segments of the drum. C3 assumes a homogeneous matrix for the whole drum and a uniform activity distribution within concentric (radial) rings in the drum. C4 assumes both a homogeneous matrix and a uniform activity within axial/height segments in the drum. C5 uses complete spatial information on matrix and activity distribution. The neutron assay systems could, in principle, be subdivided as in Table 2.2. However in any such a scheme an important distinction has to be made between active and passive neutron assay and, in passive systems, a further distinction between multiplicity counting ("doubles"

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and "triples") and coincidence counting ("doubles") since the latter requires matrix and activity distribution correction. However, such a scheme was not set up because the low number of participating systems made it meaningless. 3. Results All laboratories reported analysis results according to a predefined characterisation list to enable a common statistical evaluation procedure to be applied to all the results and thereby determine the performance of the participating systems. The following nuclides were involved in this Round Robin Test 110m Ag; 241 Am; 133Ba; 57Co; 58Co; ^Co; 134Cs; 137Cs; 152Eu; 154Eu; 155 Eu; 59Fe; 54Mn; 95Nb; 137Np; 238Pu; 239Pu; 240Pu; 241Pu/237U; 242Pu; 106 Ru and 106Rh, 124Sb; 125Sb; 234U; 235U; 238U, 95Zn, 65Zr. and in the reference standard: 214 Bi; 212Bi; 212Pb; 214Pb; 226Ra; 228Th; 232Th; 234Th; 228Ac; 208T1; 231Pa;

^"Pa.

The fissile Round Robin Test results are reported as 240Pu equivalent. Due to the fact most nuclides were not present in all standards, the results were too few per nuclide for each nuclide in the test drums, to perform a meaningful statistical performance assessment for the individual NDA systems. Hence, an unforeseen study and research had to be performed with as goal the development of performance assessment method. It is evident that the final data set of the performance assessment method, on which conclusions will be drawn, has to be comparable and representative. The developed performance assessment method is based on "the decrease of the Round Robin data set (all reported results) for each individual NDA system according to the same rules in such a way, that it contributes to the overall performance assessment of one of the five (Cl to C5) defined NDA gamma scanning systems". Before the applied rules are defined and the calculation method is discussed in more detail it is necessary to point out that both have been strongly discussed during the project. The main items discussed are summarised so that the reader can take these remarks into account by the interpretation of the results. The discussed items are (in random order): 1. Declared reference values. The declared reference values are considered to be the best estimate of the real specific activity of a certain nuclide in a round robin standard. 2. Quoted uncertainty. Taking reported uncertainties into account the method will be more complex and also more correct. However, the additional value will be small, due to the fact that it was found that most quoted uncertainties of the Round Robin Data set are underestimated. Therefore they are not taken into account, 3. Reporting non-present nuclides and not reporting present nuclides. Both are serious errors, however no attention is given to both points, due to the fact that no satisfying solution could be found to weight these errors in the assessment method. 4. Detection limits. These are not taken into account. The MDA is depending on a large number of parameters such as type and efficiency of the detector, measuring time, size and aperture of the collimator etc. and MDA levels can easily differ orders of magnitude.

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5. Presentation of the data. The method uses the parameter "the relative standard deviation R.Devx" Due to the definition of R.Devx the range can varie from "0 to •". The graphical presentation of the over-estimations will be presented better then under-estimations. 6. Calibration. Every participant has measured some standards outside their qualified calibration range (e.g. density, matrix homogeneity, non-uniform activity distribution). 7. Classification of the individual NDA systems. Specialists of the different laboratories have made the classification of each individual NDA system -see Table 2.2-, but have not taken into account all small differences in evolved procedures. 8. Elimination of complete standards. Most standards are a good representative drum for the type of drum in a participating country and are checked by a NDA system specially developed for these types of waste packages. Therefore the elimination of complete standards will influence the results of some individual NDA scanning systems and are not corrected for this effect. The following rules are defined and applied to all reported data of all individual NDA gamma scanning systems and combined for the overall performance of the five NDA systems:: 1. Select all reported data of all fissile and non-fissile standards (not the reference standard) without exclusion of nuclides of which declared values are known. This gives a total of 621 useful results out of the about 700 available results. About 80% of the 80 additional reported activities are below the 1 kBq/kg. The remaining nuclides are 110mAg, 241Am, 133 Ba, 58Co, ^Co, 134Cs, 137Cs, 152Eu, 154Eu, 54Mn and 125Sb covering an energy range of 60keVuptol600keV. 2. Decrease the data set with all reported results of which the declared value is at the level of the minimum detectable activity (MDA). As MDA level is presumed a specific radioactivity of 1 kBq/kg. A total of 549 results remains. 3. Decrease the data set with all reported results of standards that contain an inactive shielding. This rule is adopted, due to the fact that by studying the detailed results of each unique NDA system large deviations exists when inactive shieldings are involved. This is one of the examples that 220 litre standards are outside performed calibrations for a number of NDA systems. A total of 447 results remains. 4. Decrease the data set with results that are measured by less then 60% of the participating NDA systems. A total of 438 results remains. 5. Decrease the data set with outliers of each unique NDA system. These Max and Min values are identified separately for each unique NDA system by studying the total remaining data set of each individual NDA system. A total of 403 results remains. The difference between Measured and Declared data was calculated by the formula: „_ (Measured,, - DeclaredY ) R.Dev = — Declared^ where: R.Devx Measuredx Declaredx

is the "Relative Deviation", when isotope X has been measured in a standard by the NDA gamma scanning system is the "Measured" activity by the NDA waste scanning method is the "Declared" activity of the nuclide in a standard

By calculating R.Devx for all reported results of all nuclides in all standards and combining the results a Mean Relative Deviation can be calculated for a given type of NDA waste scanning method (i.e. Cl to C5, see above). An assessment of the accuracy of a given

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scanning method can be gained of the range of R.Devx given by its maximum and minimum values, respectively the Maximum Rel. Dev. and Minimum Rel. Dev.. Figure 2.2 presents the results of these calculations for respectively the defined Cl to C5 systems. The number next to the bar of Maximum Rel. Dev. corresponds with the number of values of R.Devx used to calculate the Mean and the Range. Three data sets are presented. The first set is comprised of all data. The last two have reduced data sets after the third and fifth exclusion rule. Figures 2.3 presents the results of the unique NDA scanning systems in the way of the "Bundesambt fur Strahlenschutz (Bfs)" and are ranked according to the calculated Mean. Bfs has build up a large experience in interpreting results of Round Robin Tests and intercomparisons and have discovered that this way of presentation is able to give information about the distribution of results of individual NDA scanning systems and the existence of systematic biases in detection methods, analysis etc.. Therefore it is not the score of each individual NDA scanning system, that is important but the pattern that comes into existence. Consider this Round Robin Test. All participated NDA scanning systems have no any direct or indirect connection with one of the other participating NDA systems. This means that the results of each NDA system can be seen as independent results. Further, there is no scientific reason known at this moment that one may not assume that these thirteen independent results will not be distributed normally. This means that the expected pattern has to be of a normal distribution -horizontal S-curve-. It is clear that this is only true when results are not biased. Quoted uncertainties of the results will have no influence on this pattern, because the uncertainty is a unique property of a NDA system and provides information about the validity of reported activities measured by this NDA system. 3.

Discussion and conclusions

Fissile Round Robin Test Discussion of the results of the fissile Round Robin test is limited in this paper due to the small data set. Large scatter and some ambiguity prevent the drawing of meaningful conclusions. The main reason for this lack of data is that some selected standards had a fissile content below the detection limit of some systems. In the final report of this European Commission project more information is noted. However, the recommendation that can be made for any future test, is that at the outset, the participating neutron specialists have to establish and agree a simple benchmark test that can easily be performed by laboratories with either an active or a passive neutron system. Non-fissile Round Robin Test It can be seen from Figure 2.2 that the range defined by the Maximum Rel. Dev. and the Minimum Rel. Dev. is lower for the "advanced" scanning methods C2, C3, C4 and C5 compared with the most simple method Cl. This means in general that advanced systems are much more sensitive for the interrogation of inhomogeneous matrices and for non-uniform activity distributions than simple systems. Why these systems are more sensitive and respond in this way will be investigated in the near future by comparing the methods in close detail. Also, it should be noted that, at present, no common correction algorithm exists that can be routinely applied to all combinations of inhomogeneous matrices and non-uniform activity distributions.

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Figure 2.2 Calculated Mean Rel.Dev., Maximum and Minimum Rel. Dev. as function of defined NDA Waste Scanning methods for different data sets from the non-fissile Round Robin Test

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Figure 2.3 Calculated Mean Rel.Dev., Abs ReLDev., Maximum and Minimum Rel. Dev. of the individual NDA Waste Scanning methods for different data sets from the non-fissile Round Robin Test ranked as function of the Mean ReLDev.

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In Figure 2.3 the pattern of the Mean shows a positive tendency, but changed rapidly to negative. Meaning, that most NDA scanning systems under-estimate the specific activity. Some other conclusions that have been drawn are: • All participating NDA waste scanning methods were are able to report specific activities with a standard deviation of about 10% for their routine wastes. This is due to the high level of calibration at the individual laboratories. • For well-calibrated systems, the standard deviation of reported activities of nuclides emitting photons of at least 150 keV will always be better then 50% -value of the most simple method Cl- for all non-routine packages without internal shielding and point sources. • For a number of packages the deviation between reported activity and the real activity will be much higher then 50%. These packages will mostly not fulfil the definition of a routine waste package and will probably contain a shielding and/or point sources. • The average systematic deviations of all the waste scanning methods (Cl to C5) is depending on the system and can go up to minus 30%. • The MDA of nuclides emitting photons above 150 keV is for all participating NDA scanning methods below 1 kBq/kg. The overall conclusion, which can be made of these Round Robin Tests, are, that the results of the non-fissile Round Robin Test are good in comparison with the project objectives, while the results of the fissile Round Robin Test are not satisfactory. 5 Recommendations for non-destructive assays of radioactive waste packages One of the main goals of this Round Robin was also to distil out of the gained experience and results of the Round Robin Tests a set of recommendations or general guidelines (Ref 3): • to improve the QA/QC of the NDA methods of waste quality checking laboratories • for the installation or set-up of a new radioactive waste quality checking laboratory. The major recommendation for the improvement of the QA/QC of NDA fissile scanning methods is the development of a simple benchmark test. This has to be easily performable by all laboratories and by both active and passive neutron systems. This will enable an assessment of the performance of these systems and a comparison of results. The recommendations for the improvement of the QA/QC characterisation of waste packages by NDA non-fissile scanning methods will include: • Modification of the specification list of a radioactive waste package, so that the most appropriate available NDA technique directly can be used and that no time (or money) is lost searching for the appropriate NDA technique. • Selecting and defining NDA scanning methods for new QA/QC laboratories. The results of the non-fissile Round Robin Test are used to assess the state of art performance of the five defined NDA scanning methods. These assessments are combined in Table 5.1 with assessments of installation and maintenance costs, analysis costs and the required skill of operators. The range of assessed performances are given according to the level of understanding of matrix homogeneity and activity distribution due to the fact that performed calibration(s) will not always be adequate for all non-homogeneous matrixes and non-uniform activity distributions. The lower limit of "<10%" is assessed to

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applicable for those waste packages that are very comparable with the actual calibration source/package used. Table 5.1 Table for selection of a NDA non-fissile scanning systems. Based on the assessment of the state of art performance for the five defined NDA non-fissile waste scanning methods (see Section 2.2) for nuclides emitting photons above 150 keV and for waste packages where activities are not shielded and are not acting as point sources.

•

State of the Art Performance Mean Relative Deviation

Type of System

Installation & Maintenance Costs

Analysis Costs

Skill of Operator

Cl

Low

Low

Low

<50%

<10% to <50%

C2, C3, C4

Medium

Medium

Medium

<25%

<10% to <25%

C5

High

High

High

<10%

<10%

Prior knowledge about Matrix Homogeneity & Activity Distribution None Full

Existing waste quality checking laboratories can improve their NDA gamma scanning methods by installation of procedures to detect and correcting for activities that are shielded or acting as point sources. At present, no satisfactory solution of this problem exists for routine analyses.

Acknowledgement All partners in this project wish to express their gratitude to the EC for their support during this project and also wish to thank all authorities involved in granting licences and permits for their assistance, comments and remarks. References 1. 2.

3.

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Accord European relatif au transport international des marchandises dangereuses par route Trb. 1959,8IE Biicherl, T.; Kaciniel, E.; Lierse. Ch.; Synopsis of Gamma Scanning systems; Comparison of Gamma Determining Systems and Measuring Procedures for Radioactive Waste Packages. Reference: European Network of Testing Facilities for the Quality Checking of Radioactive Waste Packages Report WG-A-01. Van Velzen, L.P.M.; Janssen, B.; Chabalier, B.; Delepine, J.J., Brunei, G.; Pina, P.; Morales, A.; Bardone, G.; Dodaro, A.; Troiani, Fr.; Pedersen, B.; Berndt, R.; Sanden, H.J.; Filss, P.; Kroth, K.; Odoj, R.; Bucherl, T.; Lierse. Ch.; Bruggeman, M.; Van Iseghem, P.; Carchon, R.; Lewis, A.; Daish, S.; May, R.; Botte, J.; Hendrickx, J.P.; Round Robin Test for the Non-Destructive Assay of 220 litres Waste Packages; EC rapport in press.

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C584/002/2000 The feasibility of surface storage for high-level waste LCAVE Risk Assessment Limited, Tunbridge Wells, UK

ABSTRACT The present ethical basis for the management of HLW in the UK is challenged on the grounds that it does not provide a just allocation of resources, on a global basis, between present and future generations, and it does not reflect correctly the intent of the ICRP's recommendations. The relationship in the UK between HLW management, nuclear power and fuel reprocessing (now and in the future) is indicated. The combination of these factors suggests that monitorable, retrievable, surface storage (MRSS) of re-processing waste for 1,000-2,000 years should be the preferred option. Factors relevant to public acceptance of MRSS are discussed.

1

INTRODUCTION

As shown by the difficulties experienced in most countries seeking to find acceptable sites for the storage or final disposal of high level radioactive waste (HLW), it is unlikely that it will be possible for the UK to implement a policy of deep underground geological disposal (DUGD) for many years. Thus there is an opportunity to consider whether, given the present pace of technological development, it is likely that an alternative solution could be found within the next 100 to 1,000 years which would be preferable on ethical grounds and acceptable in principle to the UK public in this generation. Some possibilities are: • • •

Transmutation of HLW to short lived isotopes. Advances in medicine that would reduce or eliminate the risk of radiation-induced cancers. Transfer of HLW to uninhabitable regions of the solar system.

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The use of monitorable, retrievable, surface storage (MRSS) with a design life of, say, 2,000 years would provide the necessary time for one or other of these developments. There is, of course, the possibility that within this time frame world order would collapse completely due to some catastrophe (e.g. a major nuclear war or an asteroid strike), leaving a small population eking out an existence at survival level. In that situation the hazard presented by HLW would appear trivial (See Section 2.3). The merit of MRSS as an interim solution is also dependent to some extent on the likelihood of the need for a large re-expansion of nuclear power within the next 50 to 300 years, in order to meet the demand for new energy sources (See Table 1). TABLE 1 - PHASES IN THE USE OF NUCLEAR ENERGY AND THE MANAGEMENT OF HLW Phase

Approximate Time Frame

Relevant Activity in UK During the Phase (a) (b) (c)

1

2000-2050 (d) (e)

2

3

2050 to 2100 or (a) (b) 2300 (say). See Footnote * (c) End of Phase 2 until replacement of nuclear energy by a better alternative.

(a)

(b) (c)

Use of nuclear power continues. Reprocessing of irradiated fuel continues; plutonium stockpiled. Design and costing of long-term MRSS facilities carried out. R&D into alternative methods for final disposal continues. "Ring-fenced" fund established to meet net costs for management of Phase 1 HLW and storage of plutonium. Possible temporary pause in the use of nuclear power. Re-processing of back-log of irradiated fuel continues. Construction and commissioning of long-term MRSS facilities begins. Nuclear energy re-introduced for power generations and heat, using stock-piled plutonium in breeder, or nearbreeder, reactors. Reprocessing of irradiated fuel continues, to provide a sustainable energy source. Development and commissioning of final disposal facilities.

[Note* The duration of Phase 2 depends largely on the feasibility of utilising the large reserves of methane which are in the form of methane hydrate.] These will become essential as the finite supplies of fossil fuels become depleted, or considerations such as global warming necessitate a reduction in the rate at which they are consumed. Thus at some time in the future the re-expansion of nuclear power is likely to be necessary; this would have a five-fold effect on the management of HLW. These factors are as follows:•

142

Future generations within the next few hundred years would have a much larger amount of HLW to manage than that accumulated in this generation and thus would have a powerful incentive to find the optimum solution.

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• •

•

•

The continued use of nuclear power would ensure that rigorous regulation of the whole of the nuclear industry would be maintained. The stock of plutonium from the operation of the existing power stations would provide a valuable energy source for future generations (e.g. if used as the initial charges for a program of fast breeder reactors, it could provide a fully sustainable source of energy) and thus could, at least in part, satisfy the requirement for equity between generations. The strong likelihood that the plutonium would be required in the relatively near future would render continuous reprocessing of irradiated fuel, using the existing capacity, the most economic way of providing this resource for future generations. At the same time it would simplify the management of HLW, as the half-lives of nearly all the other isotopes present in the waste would be much less than that of plutonium. Storage of the accumulated plutonium in a single, high security facility should, in UK conditions, avoid the possibility of any of the material being acquired by terrorists; the non-plutonium bearing HLW would be an unattractive target for terrorists or saboteurs.

All of these factors should assist in securing public acceptance of the MRSS concept in this generation. It is visualised that, in the UK, the various activities leading to the production of plutonium and HLW, together with their use, storage and disposal would occur in the three phases shown in Table 1. As much of the opposition to nearly every proposal for the management of HLW purports to be based on ethical considerations, the principal need is to examine the validity of the current arguments. The economic and other aspects are discussed in (1).

2 THE ETHICAL ASPECTS 2.1 Two Schools of Ethics Since the time of Aristotle and Plato there have been many different schools of ethics, often with a conflict of views between contemporary schools. The present generation is no exception, there are two main schools: the deontological school, which considers the rights of the individual, both in the present and future generations, to be paramount and the utilitarian, which is more concerned with the greater good of the greater number, both in this and subsequent generations. 2.2 The influence of the deontological school on the utilisation of nuclear power and the management of HLW Looking back some three decades it appears that the opponents of nuclear power, notably in the US, realised that the ideas of the deontological school could be used to provide an ethical basis for their opposition. Moreover, because of its concern for future generations and the extremely long term problems presented by HLW, the views of the deontological school provided an ideal means for demanding extremely rigorous requirements and then demonstrating that it was unlikely that they could be met, e.g. (2) and (3), leading to the conclusion that the use of nuclear power should cease as soon as possible. It seems likely that the ideas developed in the US are the basis of those put forward by "Greenpeace" and "Friends of the Earth" in the UK, which were largely instrumental in defeating BNFL's planning application for the "Rock Characteristics Laboratory" at Sellafields.

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The ideas of the deontological school also appear to have influenced the persons drafting the IAEA's "Principles of Radioactive Waste Management (4), which have been the "Bible" for regulations in this field. A major weakness in the position of the deontological school, which undermines the moral high ground claimed by the opponents of nuclear power, is its failure to consider the rights of the underprivileged member of this generation compared with (by inference) the average members of future generations. This aspect is discussed further in the next Section. A further weakness of the arguments of the opponents of nuclear power in this respect is the very clear precept of the deontological school that, if this generation depletes a scarce resource (e.g. fossil fuels), it should provide a replacement for future generations (5). Thus, in the absence of proof that non-nuclear renewable energy sources would be able to replace fossil fuels in a way that will be acceptable economically and environmentally, there should be no objection on ethical grounds to the continuation of the use of nuclear power, reprocessing the fuel (both proven technologies) and stockpiling the plutonium as a future energy source. It should be noted that, on ethical grounds, the replacement energy sources should have as low a cost as possible consistent with other constraints, until such time as there is no scarcity of resources for improving the lot of the least privileged members of society, as discussed in the next Section. In spite of these deficiencies in the ethical arguments it seems that the purveyors of deontological ethics have been successful in persuading the general public in most countries that HLW presents a more serious hazard than the nuclear power stations which produce it; as a corollary, in several countries the further use of nuclear power is now uncertain. Possibly this is due to the emphasis on the longevity of the HLW, although paradoxically opinion polls, e.g. (6), show that in general members of the public take little interest in possible events much more than 100 years ahead, i.e. the period encompassed by the lives of their children, grandchildren and possibly great-grandchildren. 2.3 An alternative ethical approach As noted above the deontological school of ethics, although very concerned about the welfare of future generations, has little to say about how scarce resources should be used to benefit the poorer sections of the present generation, possibly to the detriment of future ones. The scarcity of resources for such purposes is shown by the inability of most of the developed countries to contribute the promised 0.7 per cent of their GNPs to UN aid programmes; the average is less than 0.3 per cent. The utilitarian school, however would, in principle, seek to use scarce resources to provide an optimum distribution of benefits between the present and future generations, taking into account the current disparities in factors such as the provision of health care and opportunities for education within the present generation, on a world-wide basis. For example, Barry, a philosopher who is sympathetic to some of the views of both schools of ethics, argues that "Justice requires future generations should not be made worse off than our own but it does not require that they should be made better off than our own, bearing in mind that justice also requires some re-distribution of resources in this generation" (7). The moral obligation of the developed countries to reduce the risk of malnutrition and hunger in the developing countries is stressed in (8).

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In relation to HLW the utilitarian ethics school would seek to reverse the tendency to overregulation, regarding this as self-defeating, (9). In addition, it would take into account the disparity in the cost of a DUGD facility compared with an MRSS facility of equivalent capacity with a design life of, say, 2000 years (a factor of at least one order of magnitude) and the fact that the DUGD concept is not necessarily the optimum solution. Thus deferring an attempt to establish a DUGD facility would lead to a substantial saving of scarce resources. Utilitarian ethics would require that this saving should be divided partly between schemes to improve the present conditions of the most deprived members of society, preferably on a global basis, and partly on the development of methods for ultimate disposal of HLW, for the benefit of future generations. The latter could include transmutation, and cancer research, as well as the development of a more realistic basis for the design of a DUGD facility, aimed at making the risk from a DUGD commensurate with those from other sources (including non-radioactive, non-biodegradable, toxic waste disposal facilities). Meanwhile MRSS could provide a satisfactory solution in the following sets of conditions. Scenario 1 - World order is maintained and technological progress continues to accelerate at the present rate. Scenario 2 - World order breaks down and the technological basis of civilisation is lost (e.g. due to wars involving the use of weapons of mass destruction or the impact of a large asteroid). These conditions are assumed to last for about 500 years. Scenario 3 - World order is maintained but the rate of technological progress is reduced. In Scenarios 1 and 3 the future population would be as competent as ourselves, if not more so, to manage HLW; there could be some economic costs but radioactive hazards would be detected in time to avoid adverse health effects. In Scenario 2 it must be borne in mind that there would be several factors tending to reduce the relative importance of the hazards from HLW. •

•

•

The population of the UK would be existing at a survival level and would be greatly reduced, probably to about the level at the end of the Roman occupation, some two million. Life expectancy would be comparable to that in the poorest developing countries today, i.e. 35 to 40 years. Thus radioactive induced cancers would account for only a small proportion of all causes of death. Bacterial contamination of drinking water would present a far greater hazard than radioactive contamination.

In (1) it is shown, using WHO and UNICEF data, that there is a very large difference, in terms of lives saved, between the benefit obtainable by expenditure of a given sum on improving healthcare in the developing countries now or on seeking near-zero risk from HLW disposal, even in Scenario 2 conditions. For example £100 million, say, spent in this generation on seeking to maintain the radioactive hazard from drinking water in Scenario 2 conditions, at the levels currently contemplated, might save about 20 lives over 1,000 years; the same sum spent on helping the poorer developing countries to provide clean drinking water could save the lives of at least 500,000 children in the under-5 age group, in this

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generation alone. The scale of the problem is shown by WHO statistics: worldwide, about 2 million children in this age group die each year due to water-borne diseases alone. 2.4 Other aspects of inter-generation equity As indicated in the Introduction, it is quite likely that within, say, three generations the UK will need its accumulated stock of plutonium as a low cost, sustainable source of energy, i.e. Phase 3 of Table 1. Thus later generations, in the not-too-distant future, could receive a substantial direct benefit; a further benefit could result from the experience gained by this generation in the design, construction and operation of nuclear power stations, if this experience could be recorded in a suitable form, or preferably, if at least a small nuclear programme were continued in one or more countries. As these future generations would benefit from the plutonium separated from the HLW's held in MRSS, this should more than offset the detriment from the latter. Intermediate generations would incur costs in monitoring the MRSS and ensuring the security of the plutonium stocks. These costs would be offset to some extent by the reduction in CO2 emissions due to this generation's nuclear power programme, enabling them to use more fossil fuels within a given "CO2 emissions quota". On a global scale, the on-going nuclear power programmes, by slowing down the rate of depletion of the fossil fuel reserves, will also have helped to reduce the inevitable rate of increase in the cost of fossil fuels as depletion proceeds. To secure intergeneration equity for these intermediate generations provision should be made, by establishing a "ring-fenced" fund, to cover the net additional costs arising from the management of the HLW. 2.5 Proposals for a new approach to the ethical problem Up to about 1993, the writings of Dr K Schrader-Frechette, e.g. (3), a prominent member of the deontological school in the US, seemed to have been the main inspiration, at least in the US, for the opponents of the DUGD concept in particular and nuclear power in general; although a neutral reader might consider some of the arguments strained to the point of weakness (e.g. the need to replace finite resources consumed by this generation is ignored). However, in a more recent publication (10) this author has expressed the opinion that there should be an acceptable compromise somewhere between the views of the two opposing schools, although still leaning towards the deontological end of the spectrum. Unfortunately no explanation is offered for this change in attitude. This brief review of the ethical situation suggests that there is a need to reconsider the ethical basis for HLW management. In an era where there are great differences between living standards in the developed and developing countries but resources to reduce the differences are scarce, it should be generally acceptable, on ethical grounds, to allocate a higher proportion of those resources to the certain needs of the under-privileged in the present generation and less to the possible needs of future generations. This view has been put forward in the past, notably in the "Collective Opinion of the NEA Radioactive Waste Management Committee" (RWM) on "The Environmental and Ethical Basis of Geological Disposal of Long-Lived" Radioactive Wastes (1995). The Committee had explicitly excluded the intergenerational and intragenerational equity aspects in its brief to the authors of the background papers on ethics, on the grounds that it was too dependent on national political factors. Nevertheless, there was some debate in the discussion sessions and in its "Collective Opinion" (lla) the Committee stated "When considering resource allocation, risks from radioactive wastes must be kept in perspective with competing projects in the area

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of human health and environmental protection" but did not pursue the matter further. Thus the NEA's view on this aspect represents a correct interpretation of those of the International Commission on Radiological Protection (ICRP), as can be seen from the following points from its "1990 recommendations". The ICRP's fundamental principle is that radiation exposures of individuals and of the public as a whole should be kept as "low as reasonably achievable, economic and social factors being taken into account", particularly where it is not certain that the exposures will occur, (12a). The ICRP also provides recommendations concerning the optimisation of expenditure on radiological protection; clearly these recommendations are based on the utilitarian system of ethics, not the deontological one. The ICRP emphasises that its recommendations are confined to protection against ionising radiation and that although its risks need to be treated with care rather than fear; its risks should be kept in perspective with other risks. The ICRP comments that all those concerned with radiological protection have to make value judgements about the relative importance of different kinds of risk and about the balancing of risk and benefits" (12b). The ICRP also notes that "To search for the best of all available options is usually a task beyond the responsibility of radiological protection agencies (12c). As international organisations such as the IAEA, the EU and NEA (an agency of the OECD), together with most national governments, base their recommendations and regulatory requirements on the ICRP's recommendations, it is important that these should be interpreted correctly. In relation to the problem of maintaining a correct perspective on radiation risks relative to those from other sources, within the OECD countries only the Netherlands appears to have arrived at a solution: in 1992 the Government, after extensive public consultation, enacted legislation which put all toxic wastes (nuclear and non-nuclear) on an equal basis and has adopted a policy of retrievable interim storage for both (l1b). Below governmental level the need to consider both categories of toxic wastes equally has been emphasised elsewhere e.g. (13). In the UK it appears that, at present, the ICRP's recommendation to maintain an appropriate perspective on radiation risk in relation to other risks are being ignored, at least in the specific context of the management of HLW in the UK. It also seems to have been ignored by the IAEA. Acceptance of the need for a change in the ethical basis leading towards a more utilitarian approach, as advocated by the ICRP, would support the adoption of the MRSS concept, as it provides a means of reducing the immediate cost of HLW management and providing more time for research and development to determine the optimum solution for ultimate disposal. Finally, in considering the validity of the deontological arguments in relation to the management of HLW, it should be borne in mind that, without more resources for health care in the developing countries, millions of the children of this generation will not live long enough to be the parents of the next generation.

3. SECURING PUBLIC ACCEPTANCE OF THE MRSS CONCEPT IN THE UK 3.1 Factors affecting public acceptance of the MRSS concept in the UK There appear to be a number of factors likely to affect the UK public's acceptance of the MRSS concept. These and the counter-arguments, may be summarised briefly as follows:

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3.1.1 Fear of inadvertent interference or deliberate interference by terrorists or saboteurs. Separation of the plutonium, to be stored under high security arrangements, from the HLW itself, should allay the fear of deliberate interference with the HLW; inadvertent interference could be prevented by design. 3.1.2 With the passage of time, the storage building -would deteriorate and monitoring for leakage would be neglected Continuation of a nuclear power programme, or at least the expectation that a re-expansion was likely, and an MRSS programme would ensure that a regulatory body, covering its costs by fees from its "clients', would remain in existence, unless world order broke down. Attention should be drawn to the longevity of ancient Egyptian and Roman buildings (2,000 to 4,000 years) with little or no maintenance and to the provision that would be made in the design to detect and arrest any activity leached out by condensation. 3.1.3 Environmental changes and other external hazards could lead to the release of radioactivity. Major environmental changes, such as a large rise in sea level, could be foreseen in time to transfer the HLW to another location. The designers of nuclear power stations have demonstrated to the UK regulatory authorities that the risks from other, less predictable, external hazards (e.g. earthquakes) can be reduced to an acceptable level; for a MRSS the technical problems would be simpler. 3.1.4 Concern that further work on alternative methods of disposal would be abandoned. Given the "breathing space" provided by an MRSS with a design life of over 1,000 years, there would be ample time to determine whether technological developments (e.g. transmutation; advances in medicine that would reduce the hazard to one of the chemical toxicity or reliable space transportation) would provide alternative solutions to DUGD. 3.2 Some steps to secure public acceptance of the MRSS concept in the UK A vigorous PR campaign could lead to public acceptance: the principal features should be:•

Emphasis on the need to re-consider the question of intergeneration and intrageneration equity on a global basis, as discussed in Section 2.3, above. • The fallacy that non-nuclear renewable sources could provide an economic, continuous electricity supply in UK conditions (1). • The value to future generations of the plutonium already produced by the UK nuclear programme, as an economic, environmentally benign, energy source. • The strong possibility that, within 50 years, advances in medicine will have reduced the health hazard of radioactive materials to one of chemical toxicity. • The technical advantages of removing plutonium from the HLW prior to storage. • Safety aspects of an MRSS would be consistent with an agreed, clearly defined, ethical basis for both intergeneration and intrageneration equity. • Unless in the not-too-distant future the public in the UK and other developed countries were willing to accept the small risks associated with nuclear power and its attendant wastes, it is highly probable that they would be contributing to the early deaths of many millions of children in the developing countries.

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Private and government organisations concerned with reducing child mortality in the developing countries should be urged to support the PR campaign. The principal design features of a MRSS that would probably be needed to secure public acceptance are briefly as follows: • • • • • •

Design life of 1,000-2,000 years without major refurbishment, using proven long-life, (2,000 years) Roman pozzolana concrete. Waste stored dry, in vitrified form, using low leach-rate glass. Overall design in accordance with Nil's "Safety Assessment Principles for Nuclear Plant", including resistance to the full range of internal and external hazards. Air cooling by natural convection (i.e. no mechanical plant). Two lines of protection against leaching: use of heat from waste to maintain dry conditions; monitorable, leak-proof, drip tray. Facilities for retrieval.

4. CONCLUSIONS Because of its effects on the demand for scarce resources, the future policy for the management of HLW and the use of nuclear energy in the UK can have an important effect on the amount of aid that is likely to be provided to the developing countries in the near and the not-too-distant future. In the UK there is a general need for a change in the ethical basis on which decisions that affect, directly or indirectly, the welfare of the present and future generations are made, in order to achieve an equitable distribution of scarce resources between them. In the context of the management of radioactive wastes, a re-statement by the ICRP of its "1990 Recommendations", emphasising the need for intragenerational as well as intergeneration equity in this case should be helpful, as it could lead to a revision of the existing principles enunciated by the IAEA and other influential organisations. In the specific case of the management of HLW in the UK, a contribution to this more equitable distribution of resources between the generations could be made by the adoption, as the preferred solution, of the MRSS concept, thus reducing the immediate demand on the UK's scarce resources, as compared with the continued support of the DUGD concept as the preferred solution. Continuation of reprocessing of irradiated fuel and stockpiling of the plutonium for the use of future generations could also contribute to the more equitable distribution of resources between the generations. Adoption of the MRSS concept would provide more time for further research and development to determine the optimum solution for final disposal. Public acceptance of this change in UK policy would be facilitated by a clear exposition from Government of the factors leading to the change, including the likely affect of the inevitable changes in the principal energy sources; on the management of HLW in the UK. Securing public acceptance could also be helped by the well-publicised participation in the debate of organisations concerned with improving child health in the developing countries.

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REFERENCES (1) Cave L. (2) Schrader-Frechette. K.

(3) Schrader-Frechette. K.

(4) IAEA (5) Rawls J. (6) Duncan I. J.

(7) Barry B.

(8) AikenW. and Lafollette H. (Ed). (9) Keeny R. (10) Schrader-Frechette K.

(11) Nuclear Energy Agency

(12) ICRP

(13)Okrent D.

"The case for Surface Storage of HLW in the UK". RAL Report 02/00. "Nuclear Power and Public Policy: The Social and Ethical Problems of Fission and Technology." Pub. By Reidel, Boston 1983. "Burying Uncertainty: Risk and the Case Against Geological Disposal of Nuclear Waste". Pub. By University Calif. Press. 1993. "Principles for the Disposal of Radioactive Waste". "A Theory of Justice 2nd Edition, Harvard Univ Press 1999. "Some Aspects of the Relationship between Society and the Disposal of Radioactive Waste". Paper presented at "24th Annual Symposium". September 1999, London. "Circumstances of Justice to Future Generations" in "Obligations to Future Generations". Sikora R.J. and Barry B. (Ed). White House Press 1978. "World Hunger and Moral Obligations". Earthscan 1996 "Mortality Risks Induced by Economic Expenditure". Risk Analysis Vol 10, Part 1,1999. "Risk and Ethics" Background Paper No. 11 for IAEA Intl. Conf. on "Radiation and Society: Comprehending Radiation Risks". Paris 1994. "A Collective Opinion of the Radioactive Waste Management Committee of the OECD Nuclear Energy Agency": on the "Environmental and Ethical Basis of Geological Disposal of Long-Lived Radioactive Wastes". OECD 1995. l0a The Collective Opinion, p. 13. [The "Collective Opinion" was based on the discussion of a number of invited background papers from UK, other European and US authors]. l0b Appendix 5 of the Background papers. "1990 Recommendations of the International Commission on Radiological Protection". ICRP Publication 60. Pergammon Press. 3(a) Section 1.2. 3(b) Section 1.4. 3(c) Section 4.3.1. "On Intergenerational Equity and its Clash with Intragenerational Equity and on the Need for Policies to Guide the Regulations of Waste and Other Activities Posing Very Long Term Risks". Risk Analysis Vol. 19. Oct 1999. P 877-902.

ACKNOWLEDGEMENTS The views expressed in this paper are those of the author but the comments provided by various persons on earlier drafts of this paper, particularly by Mr. P. Beck, are gratefully acknowledged. COPYRIGHT. Reserved By Risk Assessment Ltd.

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C584/012/2000 Application of in-line monitoring to waste minimization during soil remediation TJ MILLER Decontamination Group, AWE (Aldermaston), Reading, UK

ABSTRACT The Decontamination Group (DG), at the Atomic Weapons Establishment (AWE), has successfully applied in-line radiological monitoring techniques, to minimise the volume of controllable radioactive waste arisings, during the remediation of uranium contaminated land. The approach adopted was to locate "hot spot" contaminated areas, using a sensitive x-ray counter (IS 610) developed at AWE, then excavate the "hot spots", together with the minimum of associated uncontaminated material, for radiological assay (IS 610) at a monitoring station local to the contaminated site, so that only genuinely contaminated waste was packaged for disposal and uncontaminated material was returned to its point of origin. Figures are presented, comparing the improved performance of this approach, with traditional remediation techniques, which involve an iterative process of sample analysis and excavation, until the desired end-point is achieved.

1 INTRODUCTION The AWE DG maintains a capability to develop economical, efficient and effective decontamination and monitoring practices, which minimise the volume of controllable radioactive waste arisings from AWEs operations. Traditional land remediation practices, relying on an iterative process of sample analysis and excavation, have proven to be expensive, time consuming and ineffective at meeting the stringent radiological end-points required. However, the development, at AWE, of a sensitive x-ray counter (IS 610), has enabled much cheaper, quicker and more effective in-line monitoring practices to be introduced. 1.1 Traditional remediation practices The contaminated site is cordoned off and a grid, usually one metre squares, is superimposed. Small samples, typically a few grammes, of the topsoil are removed at the grid intersections and sent away for laboratory analysis. When the results are known, the entire surface, often the whole area where above background samples were found, is then removed, usually down to a depth of one tenth of a metre, for controlled disposal as contaminated waste. The process of sampling and excavation is then repeated until the samples are uncontaminated or below an agreed end-point which is acceptable to all parties. For depleted uranium (DU) this could be

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11.1 Bq/g (1), but AWE normally aims for 2.5 Bq/g, with more stringent end-points for other isotopes. This approach is costly and lengthy, with no guarantee of reaching the desired endpoint, since large numbers of samples must be processed and, inevitably, large quantities of uncontaminated material are removed, along with the contamination, for controlled disposal, but some contamination may remain behind. 1.2 In-line monitoring practice A sensitive x-ray counter (IS 610) is employed to rapidly pinpoint "hot spot" contamination areas within the field. The "hot spot" topsoil is then transferred to a local monitoring station, sited in an uncontaminated area which is several metres outside the field, for assay by another xray counter (IS 610). If the removed material is below the end-point criterion, then it is returned to its point of origin. Only genuinely contaminated material is packaged for controlled disposal, according to its contamination level. Finally, the field is resurveyed to confirm that it is free of contamination. This approach is rapid and inexpensive because there is constant feedback on the progress of the remediation by direct monitoring, so that excavation stops when the end-point has been achieved and only contaminated waste, with the minimum of associated uncontaminated material, is removed for controlled disposal. 1.3 IS 610 x-ray counter The IS 610 was originally developed by AWE (2) for the detection of L x-ray photon and gamma emissions, from low level ground contamination, by the various isotopes of plutonium and uranium and their radioactive decay products, such as americium and thorium. It can be used as a hand held monitor, or mounted on its tripod (figure 1).

Figure 1 IS 610 mounted on its tripod

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The IS 610 detector consists of a NaI(Tl) crystal, with a diameter of 75 mm and a thickness of 1 mm. There are three regions of interest: channel 1 (Cl), 10-24 keV; channel 2 (C2), 47-72 keV and channel 3 (C3) 10-72 keV. Typical sensitivity and background counts (C2) are 2 cps/Bq/cm2 and 8 cps respectively (3). The normal field of view, when mounted on its tripod 30 cm above the ground (optimum working height), is a 4 m diameter circle. Detection limits improve with counting time (figure 2), but the standard, factory set, counting time is 100 s. However, the counter may be used in RUN mode, where the cps display is updated every second.

Figure 2 IS 610 detection limits Detection limits may be reduced by attenuating materials. However, the emissions from DU penetrate several cm of soil (figure 3), so the IS 610 can assay bulk samples, provided they are monitored as thin layers, contained in trays.

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Figure 3 IS 610 response to 60 kBq (4 g) DU powder under soil

2 CASE STUDY A small area of DU contaminated ground, approximately 40 m long and 3 m wide, was surveyed conventionally by taking samples of topsoil at the intersections of a m2 grid and sending them for laboratory analysis. Some samples had approaching 100 Bq/g alpha activity and averaged over 10 Bq/g. A subsequent IS 610 survey quickly identified a number of "hot spots" in areas where the analytical samples were high. The ground was then remediated using the in-line monitoring approach. Confirmation of achieving the desired end-point, of below 2.5 Bq/g, was obtained by resampling in addition to remonitoring the whole area.

2.1 IS 610 calibration DU powder was mixed with soil to produce trays containing a range of DU/soil standards from 0-6 Bq/g above background levels. The standards were then counted with the IS 610 in various mass and dimension configurations. The counting efficiency, E, (cps/Bq/g), was derived from the slope of the calibration plot (figure 4).

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Figure 4 IS 610 calibration for 30 Kg DU/soil standards in m2 trays Decision levels (Dcl), detection levels (Dtl), errors (Er) and activities (A), were calculated from the sample counts (S), background counts (B), count times (T) and efficiencies (E), using the following formulae (4): Del = B + (1.64 x B1'2)

counts

(Eq. 1)

Dtl = Del + (1.64 x Del1'2)

counts

(Eq. 2)

Er = 1.96 x (S + B)"2

counts

(Eq. 3)

A =(S-B/ExT)

(Bq/g)

(Eq. 4)

2.2 Field survey The whole area (40 x 3 m) was surveyed, in RUN mode (cps display updated every second), by walking slowly (2 kph) along the grid and using the hand grip so that the detector faced downward and slightly forward. All m2 squares having > 5 cps above background (C3) were noted and identified as 'hot spots'. These 'hot spots' were then surveyed in INT mode (100 s count), by setting up the IS 610 on its tripod so that the detector was 30 cm above the centre of the m2 square containing the 'hot spot', and collecting and recording the cps above background.

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2.3 "Hot-Spot" treatment All "hot-spots" were removed (shovel and bucket) for assay at the monitoring station. The IS 610 was used in standard INT mode and the soil was spread out evenly in a m2 tray to a depth of 2.5 cm. All trayloads below 2.5 Bq/g (above background) were returned to their point of origin. Those above 2.5 Bq/g were packaged for controlled disposal.

3 RESULTS 3.1 IS 610 performance Given the standard counting time of 100 s, the standard IS 610 counting geometry and a sample mass of 30 kg, spread to a depth of 2.5 cm in m2 trays, it was possible to achieve sub Bq/g detection levels for DU in soil. Efficiency factors were improved by using larger masses of sample (Table I). Table I IS 610 efficiency factors for bulk soil assay Soil Mass (kg)

10 3 10 30

Soil Dimensions (cm) 25 Diameter x17 height 48 x 57 x 0.9 height 48 x 57 x 3 height 100 x 100 x 2.5 height

Efficiency (cps/Bq/g)

1.7 0.6 1.7 2.7

3.2 Field survey and assay A few "hot spot" areas were located in places where samples had also given high results. Removal of the topsoil from the "hot spot" areas and assay at the side of the field showed that only one 30 Kg tray in 17 was contaminated above the end-point criterion of 2.5 Bq/g (Table II). This was disposed of as controlled waste and the other 16 trays were returned to their point of origin. Subsequently, the field was resurveyed with the IS 610 and found to be uncontaminated. This was confirmed by conventional sampling and laboratory analysis.

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Table II IS 610 field assay of 30 Kg trays of soil Tray

cps

Bq/g

1 2 3 4 5 6

4.3 4.5 3.6 3.1 4.9 3.3 2.0 1.9 3.4 2.4 3.4 0.7 2.1 1.4 1.3 8.0 3.8

1.6+/-0.1 1.7+/-0.1 1.3+/-0.1 1.1+/-0.1 1.8+/-0.1 1.2+/-0.1 0.7+/-0.1 0.7+/-0.1 1.3+/-0.1 0.9+/-0.1 1.3+/-0.1 0.3+/-0.1 0.8+/-0.1 0.5+/-0.1

7

8 9 10 11 12 13 14 15 16 17

0.5+ /-0.1 3.0+/-0.1 1.4+/-0.1

4. COMPARISON OF TRADITIONAL AND IN-LINE TECHNIQUES The costs, timescales and effectiveness of the in-line monitoring approach are all far superior to the traditional approach. Table III gives a simple comparison of costs. Table III Comparative costs for remediation of a small area of DU contaminated ground Operation

Traditional procedure (£)

DG procedure (£)

Site characterisation Excavation of contamination Confirmatory monitoring Waste disposal Totals

3,000

100 300 50 30 480

300 3,000 6,000 12,300

4.1 Site characterisation A conventional survey for a small area would cost around £3,000 and it would be several months before the results were known (5). Also, grid sampling in the presence of "hot spots" is a hit and miss method of determining the overall contamination distribution and its boundaries. This can lead to the excavation of much clean soil, along with the contamination, or missing contaminated areas altogether and removing no soil. By contrast, the IS 610 is able to rapidly home in on "hot spots" in RUN mode. The contaminated areas may then be monitored more accurately in INT mode. The whole operation takes only a few hours and would cost only £100.

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4.2 Excavation of contamination On the basis of the conventional survey alone it would have been recommended that an area of 60 m2 be excavated to a depth of 0.1 m, generating some 6 tonnes of waste. This operation would cost around £300 and may need to be repeated several times until subsequent surveys indicate that radiological end-points have been met. By contrast, the in-line monitoring technique generates only a small quantity of waste, since only contaminated surface soil is excavated. This is then assayed at the side of the field and only removed if it is above the end-point criterion. Clean soil is returned. The net cost for this operation would be around £300 for the small "hotspot" areas requiring treatment. 4.3 Confirmatory monitoring A second conventional survey would be required to confirm that radiological end-points had been met and would cost a further £3,000 and have the drawbacks noted for the initial characterisation. If the site were still contaminated, further cycles of excavation and monitoring would lead to rapidly escalating costs. However, a second IS 610 survey would be more rapid than the first, since only the excavated areas, where the "hot spots" were, would need to be examined. This would only cost around £50. 4.4 Waste disposal The traditional technique would generate around 6 tonnes of waste which would cost £6,000 if sent to Drigg as low level waste. By contrast the in-line monitoring technique ensures that only the contamination is removed with the absolute minimum of associated soil. This was only 30 Kg, with minimal disposal costs. This paper was published in the WM'00 Conference Proceedings, February 27-March 2 (2000), Tucson, Arizona, USA, on CD-ROM, 31-1.

5 REFERENCES 1. Radioactive Substances Act 1960 (RSA 1960) in conjunction with Exemption Order SI1002 (1986) (exempts natural uranium isotopes from controlled disposal below 11.1 Bq/g). 2. N.Harris, IS 610 x-ray monitor user manual, AWE, Aldermaston, Reading, Berkshire, RG7 4PR, UK, March 1994. 3. L.W.Hensman, use of the IS 610 for ground contamination measurement, Safety Division Technical Note 21/92, AWE, Aldermaston, Reading, Berkshire, RG7 4PR, UK, December 1992. 4. B.B.Warren, Environmental Monitoring Group Report no.35, AWE, Aldermaston, Reading, Berkshire, RG7 4PR, UK, 1981. 5. D.Urquhart, Environmental Monitoring Group Manager, costs of analysing environmental samples, AWE, Aldermaston, Reading, Berkshire, RG7 4PR, UK, 1994. © British Crown Copyright 2000/MoD

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C584/017/2000 Contained water management within the Chernobyl 'shelter object' A A KORNYEYEV Energoatom, Ukraine C R WILDING and T H GREEN AEA Technology, Didcot, UK A P KRINITSYN National Academy of Sciences of Ukraine, Chernobyl, Ukraine

1

INTRODUCTION

On April 26th, 1986, the biggest accident in the history of the nuclear industry occurred inside Unit 4 of the Chernobyl Nuclear Power Plant (ChNPP) in the Ukraine. Within six months, a containment was built over the remains of the reactor. It was not feasible to construct a leaktight containment in such a short timescale and under such difficult radiation conditions. Consequently, the "Shelter Object" contains a number of construction defects. A "Shelter Implementation Plan" (SIP) for stabilisation of the Shelter, and conversion into an environmentally safe site was subsequently formulated and funded by the G7 countries. The SIP Plan is administered by the European Bank of Reconstruction and Development. One aspect of this Plan is management of the water contained inside the Shelter. The scope of the paper includes water contained inside the Shelter and does not address the water inside the Turbine Hall. The paper discusses the sources of water inside the Shelter, and water losses. It summarises the results from characterisation of the water and presents the current status of studies conducted for its future management.

2

SOURCES OF WATER INSIDE THE SHELTER

Precipitation Rainfall and snow-melt leak through the roof of the Shelter. About 2,200 m3 of water per year could enter the Shelter by this mechanism. Condensation Condensation occurs of moisture vapour in the air flowing through the Shelter. During the period May to August, when the temperature of the air entering the Shelter is greater than that of the walls, structures and pools within the Shelter, moisture vapour condenses from the

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incoming air. About 1,650 m3 of water per year could condense on the walls, structures and pools by such mechanisms. Since 1996 a scheme to minimise condensation of water has been implemented in the Shelter. This has involved heating of the air and suppression of natural ventilation mechanisms. Dust suppression liquids Dust suppression systems periodically release water (containing a variety of chemicals) from sprinklers at various locations within the Shelter. The dust suppression system in the Central Hall has been operating since January 1990. From January 1990 to 1999 over 1000 tonnes of dust suppression compounds have been introduced into the Shelter. On average, about 270 m3 of water per year is released from the dust suppression system.

3

WATER LOSSES FROM THE SHELTER

Evaporation Water evaporates inside the Shelter. During the period January to April and September to December, when the air entering the Shelter is cooler than the walls and structures and pools within the Shelter, moisture vapour evaporates into the air. About 2,100 m3 of water per year could evaporate by such mechanisms and is carried out of or redistributed within the Shelter as moisture vapour in the air. Surrounding structures It has been experimentally determined (by visual observations and tracer studies) that water from the Unit 4 continuously moves through the dividing wall between the Unit 4 Nuclear Island Auxiliary Systems Room 001/3 into Unit 3. It then finds its way into the Unit 3 drainage water collection system. Surrounding ground Since 1990, investigations of contaminated ground in areas close to the Shelter have been made. This has involved drilling a network of bore-holes round the Shelter and sampling and analysing the water. Trace levels of Cs-137, Sr-90, uranium and plutonium have been found in the water.

4

WATER BALANCE

About 4,120 m3 of water could enter the Shelter each year by the mechanisms discussed above. The levels of the pools of water have been observed to remain fairly constant over a long period (i.e. recent years). The permanent water volumes within the Shelter are about 400 to 700 cubic metres, depending on the time of year. About 70% of the water accumulations are from the Nuclear Island Auxiliary Systems Room 001/3. If it is assumed that the evaporated water is carried out of the Shelter, the remainder of the water, 2,020 m3 (i.e. 4,120 m3 2,100m 3 ) could flow out of the Shelter each year, by leakage to groundwater and/or to neighbouring structures. The water flow is illustrated in Figure 1 and the balance in Figure 2.

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Figure 1: Flow of water through the shelter

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16s

Figure 2: Summary of shelter water balance Recent observations and tracer study results indicate that the water ingress and egress values discussed above could be overestimates by a factor of about two.

5

WATER CHARACTERISATION

Sampling Investigations of water within the Shelter began in 1991. To establish the composition of the water, 31 sampling points were identified that would provide the most information on principal water flowpaths. At these points samples of water are taken at regulator monthly intervals. The concentrations of radioactive isotopes of caesium, strontium, uranium and transuranic elements, heavy metals and the macrochemical composition were/are determined. For selected samples the activity on the suspended solid phase (from 0.03 to 1 |J,m) of collected samples has been determined. Water analysis The radiochemical concentration of the Shelter water is summarised in Table 1.

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Table 1: Radiochemical concentration of the shelter water

Radiochemical Cs-137 Sr-90 Plutonium Isotopes Uranium Isotopes

Minimum Value from all water samples taken 0.1MBq/l 0.2 kBq/1 2Bq/l 0.007 mg/1

Maximum Value from all water samples taken 100 MBq/1 10 MBq/1 l,300Bq/l 57 mg/1

Averages Values in Room 001/3

6.8 MBq/1 0.47 MBq/1 80 Bq/1 3.2 mg/1

Data on average concentrations of radionuclides dissolved in the Shelter water as a function of time are shown in Figure 3.

Figure 3: Variations of average nuclide concentrations of shelter water with time Sludge analyses When colloidal particles enter the accumulations of stagnant waters at lower levels of the Shelter, they precipitate and form sediments. The volumes of these sediments (sludges) has been estimated at approximately 100 m3 in room 001/3 (Level -1.0 metres). The radiochemical composition of the sludges are shown in Table 2 below.

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Table 2: Radiochemical composition of the sludges Property/Content Sr-90 concentration Cs-134 concentration Cs-137 concentration U concentration Pu concentration Mass of Uranium Radiologically significant radionuclides

Averages Values in Room 001/3 5.8E7 Bq/kg of dry precipitate 6.3E7 Bq/kg of dry precipitate 3.7E9 Bq/kg of dry precipitate 550 mg/kg of dry precipitate 6.8E5 Bq/kg of dry precipitate >50kg >2.7 TBq

Colloid analyses There are only limited data available on the radioactivity associated with the colloidal (0.1 to 1 micron) and the ultracolloidal (0.1 to 0.01 micron) phases. The samples were ultrafiltered, using nitrogen gas to pressurise the liquid through the filter medium. The results indicate that Cs-137, Sr-90, Pu and U are present in the solid phase particles with colloidal and ultracolloidal dimensions. It has been speculated that the alkaline pH of the water causes plutonium species with oxidation states (III) and (IV) to precipitate from solution in colloidal form. The amount of plutonium associated with the colloidal phase can be similar to that found in solution. It cannot be ruled out that the majority of the Pu in some cases might not be in solution. There are no data available for colloidal plutonium in Room 001/3, the room which contains the largest water accumulations.

6

WATER MANAGEMENT STRATEGY

In order to determine an optimum water management strategy it was necessary to identify the options available for each stage of the process, from collection through to disposal. These process stage options were then accepted or rejected by qualitative arguments. The acceptable process stage options were then combined to produce a number of strategic options. These were then assessed using semi-quantitative arguments to determine the preferred strategic option. Technologies which involved a considerable amount of work in the Shelter were rejected, since implementation could result in unacceptable construction and operational doses. For example, sludge collection by filtering the water in the Shelter was rejected. Not only was this a high dose rate option, but it would also generate an intermediate activity solid waste stream which would require provision of additional containers and premises for safe storage. The preferred strategy is to collect the Shelter water and sludges, then pre-treat them in a Shelter Water Pre-Treatment Facility (SWPTF) for removal of the organics and transuranics. The objective is to pre-treat the water to meet the ChNPP evaporator acceptance criteria. The pre-treated water will be evaporated and the ChNPP evaporator concentrates processed in the Liquid Radwaste Treatment Plant (LRTP). The sludges which arise from the SWPTF will be

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temporarily stored and eventually conditioned and disposed of as long lived waste. This is illustrated in the process flow diagram shown in Figure 4.

Figure 4: Shelter water pre-treatment facility flow diagram The main functions fulfilled by the SWPTF are thus: •

Hydraulically connect to Room 001/3 those rooms within the Shelter which contain water.

•

Collect the available water and sludges in Room 001/3.

•

Transfer the water and sludges to a sludge storage tank.

•

Store the sludge for an interim period.

•

Transfer the water to a treatment tank.

•

Reduce the organic content of the water by addition of oxidising chemicals (e.g. hydrogen peroxide).

•

Reduce the transuranic content of the water by adding sodium hydroxide to increase the pH of the water and thus precipitate the transuranics.

•

Transfer the liquor to a settling tank.

•

Transfer the pre-treated water to a sentencing tank and from there to the ChNPP water collection system. Evaporate the water using the ChNPP evaporator. The sludges from the evaporation process will be sent at the discretion of ChNPP to the Liquid Radwaste Treatment Plant for further treatment, including encapsulation in cement and then disposal.

•

Interim storage of the solid residues.

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•

Eventual encapsulation of the solid residues.

•

Eventual storage and disposal of the solid residues.

One major question was the possibility of a criticality. At an early stage, a criticality assessment was carried out on the assumption that all of the Shelter fissile material in the water, colloids and sludges was located inside one tank. The results indicated that there was not enough fissile material under any circumstances to cause a criticality. Therefore the strategy of collecting the water and sludges in one tank was sound from a criticality point of view. Arguments for location of the Facility outside the Shelter include those shown below. •

In the event of Shelter collapse, water will enter the Shelter and will require to be treated by the SWPTF. If the Facility was inside the Shelter it could be damaged by roof collapse.

•

The dose rates inside the Shelter rooms are higher than the ambient levels outside the Shelter. It would be a violation of the ALARA principle to expose operators to these higher dose rates on a regular basis.

7

FACILITY THROUGHPUT

The assumption was made that the current rate of water ingress would remain constant and that no measures would be taken in the short term to reduce water ingress. There are a number of possible scenarios which would alter the water volumes to be treated, and these are discussed below. 1. Work could be undertaken to block off some of the holes in the roof. This would reduce the amount of rain and snow which ingresses the Shelter. 2. A containment is built over the Shelter. This would reduce the amount of rain and snow which ingresses the Shelter to zero levels. There would still be a small amount of condensation water, but this has not been quantified. A containment system will not be built for practical reasons for at least 7 years or longer. 3. Dust Suppression Liquids: There could be a decrease or increase in the quantity of dust suppression liquids used in the Shelter. 4. Decontamination Waste: Liquid decontamination reagents could be used in the Shelter. These will need to be treated prior to discharge. 5. SIP Package D wastes: Package D are currently evaluating the use of wet chemical techniques for removal of some of the FCM. If this is technique is adopted, some contaminated water will require treatment. The amount of water is unknown. Figure 5 shows a hypothetical profile illustrating the volume of water which could be treated in the Facility over a twenty year period. Note that this analyses using more recent data

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which indicate that less water flows through the Shelter than that shown in Figure 2. A summary of the main points in Figure 5 is given below. Years 1-6: The Shelter water is treated for 6 years at a rate of 2,000 cubic meters per year. Years 7-11: Some repairs are made to the Shelter roof. There is a decrease in the volume of water to be treated. Years 11-15: In year 11, small amounts of decontamination waste are treated in the facility. This causes a slight increase in the facility throughput. Year 16: The Shelter containment is completed. About 400 cubic metros of water remains in the Shelter and this is treated in year 16. Water ingress is zero, but some condensation water enters the Shelter every year. Years 17-20: Dust suppression solutions, condensation water, decontamination water and Package D waste are processed from year 16 to year 20.

Figure 5: Illustration of water volumes which could be treated at the shelter

8

PRELIMINARY SAFETY ANALYSIS REPORT

A Preliminary Safety Analysis Report (PSAR) has been carried out on the conceptual design. The radiation doses and risks during construction, operations and under accident conditions were found acceptable, provided ALARA considerations were applied where relevant.

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A number of improvements to safety were identified and these will be addressed during detailed design.

9

CONCLUSIONS

There has been an extensive amount of work done to characterise the water within the Shelter and to determine the origins and volumes of water. The results indicate that the water compositions within the major water locations have stabilised and the chemical and radioachemical compositions are well characterised. Based on data generated the preferred water management strategy is to remove the water from the Shelter and pre-treat before sending to ChNPP for evaporation. The pre-treatment will remove transuranics and organics. The sludges will also be removed and together with sludges arising from pre-treatment will be temporarily stored prior to conditioning for future disposal as long-lived radioactive waste.

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C584/020/2000 ALARP as applied to high-level waste - the regulatory approach at Sellafield C H WAKER Nuclear Safety Directorate, Health and Safety Executive, Bootle, UK

SYNOPSIS Traditional risk management in parts of the nuclear industry assumes that if numerical risk levels, in terms of probability times consequence, are at or below numerical criteria then the requirements of ALARP are satisfied. However, risk in terms of health and safety law is to do with the potential to harm people, and if it is reasonably practical to reduce that potential then that is what the law requires, regardless of numerical risk figures. Furthermore, Nil's SAPs also set out as the basis for its safety analysis a fundamental hierarchy of deterministic principles. The first of these is to avoid the hazards and maintain safety by inherent and, where possible, passive design features. Where this is not reasonably practicable the design should be such that the sensitivity to faults is minimised. This hierarchy is applied with a rigour proportionate to the potential to do harm. These principles have been applied to NII's regulation of the storage of liquid HLW at Sellafield. They are the basis for a regulatory strategy which has the objective of reducing the hazard progressively by requiring the reduction of the stock to all but a minimum buffer quantity by around 2015. Some of the considerations NII has required of BNFL in its studies for defining a minimum buffer stock require a wider holistic view of the upstream and downstream processes to determine ALARP. This has caused NII to look at operational options itself, in order to benchmark BNFL's proposals and to constructively challenge assertions and assumptions.

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1. INTRODUCTION 1.1 In February 2000 NII published a report on the storage of liquid high level waste (HLW) at Sellafield (1). In it the safety issues related to the storage of HLW were discussed and it was explained why NII accepts that the operation of the plant is acceptably safe. 1.2 Liquid High Level Waste or Highly Active Liquor (HAL), is derived from the reprocessing of irradiated nuclear fuel at the Sellafield Works of British Nuclear Fuels plc (BNFL); it is stored and vitrified on the site. This waste has accumulated since the 1950s from earlier reprocessing and is also being produced from current operations. The HAL, which is a concentrated solution of fission products in nitric acid, is stored in a number of water-cooled Highly Active Storage Tanks (HASTs) housed in the Highly Active Liquid Evaporator and Storage Plant known as "B215". 1.3 HSE and BNFL have the mutually agreed policy that the HASTs should be emptied and the HAL converted into a solid form as soon as is reasonably practicable. The solid form of HLW adopted in the UK is borosilicate glass in which the fission products are incorporated, a process known as vitrification. Our policy is that this should be achieved by a target of around 2015, based on a judgement of what can reasonably be achieved. 1.4 A characteristic of HAL is that it generates heat so that, if not adequately cooled, it has the potential to reach a significantly increased temperature and, in the extreme, could boil. Overheating could lead to a reduction in the effectiveness of the containment system by, for example, over loading the ventilation clean up and filtration system. The need for high reliability cooling systems is therefore essential. By contrast the storage of HLW as a solid avoids any dependence on the continued availability of installed services such as electricity and water, or the need for operator control, because the fission products are immobilised in the solid matrix and the glass is cooled by natural air circulation. It is therefore passively safe. 1.5 Thus, despite the finding that B215 is acceptably safe NII believes that safety can and should be systematically and progressively improved by reducing the amount of HAL stored to a buffer stock, thus reducing the hazard potential associated with its storage. It also believes that the current operational mode should be reviewed to see if further improvements to safety can be achieved. It intends that these issues are controlled in accordance with the regulatory framework described below. 2. THE UK REGULATORY FRAMEWORK 2.1 The basis for the HSE policy that improvements to safety should be undertaken is the requirement in the Health and Safety at Work etc. Act 1974 (HSW Act) that employers must take measures to avert risks arising from their activities so far as is reasonably practicable. This requirement can be reformulated as requiring an employer to reduce the risk posed by its facility to As Low As Reasonably Practicable (ALARP) and is embodied in HSE's document on the Tolerability of Risk (2). 2.2 In considering safety issues it is useful to distinguish between the terms "hazard" and "risk". "Hazard" is usually defined as the intrinsic property of something which provides it with the potential to cause harm. "Risk" is the combination of the probability of the harm

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being realised and the consequences of such harm. In the case of HAL the hazard results from its high radioactive inventory, its mobility and its heat generating properties. 2.3 The implication, as set out in HSE's discussion document on reducing risk and protecting people (3), is that a successful approach to managing risks must ensure that hazards are properly addressed. It is also worth noting that the ACOP for the Management of Health and Safety at Work Regulations 1999 (4) states (paragraph 27) "it is always best if possible to avoid a risk altogether". 2.4 The requirement to reduce hazard further and to strive towards inherent or passive safety is reflected in the Nil's Safety Assessment Principles (SAPs) (5). The major part of the SAPs comprise engineering principles the first two of which (P61 and P62) are key. These form a hierarchy, the first of which is that hazards should be avoided by design and that safety should be maintained by passive means. Where this cannot be fully achieved, the next key principle requires the sensitivity of the design to faults to be minimized. It is these hierarchical safety principles, whose basis is in general Health and Safety law, that form the driver for HSE's policy that the HASTs should be reduced to a buffer stock as soon as reasonably practicable. A further requirement of these principles is that the design basis accidents must be assessed in the safety case using robust deterministic arguments with appropriately conservative assumptions. 2.5 To be compliant with the SAPs, a safety case should have at least three "legs" - a demonstration of sound engineering that the design is fit for purpose (based on fundamental safety principles and good practice), a conservative deterministic analysis of the design basis to show robust tolerance to more frequent faults, and a probabilistic (or quantified) risk analysis based on best estimate data for comparison with risk criteria and to search out weaknesses in the plant and its operations. The relative "weight" of these legs will depend on the specific technical considerations and the hazard. Probabilistic arguments are also important in order to ensure that the overall plant risk and the balance of risk across the plant are acceptable. Paraphasing from reference 4, it may be said that risk assessment must be suitable and sufficient for the purpose of identifying the measures needed to control the hazard. 2.6 A regulatory body must be aware of public and political concerns. Recent research for HSE into public perception suggests that there is more interest in large accidents irrespective of their likelihood (6). It is concluded from the above that intrinsic hazard is appropriate as a measure of the threat to safety and provides a meaningful concept for communication with the public. The policy on B215, that the hazard should be reduced, will address many of the concerns held by the public and appropriately focuses on the scale of the potential consequences rather than the risk in probabilistic terms. This is consistent with Reference 3. 2.7 The primary tool for nuclear regulation in the UK are the conditions attached to the nuclear site licence which HSE grants intend to use a nuclear site for defined purposes. This comes from from the Nuclear Installations Act (as amended) 1965, a relevant the HSW Act.

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powers derived from the to corporate bodies who delegated powers derived statutory provision under

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2.8 Licence condition 32[4] relates to the accumulation of radioactive waste on the site and allows the NII to "specify" "limitations as to quantity, type and form" of the accumulated waste. NII intends to utilise this condition in controlling the amount of HAL in B215. 3. HIGH LEVEL WASTE PRODUCTION AND MANAGEMENT AT SELLAFIELD1 3.1 HAL was first stored in B215 in 1955. These early arisings were generated from the reprocessing of irradiated nuclear fuel from the Windscale Piles and the early Magnox reactors. Some of this early HAL is still stored in the so called "Old Side" of B215. HAL is currently being produced at Sellafield as a result of the reprocessing of both Magnox and Oxide spent nuclear fuels and is stored in the "New Side" of B215. 3.2 A liquid waste stream, known as highly active raffmate (HAR), is produced in the first solvent extraction stages of the reprocessing plants. This raffmate contains more than 99.9% of the radioactive fission products present in the spent fuel. The HAR from both Magnox reprocessing and the Thermal Oxide Reprocessing Plant (THORP) is transferred separately to B215 via pipe bridges but is in volumes which are impracticable to store without treatment 3.3 The function of B215 is to receive, concentrate, blend and condition the HAR and thus store as HAL prior to vitrification. On receipt the HAR is fed on a semi-continuous basis to one of three evaporators and treated until the required HAL condition is achieved. The HAL is then transferred to the HASTs designated to hold the HAL derived from that particular source (i.e. Magnox or THORP) where further in-situ evaporation is carried out. 3.4 The HAL is subsequently blended and treated to the required composition by mixing the different types of HAL and the addition of additives prior to being exported to the Waste Vitrification Plant (WVP) for vitrification. The blending programme is intended to minimise the number of vitrified product containers which have to be made in WVP by ensuring that higher burn up material from oxide fuels is blended with Magnox material (or historic wastes from the "Old Side") to optimise the amount of waste oxide incorporation in the glass and avoid it being limited by constraints such as the heat rating. The planning of the associated liquor movements and the subsequent blending operation is thus complex. 3.5 The B215 facility currently consists of 3 evaporators and 21 HASTs. It has evolved into its present form over a period of 45 years with individual HASTs which have been in operation for very different lengths of time, ranging from only 10 years for the newest tanks, to 45 years for the oldest. The "Old Side" tanks, HASTs 1-8 ( Fig 1), were commissioned between 1955 and 1968 and are each of 70 m3 nominal capacity. These tanks have either one or three internal cooling coils, through which water is circulated to cool the HAL. 3.6 HASTs 9-21 (Fig. 2 and 3) - the "New Side" - are all of 150 m3 nominal capacity and were commissioned between 1970 and 1990. These HASTs each have seven internal cooling coils, and one or more external cooling jackets and include systems for agitating the contents. The design of these larger HASTs evolved from HAST 9 onwards. The main areas of development were the extension of the cooling jackets to the full operational height of the vessel from Tank 12 onwards, and modifications to the HAST emptying systems. 1

A more detailed description may be found in Reference 1.

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3.7 The HAL is maintained within a temperature range of 50-60°C requiring the active involvement of the operator, as does the maintenance of other properties of the HAL Cooling water is supplied to the HASTs via an open loop cooling water system from one of two sets of forced draft cooling towers. The cooling system and its associated pumps and power supply system have a high reliability with a large degree of redundancy. 38 The HASTs provide the primary containment for the HAL and are housed in the cells, (heavily shielded enclosures) to provide a secondary containment. The gases ventilated from the HASTs are routed to a dedicated vessel ventilation system which removes entrained radioactivity from the off-gases prior to them being discharged to the environment. 3.9 In the event of the need to take a HAST out of service, spare capacity is available to allow the contents of the HAST in question to be transferred to another tank. The spare capacity is equivalent to one spare tank for every three working HASTs, and is known as the "one-in-four" spares policy. This policy has been a requirement which NII has consistently maintained over many years. 3.10 The characteristics of the waste vary from tank to tank. The degree of radioactive decay of the fission products is related to the storage time. Consequently, the isotopic composition of the waste changes and the heat generation rate decreases as the storage time increases. The current heat generation rate of the HAL varies from about 20 - 460 kW per tank from Magnox fuel and can be up to twice the higher figure for oxide fuel. HAL from oxide fuel also has a higher specific radioactivity as a result of higher burn-up in reactors. 3.11 WVP came into active operation in 1991 and comprises two vitrification lines known as Lines 1 and 2. A number of operational problems have resulted in the throughput of WVP being significantly lower than originally intended. As a result BNFL has constructed a third vitrification line which is being commissioned. A discussion of the operational experience of Lines 1 and 2, and the improvements carried out to improve plant throughput is contained in Ref 1 and 7. 3.12 In WVP the HAL is first evaporated, dried and partially denitrated to produce a fine powder known as calcine. The calcine is then fed with crushed glass into a melter, in which the glass melts and the calcine dissolves, to create a molten mass. This is then poured into a stainless steel container, allowed to cool, and a lid is welded on. The container is transferred in a shielded flask to the adjacent Vitrified Product Store which includes an "export facility". 4. THE REGULATORY POSITION 4.1 Background 4.1.1 Subsequent to the publication by NII of reference 1 BNFL were required to respond to the report by 18 August 2000 and agree emptying curves with NII or NII would impose them. BNFL responded comprehensively to all 22 recommendations in NII's report before the deadline. In relation to the requirement to empty the tanks to a buffer stock by 2015, an emptying curve in terms of volume has been offered based on realistic assumptions about the ramp up of vitrification capacity and using the most up to date business plans for Magnox and THORP business. The submission also outlines areas for further development with milestones

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which should ensure continuous improvement to vitrification performance and tank operational regime. 4.2 Key Issues 4.2.1 Although quantified risk analysis in the BNFL safety case shows that the overall risks arising as a result of the operation in B215 are consistent with the requirements in Nil's SAPs, the P61/62 drivers raise the issue as to how the principle of ALARP is to be judged and the need for the harm potential to be further reduced. That debate is still ongoing as a result of the responses received by HSE to its discussion document at reference 3. However, Nil's approach has been to expect the licensee to establish the "modern standard" - i.e. what would it look like if designed today. This can be done by a process of "optioneering" to identify what is possible from a safety perspective and then to argue on the grounds of reasonableness and practicability how close the ideal can be achieved. 4.2.2 Whilst reprocessing continues there will always be a need for a buffer stock of liquid HAL to be held in order to allow operational flexibility between the reprocessing and vitrification plants and to enable the necessary conditioning of the feed stock to the vitrification plant. Thus achieving more inherent safety means reaching the point where there the storage of HAL is minimised , and in the meantime operating the existing plant such that it is inherently safer. Judging what this should "look like" must be done in a holistic manner. It will include ensuring the best use of the vitrification capacity and be clearly seen to be minimizing the harm potential whilst enabling the licensee to meet its reasonable and legitimate commercial aspirations. 4.2.3 There are a number of related issues in the case of HAL storage. Firstly there is the question of defining what is meant by a minimum buffer stock. The minimization of the amount stored will reduce the hazard potential, most obviously by reducing the consequences of any accident and should reduce the potential for harm to as low as reasonably practicable. The current design of tank and the operational mode effectively limits this to at least one tank. Our own assessment studies indicate that in the ideal small volumes may be achievable, a situation we have termed "near real time" vitrification. Nil's report required BNFL to challenge the present operational limits, including the consideration of smaller intrinsically safe tank designs for housing the post 2015 buffer stock. BNFL's response to the NII report has identified a number of key development studies with milestones towards this. 4.2.4 Secondly there is the question of the emptying strategy and the time and rate of approach to achieving the buffer stock. NII wishes to see the vitrification capacity exceed the rate of arisings so that the accumulation over the years can be reduced. This state has yet to be achieved and depends on the vitrification performance improving. NII welcomes the milestone programme that BNFL has put forward in its submission aimed at tackling the technical issues associated with WVP performance. We believe this is the key to success with its proposal. 4.2.5 The emptying strategy is technically complex and, in addition to vitrification performance, depends upon the assumed reprocessing inputs. Another considerations is the options for blending the different types of HAL so as to maximise the incorporation of oxides

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per unit volume of glass. This is important whilst vitrification capacity is limiting since it maximises volume reduction for a given production rate of vitrified product containers. 4.2.6 Thirdly there is the question of the parameters that should be used for monitoring the systematic improvement in safety, which is the whole purpose of the emptying strategy. Emptying discussions to date have used volume as the monitoring parameter. However, it has to be recognised that in the context of safety this has to be broken down to the constituent parts (i.e. THORP, Magnox - including relative age - and historic). This is because the hazard and risk are, to a first order approximation, strongly related to the oxide component, due to its radioactive inventory which dominates the potential dose in an accident, whilst the propensity to overheat is related to the material's age. For example, the historic HAL in the "Old Side" has cooled such that it cannot boil. 4.2.7 In its public report (1), NII identified a number of physical parameters in discussion with BNFL, but these parameters are only potential surrogates for measuring hazard or risk. It is recognised that none of the identified monitoring parameters alone is a satisfactory surrogate for the "risk". For example, another problem with volume that it is sensitive to both the performance of the evaporators and in tank evaporation. Whilst useful for monitoring improvements in safety as the plant moves towards the buffer storage state, and for informing decisions on options for that state, none of the parameters provide a substitute for a full safety analysis. BNFL has proposed that a safety index be developed and has included this in the milestone programme with its submission. The development of a meaningful index, which will show how the hazard changes with time, is a an important requirement to which NII is asking BNFL to give some urgency. 4.3 Current Status 4.3.1 NII has welcomed the response and recognises the considerable effort by BNFL to put it together on the time scale. The issues and the modelling by NII of options, in order to make a technically sound and quality assessment, are complex.'Discussion with BNFL to reach a technically feasible and acceptable position is not yet complete, so outcomes cannot be reported here.2 5. CONCLUSION 5.1 BNFL has made a large number of commitments which NII welcomes. A final decision on the tank emptying curves which will be used for regulatory purposes has yet to be reached. When achieved NII will use a Specification under Licence Condition 32[4] to ensure that BNFL meets its commitments. 5.2 A monitoring programme is being developed which will enable progress in the longer term to be reported upon as part of normal regulatory business via Nil's reports to the Sellafield Local Liaison Committee. 5.3 This short paper has of necessity touched upon only the more significant aspects of long and complex process of assessment and determination of an ALARP position. In the 2

However, NII promised in its report to publish an addendum in 18 months (i.e. mid 2001) explaining the outcome of its assessment and the final regulatory position.

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search for an agreed ALARP solution NII has sought to achieve a technically robust position that meets ALARP in all its facets, both deterministic and probabilistic. ACKNOWLEDGEMENT Although the author is leading for NII on the regulation of the B215 HAST emptying, it goes without saying that he is strongly supported by a team of inspectors that are working together in the assessment of the complex issues and their resolution. The author wishes to acknowledge their contribution to the thinking and analysis that is reflected in this paper. The paper is, however, the view of the author on a developing situation and therefore it does not represent Nil's formal view. The author is however grateful to NII for permission to publish. REFERENCES 1. The Storage of liquid High Level Waste at BNFL Sellafield, Health and Safety Executive (www.open.gov.uk/hse/nsd), February 2000. 2. The Tolerability of Risk from Nuclear Power Stations. HMSO Publications. ISBNO 11 886368 1, 1992. 3.

Reducing Risks, Protecting People, A Discussion Document, HSE Books, May 1999.

4. Management of Health and Safety at Work: Approved Code of Practice for management of Health and safety at Work Regulations 1992: HMSO Publications. 5.

Safety Assessment Principles for Nuclear Plants. HMSO Publications.

ISBNO 11 882043 5. 1992. 6. Public Perception of Risks Associated with Major Accident Hazards, HSE Contract Research Report 194/1998. 7. Progress with highly active waste vitrification at BNFL Sellafield. The Nuclear Engineer, Volume 36, No. 2, March/April 1995.

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S

Figure t

Cut away section of a smaller (70m3) HAST

Figure 2 Principal components of the high active waste storage tank

Figure

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3 Internal arrangement of a larger (150m3) HAST Note: The figure is based on HASTs 18 - 21

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C584/027/2000 Radiation safety problems arising with damaged nuclear submarines utilization V A MAZOKIN, M E NETECHA, YU V ORLOV, G A STANISLAVSKI, G A VASILIEV, and V V BORISOV Research and Development Institute of Power Engineering, Moscow, Russia

Reactor compartments (RC), which have lost their primary circuit tightness due to accident, involving the coolant ingress in the reactor compartment sections, while the reactors have (had) the cores with damaged fuel elements, are referred to RC with damaged nuclear steam supplied systems (NSSS). Radiation situation in the compartments after accident is characterized as hazardous and intolerable. Additional safety activities are required for the work in the compartments. The presence of nuclear fuel in such NSSS reactors means that the compartments are potential nuclear hazard. Before 1999 there were seven nuclear submarines (NS), their reactor compartment state which can be classified as the damaged one. They are NS of project 675 serial No. 175, 533, 541, project 671 serial No. 610, project 705 serial No. 900, 910, project 705k serial No. 105. Late in 1999 - early in 2000 irradiated fuel was unloaded from reactors of a damaged NS. It was the first complicated enough effort aimed at removing fuel from damaged cores, which permitted transfer of the NS mentioned to the category of conventional NS subject to utilization. The second damaged compartment has been removed from operation by now and is at a standstill in Gremikha. To formulate the main problems arising from utilization of NS with damaged NSSS, let us consider the technical state and radiation conditions in the reactor compartment in the postaccident period (see Table 1) - slide 1.

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TECHNICAL STATE AND RADIATION SITUATION IN THE AREA OF RC OF DAMAGED NS Nos.

1.

NSSS type VM-A

Accident character Date of accident 1979 Primary circuit coolant leakage and loss of integrity of fuel cladding

Radiation situation after 20 years of cooling Reactor screened section, port side - 0.5-2.3 mSv/h Reactor screened section, starboard side - 0.025-0.12 mSv/h Galleries - 0.005-0.049 mSv/h Light-weight vessel, above RC - 0.003-0.03 mSv/h In adjacent sections - 0.002-0.0025 mSv/h Beta contamination Reactor screened section, port side - 6500 particle/cm2min The rest RC sections - 40-190 particle/cm2min Major contaminant is Cs-137

Accident consequences The primary circuit lost its tightness. Reactor partition and the hold were contaminated by radioactive products. Starboard side NSSS was not damaged. Spent fuel can be unloaded from the reactors using the additional protection means for the personnel

TECHNICAL STATE AND RADIATION SITUATION IN THE AREA OF RC OF DAMAGED NS Nos. 2.

NSSS type OK300

Accident character Date of accident Primary circuit coolant leakage, 1985 accompanied by loss of integrity of a portion of fuel elements in the core

Radiation situation after 15 years of cooling Equipment screened section, reactor head, port side - up to 75 mSv/h Passage gallery - 7 mSv/h Pump screened section of floors I-III - 18-60 mSv/h Partitions of adjacent sections - 0.3-30 mSv/h Light-weight vessel above the reactor - 0.8 mSv/h Contamination of screened equipment section - 36000 particle/cm2 Major contaminant is Cs-137

Accident consequences The primary circuit lost its tightness. Reactor core was seriously damaged, inner premises of RC are severely contaminated by radionuclides. Radiation situation inside RC does not permit unloading of the cores from the reactors.

TECHNICAL STATE AND RADIATION SITUATION IN THE AREA OF RC OF DAMAGED NS Nos.

3.

NSSS type VM-A

Date of accident 1985

Accident character Self-sustained chain reaction during reactor refueling, fire in RC and ingress of sea water into

Accident consequences Reactor internals and reactor core were destroyed as well as reactor partitions and strong hull.

it

Radiation situation after 15 years of cooling Dead-end gallery - 14 - 60 mSv/h Passage gallery - 3 - 100 mSv/h Pump screened section (floor III) -1-3 mSv/h Partition of adjacent sections - 0.3 - 0.4 mSv/h Above RC at the light-weight vessel level - 3 mSv/h Major contaminant is Co-60

The technical state and radiation situation do not permit any operations in RC.

TECHNICAL STATE AND RADIATION SITUATION IN THE AREA OF RC OF DAMAGED NS Accident character Date of accident Loss of tightness of the CPS shim 4. 1989 rod claddings; radionuclide ingress in the upper cavities of the actuators shrouds Radiation situation after 10 years of cooling Streaming radiation over CPS shrouds reaches up to 30 Sv/h in a local area above the head Nos.

5.

NSSS type OK550

VM-A

1989

Primary coolant leak and loss of fuel cladding integrity

Radiation situation after 10 years of cooling Above the reactor head - 3.1-14.5 mR/h On light-weight vessel above the reactor - 0.4 mR/h Dead-end gallery - 1.5 mR/h Passage gallery - 100 mR/h Pump enclosure (floor III) - 0.9 mR/h Partition of the adjacent compartments - 0.5-1.5 mR/h

Accident consequences Radiation situation deterioration in the reactor head area. Unloading of spent fuel is possible if the additional means are applied for personnel protection.

The primary circuit lost its tightness, the hold and some premises in the RC were contaminated by radionuclides. RC can be removed if the additional means for personnel protection are applied.

The results provided in Table 1 (slide 1) suggest that, in the first place, the radiation safety problems in the NS are closely interlaced with necessity to prevent nuclear hazard and, in the second place, with the need to restore the destroyed or to create additional ecological barriers preventing radioactivity ingress to the environment. Thus, all the problems are concentrated around two major ones:

• nuclear fuel unloading from the reactors; • • bringing of RC with the damaged NSSS into radiation and ecologically safe state, bearing in mind subsequent transportation and storage in long-term storage places. The analysis of the damaged RC state and radiation situation indicates that under certain conditions at some projects (items 1, 4 and 5 in Table above) it proved possible to unload fuel from reactors and to localize the radioactivity remaining within RC, i.e. using welding to seal primary circuit nozzles, remove transition pipelines and plug their attachment points, install the additional ecological barriers (non-standard partitions along ends of RC, insulation of the lower part of the strong hull below the reactor, filling the reactor with a solidifying conserving agent) and isolate and seal the inner space of RC. The scheme of the protective barriers for RC is shown in slide 2. As was stated above, this technology has been implemented in part in 1999 at the nuclear submarine given as item 5 in Table 1. After 10 years following the accident the submarine has been brought into the ecologically safe state, namely: •

nuclear fuel was unloaded from the reactors;

•

standard heads were put onto the reactor vessels;

• primary coolant and liquid radwaste were removed from RC hold and transported for disposal; •

inner cavities of the reactor were filled with the solidifying conserving agent;

• the operations on preparing this submarine for short-term storage afloat are close to completion at present; • as soon as the local depository for RC from NS to be disposed is put in service, RC will be cut off (removed) from this submarine and subject to cooling procedure according to the standard technology. For the damaged RC, from which unloading of SNF is currently impossible (see items 2 and 3 in Table 1), it is suggested that the additional barriers are made not only inside RC, but outside them as well. Specifically, each of the above-mentioned damaged RC can be placed in a cylindrical steel casing, one of the RC larger diameter of the NS subject to utilization being used as such, for one example. For this variant of handling the damaged RC of NS (item 3, Table 1) the Central Design Office of Marine Technologies (Ts KBMT) «Rubin», RDIPE and other organizations have performed design and technological developments for bringing the

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RC in nuclear, radiation and ecologically safe state, the project being named «Sarkofag» (Sarcophagus) - slide 3. Specifically, nuclear safety of the reactor with unloaded core can be assured by removal of moderator (water) from the reactor, by its filling with solidifying conserving agent with absorbing additives, which provides a deep subriticality of the core, ruling out reactivity compensation members movement and water ingress inside the reactor, emergency external impacts of technogenic and natural character inclusively. Radiation safety is assured by providing an additional biological shield inside and outside of RC, as well as on the «sarcophagus» body. The points of installation and the thickness of the additional biological shield (made of concrete) provide gamma radiation exposure reduction to a value of 10 mR/h at most at a distance of 1 m from the «sarcophagus» body, which meets the requirements specified by the sanitary regulations for radioactive waste management SPORO-85. Ecological safety is provided by placing the damaged RC in a strong leak-tight shell made of three leak-tight missile compartments forming the «sarcophagus» and taken from the Pr. 677B NS PC which are to be utilized. The damaged RC is to be installed in the middle section of the «sarcophagus», being thus reliably isolated from the environment by its strong hull and two inter-sectional partitions on each side. In 50-60 years the radiation situation in the RC will be normalized as a result of natural decay of Co-60 dominating radionuclide, the conditions for conducting the activities of spent fuel unloading will emerge and the «sarcophagus» structures, including the damaged RC, could be utilized. The reactor vessels and iron-water shielding tanks will play the role of protective containers for remaining radioactive internals and they will be subject to subsequent arrangement in RW storage. As for the damaged NS (item 2, Table 1), where radiation situation is determined by Cs-137 radionuclide, its half life being 30 years, one should expect the normalization of radiation situation not earlier than 350-370 years later. So, for this NS PC the «sarcophagus» type package can not assure ecological safety for so long period owing to insufficient corrosion resistance. It seems that setting the RC in concrete, wrapping it by sand, gravel and rocks of the necessary thickness with water proofing and controlled leakage arrangement is a more reliable and feasible way to isolate the damaged section of this NS from the environment - slide 4. Hence, solution of nuclear and radiation safety problems during management of RC containing damaged NSSS shall proceed along the following directions: 1. Reactor compartments with damaged NSSS, their technical state and radiation situation permitting unloading of nuclear fuel from the reactors, shall be brought to nuclear-safe state by SNF removal from them; procedure for their further handling shall include the necessary measures of radiation and ecological safety. 2. In the cases when it proved impossible to unload SNF for reasons of technical state or radiation situation, the moderator (primary coolant) shall be removed from the reactor, shim rods shall be fixed in their bottom position thereby excluding their reciprocal

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movement, and the protective «container» (package) should be formed using the standard strong hull, partitions and additional protective barriers. If for some reasons the package could not provide ecological safety during long-term storage of RC, then more complex and expensive projects shall be used such as «sarcophagus» or «shelten> packages which have to be developed specifically for each damaged NS taking into account the particular circumstances of its technical state.

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C584/010/2000 Experience in nuclear decommissioning and waste management G R EDLER and D BRADBURY Bradtec Decon Technologies Limited, University of the West of England, Bristol, UK C J WOOD EPRI, Palo Alto, California, USA

ABSTRACT Chemical decontamination of radioactive components is a now a well-established procedure. Many components, particularly metallic ones, become contaminated with radioactive materials during normal operation of a nuclear power plant or other nuclear facility. In an earlier conference paper(1) a report was given on the application of the EPRI DFD Process to full reactor systems. At a recent EPRI Decommissioning Workshop'2' a number of papers were presented giving details of the benefits realised in the last two years. A summary of the decommissioning experience showed that techniques such as decontamination can offer savings in radiation exposure, waste disposal and dismantling times. New decommissioning challenges can be met with the development of new technologies. This is best achieved by close cooperation between the end user and the process developer.

1.

INTRODUCTION

Chemical decontamination of radioactive components is a now a well-established procedure. Many components, particularly metallic ones, become contaminated with radioactive materials during normal operation of a nuclear power plant or other nuclear facility. However, unless the component has been activated by a neutron flux, this contamination commonly occurs on the surface of the component only. The benefits of chemical decontamination include the reduction of radiation dose to people working on or close to, the component in question. More recently there are examples where the efficient decontamination of redundant components during decommissioning can allow the cleaned components to be released from radioactive materials controls so that they can be recycled or disposed of in conventional industry. This not only has economic advantages but

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benefits the environment as well through the recycling of valuable materials and the reduction in the volume of radioactive waste requiring disposal. Physical decontamination methods (eg such as shot blasting) are often preferable to chemical methods for cleaning externally contaminated surfaces which are easily accessible. However, these methods cannot normally be applied conveniently to complex components, structures or systems where the contamination is on the inside of the component or system. Chemical decontamination is usually preferable for such tasks. 2.

THE EPRI DFD PROCESS - RECENT APPLICATIONS

2.1 Full System Decontamination In an earlier conference paper(1) a report was given on the application of the EPRI DFD Process. This paper described the application of the process to full reactor systems after final shut down, and to a variety of components. Since then dismantling of the primary circuits has begun and information is now available on the benefit accrued due to decontamination. At a recent EPRI Decommissioning Workshop (2) a number of papers were presented giving details of the benefits. Ken Pallagi (3), speaking on the Big Rock Point decontamination(1), stated that two key issues in the decommissioning were dose reduction and alpha contamination control. As a result of the decontamination no airborne alpha was identified, no uptake or ingestion occurred, no dose assessments were required and only two positive smears in the RCP room piping were found. These measured 23 dpm/100 cm2 and 41 dpm/100 cm2. Immediate benefits identified were dose reduction and relaxing of job coverage requirements (two technician coverage versus four to six technicians). Due to the lower backgrounds, hot spots were easier to locate and evaluate. Removal or shielding of hot spots greatly reduced the dose received by workers. Further benefits from the radiation protection and ALARA perspective were minimal shielding requirements, minimal interference with work, no additional engineering controls required on the ventilation system and that during dismantlement of the primary circuit the containment building remained clean with no interference of other work. Table 1 is a summary of the dose savings in the recirculation pump room. Table 1 - Summary of Dose Savings in the Big Rock Point Recirculation Pump Room YEAR

1998 1999 April 2000

RECIRC PUMP ROOM WITH DECON 670 mSv (67 Rem) 340 mSv (34 Rem) 130mSv(13Rem)

ESTIMATED WITHOUT DECON 1150mSv(115Rem) 1630 mSv (163 Rem) 870 mSv (87 Rem)

ESTIMATED DOSE SAVINGS 480 mSv (48 Rem) 1290 mSv (129 Rem) 740 mSv (74 Rem)

Similarly Paul Plante and Glenn Collins(4) reported the benefits realised at Maine Yankee(1) as a result of the primary circuit decontamination. Table 2 is a summary of the dose benefits determined for the RCS system only.

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Table 2 - Benefit Determination at Maine Yankee for the RCS System Only Year

1998 1999 2000 Total

Actual Dose 560 mSv (56 Rem) 940 mSv (94 Rem) 520 mSv (52 Rem) (April) 2020 mSv (202 Rem)

Dose Estimate w/o Decon 1550 mSv (155 Rem)

Dose Savings 980 mSv (98 Rem)

2900 mSv (290 Rem)

1960mSv(196Rem)

870 mSv (87 Rem)

350 mSv (35 Rem)

5320 mSv (532 Rem)

3300 mSv (330 Rem)

Major Work Accomplished Asbestos Abatement RCP Removal Large Bore pipe, S/G Preps Loop Clean-up S/G & PZR Removal

The 3300 mSv (330 Rems) saved by the EPRI DFD treatment of the RCS system compared favourably with the original estimates of 2710 mSv (271 Rems) for a DF of 100 and 2560 mSv (256 Rems) for a DF of 15. Other benefits described included savings in packaging of radioactive materials from containment, minimising of dose to the public during on-site storage of radioactive materials, minimising of dose at the waste reprocessers, and lowered dose to the public during transportation of radioactive materials. Non technical aspects such as a positive impression by stakeholders and regulators were also reported. Lessons learned from these projects included •

"expect the unexpected"

•

flexibility in procedures and planning reduce delays

•

secondary waste volumes are not always determined by the decontamination process alone.

There was general agreement by the engineers that decontamination is best applied as soon as possible after final closure because of •

the use of station equipment, eg. reactant coolant pumps

•

having expert staff available on site

•

maximising the exposure saving benefits.

One trend associated with the removal of components is the diminishing benefits from the decontamination. The benefits are therefore most pronounced with high dose removal jobs early in decommissioning.

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2.2 Process Pumps The process has recently been used with great success in the USA for the decontamination of 400 series stainless steel pump components. A particular new variant of the process called "EPRI DFD Lite" is used for these components. The EPRI DFD Lite Process is applied to the pump components so that the components can be released to a workshop which has no radioactive materials controls for refurbishment and repair. This is considerably less expensive than the alternatives of discarding the components, or repairing in a radioactive workshop. In a recent paper the Southern Company(5) stated that application of the EPRI DFD Process had saved over $2M in the refurbishment of pumps in the period up to 1999 . The application of EPRI DFD Lite is performed by ALARON Corporation at their facility in Pennsylvania USA. The equipment used is an inexpensive recirculating system reported to have been assembled for approximately $5,000(6). Figure 1 shows a pump impeller being removed from the process tank.

2.3 US Department of Energy Trials have taken place at the Oak Ridge Reservation near Knoxville, Tennessee. Part of the accumulated redundant plant waste is the discarded aluminium compressor blades which have failed in service and subsequently been removed during maintenance outages. Several drums of blades were received by Decontamination Recovery Services (DRS) to be used in a decontamination trial at their Oak Ridge facility. DRS supplied a secure facility and project support such as health physics. The EPRI DFD Process was applied and engineered by a combined team from Practical Machine engineering (PMe) and Bradtec Decon Technologies. The blades tested were chosen at random from each of the four drums provided, but care was taken to ensure that a mixture of sizes was included. The isotopes of primary concern were uranium and technetium, technetium being the more abundant of the two. Over 70% of the blades were able to be released with some of the remainder being volumetrically contaminated due to re-casting in previous recycling campaigns. After the process trials both the cation and anion resins were successfully regenerated and a neutralised sludge produced. The sludge was successfully incorporated into a cement matrix. Samples of resin taken before regeneration were subjected to a Toxic Characteristic Leach Procedure (TCLP) test by an independent laboratory. The resin successfully passed the test. Security considerations did not allow the blades to be released from the DOE compound. However it was agreed that small cored samples could be removed from the blades for further laboratory investigation. The samples were tested at Bradtec's facility in Bristol. Figure 2 shows some samples before and after decontamination. The laboratory results supported those already recorded in the pilot scale work. Although the trials were technically very successful, a significant slowdown in progress has occurred as a result of the recent moratorium on recycling of materials announced by the Secretary Richardson in the United States on July 13, 2000. However, the main reason for concern leading to the moratorium is the potential for contamination in the public domain. There are many opportunities to recycle material which do not lead to the material being released into the public domain, and applications of decontamination to achieve unrestricted

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release standards are still likely to form an important future business, albeit with more restrictions placed on the outlet materials. 2.4 Feasibility Study Information released recently by the Ministry of Defence'7' indicates that after a study by them, they are considering the possibility of land storage for redundant submarine reactors. In a separate move Babcock Rosyth Defence Limited have been given approval to start feasibility and planning work on a proposal they have put forward to dismantle the reactor compartment from one of the decommissioned nuclear submarines, HMS Renown. Studsvik AB and Babcock International Group plc have recently formed a joint venture company known as Studsvik Rosyth Nuclear Services Ltd (SRNS). The new company is based at Rosyth Dockyard in Scotland. The function of the company is to provide services to the UK nuclear industry, particularly in the field of decontamination and radioactive waste management in decommissioning of nuclear facilities. Decontamination technologies are provided to SRNS by Bradtec Decon Technologies. The EPRI DFD Process is a central part of the Renown feasibility study. After decontamination one possible option for the metal components is that they could be sent to the Studsvik facility in Sweden where metal is melted to be released for recycling. Melting provides a further layer of assurance when measuring materials for unrestricted release. Internal surfaces and complex components are homogenised in the melt and processed into metal ingots. This process affords the opportunity for the precise determination of radioactivity and the production of archive samples. Melting also reduces the need for expensive space in waste repositories because only the slag has to be disposed of. Radioactive components with low radioactive content such as heat exchangers and moisture separators are routinely treated at the facility, which has an induction furnace with a capacity of 1500kg per hour. Aluminium is treated in a crucible furnace of about 250kg capacity. The facility has recently been upgraded to achieve a greater routine throughput. Materials are brought to the facility for processing not only from within Sweden but also from other countries such as Germany. Ingots from the melting process which achieve free release standards can be released in Sweden, but all secondary radioactive waste is returned to the originator. 3.

FOAM DECONTAMINATION

Although the preliminary step of filling a component or system with water for decontamination can normally be achieved, it is not always convenient. Some nuclear components are not designed to be filled with water, and the weight of the system when full would be in excess of the structural limits. An example of such a system is the Magnox gascooled reactor boiler. These items are carbon steel heat exchangers which (during operation) exchange heat from the reactor coolant gas to the water/steam circuit. The gas side of these boilers is contaminated, and was not originally designed to be filled with water. Another disadvantage of using water as the decontamination medium is that in systems with a large internal volume it is difficult to avoid stagnant pockets during the decontamination. If the decontamination chemical reagent is not efficiently circulated through the whole system

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volume during decontamination, then that part of the system in the stagnant pocket may not be efficiently cleaned. Such a problem would be encountered, for example in attempting to decontaminate a Boiling Water Reactor Steam Turbine. The restrictions on the weight of water could in principle be overcome by applying the chemical reagent in the form of foam, in which the liquid volume is highly expanded by entraining a gas. A new procedure has been developed for applying foam decontamination'8'. A foam decontamination reagent is placed inside the system or component to be decontaminated in an appropriate quantity to occupy a small proportion of the overall system volume. This proportion may be any proportion between about 1% and 10% of the system volume. Gas is introduced through a suitable inlet or inlets into the liquid volume at the bottom of the system. The gas becomes entrained to expand the liquid and thereby cause it to fill the entire volume of the system. When the decontamination reagent capacity is used up, the gas flow is ceased and the liquid is allowed to collect at the bottom of the system. The foam decontamination liquid is then removed from the system by pumping out or by gravity drain. The system surfaces are rinsed with clean water and, if necessary, returned to a dry condition thereafter. The radioactive waste management of the combined foam and rinsing solution employs traditional methods and principles. A filter may be used to remove insoluble particulate material from the waste solution. The waste solution may then be routed to a waste hold-up tank. In this tank the solution may then be mixed with chemicals added to achieve pH neutral conditions. The liquid may then be routed to an evaporator. For evaporation to take place efficiently it may then be desirable to add small amount of a suitable anti-foaming chemical. The condensate from the evaporation process can be recycled for use as rinse water or for further reagent make-up. The residue may be routed to waste drums for in-situ grouting with cement. The waste drums can thereafter be sealed and transported away for burial. Some trials of this new process have taken place with samples removed from a Magnox Boiler (see figure 3).

4.

FUTURE CHALLENGES

Nuclear decommissioning continues to raise new technical challenges and also continues to be a subject of great interest to the environmental lobby. Techniques such as decontamination can offer savings in radiation exposure, waste disposal and dismantling times. The decommissioning of nuclear reactors to a green field site in no more than ten years (Big Rock Point plans 8 years) is a measure by which the public can judge the maturity of our industry to cope with the aftermath of electricity generation by nuclear power. The American Light Water Reactors (LWRs) lend themselves more easily to early decommissioning. The need to develop suitable decommissioning technologies requires that engineering expertise and also a commercial incentive exist within the industry. We have at present the ability to develop and engineer decommissioning processes. This paper refers to one such process, EPRI DFD, which was initiated as a research programme in 1994 and by the end of

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February 1998 had been applied to two full reactor systems. This was only possible with the support of the US generating utilities via the research funding from EPRI. New fledging technologies, foam decontamination, and ideas, graphite pyrolysis'9' exist today. These technologies are best developed in conjunction with the end user to ensure a focussed and successful development programme. Ultimately everyone benefits from good process design and implementation when taking on the challenges of nuclear decommissioning. 5.

CONCLUSIONS

•

In the last two years of decommissioning at Big Rock Point and Maine Yankee, the benefits of the initial application of a full system decontamination using the EPRI DFD Process have been assessed.

•

It was agreed that considerable benefits were gained as regards reduction in personnel dose, reduction in waste generation and reduction in dismantling times.

•

Non technical benefits such as public acceptance and stakeholder confidence were also reported.

•

New decommissioning challenges can be met with the development of new technologies. This is best achieved with cooperation between the end user and the process developer.

REFERENCES 1.

Elder, G R, Bradbury, D, Wood, C J, "The application of EPRI DFD process for full reactor system decontamination post operational shut down", IMechE Conference Nuclear Decommissioning '98, Paper no: C539/016/98, page 113.

2.

Plant, P, Collins G, "System chemical Decontamination: Assessing its Benefit on Decommissioning Maine Yankee", EPRI Decommissioning Workshop, 13 June 2000.

3.

Pallagi, K, "Decontamination Benefit Assessment", Decontamination, ALARA and Worker Safety Workshop, June 2000.

4.

EPRI Innovators 2000, IN-114722

5.

Harverson, J, "ALARON's Experience using EPRI DFD Free Release Contaminated Components", EPRI Chemical Decontamination Conference, Greenville, SC. May 18-19, 1998.

7.

MoD Contracts Bulletin, 31 May 2000, p.20

8.

Foam Decontamination - British Patent Pending

9.

Mason, J B, Bradbury, D, "Pyrolysis and its Potential use in Nuclear Graphite Disposal", IAEA Technical Committee Meeting on Nuclear Graphite Disposal, 18-20 October 1999, Manchester, UK.

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Figure 1 - Pump Impeller treated by the EPRI DFD Lite Process

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Figure 2 - Aluminium Samples taken from Compressor Blades Top Row Untreated, Bottom Row Treated with the EPRI DFD Process

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Figure 3 - Foam Trials on Tube Samples from a Magnox Boiler

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C584/019/2000 Disposal of radioactive waste - a puzzle in four dimensions I J DUNCAN School of Geography, University of Oxford, UK

SYNOPSIS Society accepts that its demands for materials, energy, food and transport all create wastes some of it hazardous - and yet the public resists the disposal of such wastes, particularly if the intended site for a repository is in their locale. Waste disposal can be categorised into one of two possible regimes, firstly dilution and dispersal into the biosphere, or secondly concentration, containment and isolation from the biosphere. Waste can be gaseous, liquid or solid and varies greatly as to the degree of chemical and physical hazard. In the case of radioactive waste, the radiotoxicity decays with time and there is chemical change, two of the parameters that illustrate that waste disposal is indeed a dynamic process. 1. INTRODUCTION Some might argue that the public perception of the disposal of long-lived solid waste is a matter of placing it into a repository and trusting that it will remain there forever - a static situation. However research has shown that the public does not trust such a proposal and fully expects the system to fail and the waste, while still hazardous, to re-connect with the biosphere. Waste disposal experts counter this view with scientific expressions of degree of hazard and probability such as the risk that an individual human being will suffer a serious health effect (fatal or genetic) from any releases of radioactive material from a sealed repository should be less than one in a million (10-6) per year. It is seldom realised in the industry that such a statement is unintelligible to the public (5, p. 194). This paper develops a hypothesis that takes into account public perceptions of time and trust and considers the disposal of solid contained Intermediate Level Waste (ILW) as an example. The outcome could however be equally applicable to the disposal of all hazardous waste including Low Level Waste (LLW) and High Level Waste (HLW). Specific issues addressed include the questions: • Is the disposal of radioactive waste a static or dynamic process? • Does the public trust geological disposal? • If there is a failure in the system can it be remedied? • How can science and engineering develop a disposal method that will not only isolate the waste from the biosphere for the necessary period but will also win public acceptance and confidence?

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The traditional classification of radioactive waste has, at least to experts in the field, inferred the degree of hazard and thereby implied the isolation period necessary for each category. The isolation period for LLW is say 300 years, ILW about 5000 years and HLW at least 100,000 years. However the newly proposed categories of exempt wastes (EW), very low level waste (VLLW), short-lived low and intermediate level waste (SL-LILW), long-lived low and intermediate level waste (LL-LILW) and high level waste (HLW) should be promoted as the temporal requirements will then be more apparent to the public, although not quantified. Clearly the volumes and isolation periods for each category of waste are known in the industry, illustrating that disposal is indeed a puzzle in four dimensions. The temporal dimension of the disposal is discussed, the dynamism of the process is established and public perceptions considered in 2. DISPOSAL OF RADIOACTIVE WASTE. The perception that waste in a geological repository will re-connect with the biosphere in a relatively short period of time (less than the time required for isolation) is a commonly held belief. Is this due to the public having sufficient knowledge to be able to predict a failure of the multi-barrier system or is there another cause for this belief? This is discussed in 3. TEMPORAL DIMENSIONS. For sometime the disposal industry has toyed with the concept of retrievability of wastes. This is discussed in 4. REVERSIBILITY, RETRIEVABILITY AND REMEDIABILITY. Taking into account the issues considered in this paper a hypothesis is constructed that may assist in resolving the apparent impasse to public acceptance of geological disposal of wastes. This is presented in 5. A GENERAL HYPOTHESIS FOR WASTE DISPOSAL and leads to 6. THE HYPOTHESIS APPLIED TO ILW. The impact of this hypothesis on the science and engineering which must underwrite the safe disposal of radioactive waste is drawn together in 7. CONCLUSION. In presenting this paper to the Institution of Mechanical Engineers in a conference on Radioactive Waste the underlying expertise of the audience is acknowledged however generic background material can be found in the publications referenced in (1), (2), (3) and (4). 2. DISPOSAL OF RADIOACTIVE WASTE The regime for disposal of all waste falls into one of two categories, firstly dilution and dispersion into the biosphere or secondly, containment and isolation from the biosphere as shown in Figure 1. Waste Disposal Concepts. The biosphere is that part of the atmosphere, hydrosphere or lithosphere that supports life. Radioactive wastes are disposed of by both methods with gases and dilute liquids being dispersed into the air, sewer and soils and solid higher-level wastes being conditioned, compacted, contained and isolated from the biosphere probably within a multi-barrier deep geological repository. Radioactive waste exists and awaits disposal in considerable quantities, particularly in countries having nuclear power and a weapons program. It will continue to accumulate irrespective of the future for nuclear power and yet it could be argued that each generation should put in place the regime for the disposal of the wastes that it generates. The author's current research into public attitudes to time, space, risk and trust relevant to the disposal of radioactive waste has highlighted the dislocation between the public's perception

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of forward time horizons and the required isolation period. Siting issues are not pursued here as these have been addressed by others and by the author in papers referenced (5) and (6). There is a need to go back to the fundamentals to develop a waste disposal regime that must be more acceptable to the public and thereby advance the disposal of these wastes. Such a hypothesis needs to take into account that waste disposal is a dynamic process and for public acceptance of the scheme, their perspectives on space, time, risk and trust must be taken into account. Experts in this field believe that the multi-barrier geological disposal system envisaged for the higher level wastes provides not only a retardation of possible contact with the near-field, but would modulate the rate of reaction between the waste and the host material. It would also provide a level of redundancy should one or more of the barriers default. In essence, the system must provide effective barriers to leakage but equally importantly it must provide the time necessary for the decay of radiotoxicity. The barriers must provide the temporal buffer for this dynamic process. The following research will show that people generally accept or reject a disposal method based upon their time perspective and as that is much shorter than the required isolation period, they will usually reject the concept. 3. TEMPORAL DIMENSIONS Extensive polling of UK University students (n=688) and omnibus polling of German speaking Swiss (n=1057) and Japanese (n=2203) has confirmed that a majority of the public believes that geological disposal (at 500 metres) will fail and the wastes will re-connect with the biosphere in less than 1000 years. Question Q8. in the Swiss poll and the results obtained illustrate this point. Q8. If hazardous waste is buried in solid rock 500 metres below the Earth's surface, do you believe that it will be isolated from the living environment for: (a) 100 years 58% (b) 1000 years 19% (c) 10 000 years 5% (d) forever 13% These results are consistent with the UK outcome (not polled in Japan). When compared to the proposals for ILW disposal at 250-500 metres depth and requiring isolation for up to 5000 years it becomes clear that the public do not trust the concept with 58% inferring that the system will fail in less than 1000 years. Further polling results on public perceptions of forward time relative to self, family and community, show that a majority of people have an outer time horizon not further forward than the life on their grandchildren or say 100 years. This work has been fully described in references (5) and (6) and is illustrated by the following questions that were polled in the UK, Switzerland and in Japan. Q5. When considering the future welfare of yourself and your family, how far forward do you think? UK 92%, (Japan 91%, Swiss 87%) selected an outer time horizon of 100 years or less. Q6. When considering the environmental welfare of your home township, how far forward do you think? UK 54% (Japan 64%, Swiss 64%) selected 50 years or less; UK 90% (Japan 89%, Swiss 84%) selected an outer time horizon of 100 years or less.

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While the results of question Q8. suggest that the public has an understanding of the concept of geological disposal but still believes that it will fail in a relatively short time, questions Q5. and Q6. provide scope for another interpretation. Perhaps the relatively short expected life of the repository is not necessarily due to a critical understanding of the geotechnical aspects but is a manifestation of personal beliefs that are limited to less than 100 years for some 55-65% of the population and less than 1000 years for some 85-90%. It is the author's view that these beliefs are so entrenched in the population that discussions about isolation for 5000 years, or even more so for 100,000 years, are incomprehensible and therefore will be rejected. From these data, the outer time horizons of the population are clear and in democracies subject to plebiscites, it should be remembered that a simple majority decision is sufficient to decide the outcome of a siting issue. In the case of quantifying the forward horizon in the categories of family and community, 50% of the people have a horizon of less than 50 years! This research has led to the realisation that the time spans necessary for the isolation of radioactive waste are far beyond the time horizons of the 'common man'. This is a very significant finding for the nuclear industry and it may be the underlying factor as to why the responses to question Q8. have a relatively short horizon when compared to the scientific requirements. That is, perhaps the cause is not that there is a knowledgeable belief that the isolating barriers will break down in such a short time (58% in less than 1000 years), but that the natural outer limit for analytical thought of the 'common man' when considering self, family and community, impacts on the response. Glossy brochures depicting smiling operators, mountain streams, tree covered hills and decay diagrams on a log/log basis will not alter public opinion as to the future environmental security of a geological repository. So where does this take industry?

4. REVERSIBILITY, RETRIEVABILITY AND REMEDIABILITY The waste disposal industry has considered the option of retrievability for a decade or so. For example the concept appears in the publications of the Swedish, Finnish and UK industry, but never prominently. The principal concerns of industry are that if the concept of retrievability were adopted then there would be added cost and technical complication whereas the management firmly believes that the currently proposed systems are good enough to meet all contingencies. There is also the possibility that 'retrievability' could imply to the public a lack of confidence in the system. For example, if the wastes can be retrieved is this not tantamount to just another form of interim storage, could the waste be retrieved by 'unauthorised others' and does it not infer that it would be better to leave the wastes on the surface until a truly final disposal system can be developed and accepted by all. The origins of retrievability are probably based on the disposal of HLW in the form of spent nuclear fuel (not reprocessed to extract the reusable uranium and plutonium) where, with a change of circumstance there could be a resource, environment or security reason to recover the material. It was not envisaged that the HLW arising from reprocessing would ever need to be retrieved but as the UK now needs to consider the disposal of surplus plutonium as a waste, it brings the prospect of retrievability into focus. Likewise with ILW, whether shortlived or long-lived, there was no foreseeable need to ever retrieve these wastes. The need to reassess the concept of retrievability has arisen from the results of this research, which has identified the relatively short time horizons of the public and their distrust of long term geological disposal.

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Reversibility is a term now being used to describe the option of recovering ILW and HLW from repositories before they are closed. The possible needs to reverse the placement while the repository remains open could arise from technical, environmental, health or resource (in the case of spent fuel and plutonium) reasons. Reversal prior to closure, whilst unlikely, is technically more simple than retrieval after the repository is closed although it has a cost and once again puts the wastes back onto the surface. However if there is a just cause then surely reversal is a preferred action when compared to retrieval after closure. ILW and HLW systems incorporate multiple barriers such as the primary form of the waste (metallic, glass or ceramic), metal container, geological overpack (modified clays), near-field strata (salt, clay, rock) and distance from the biosphere. In most systems there is an in-built redundancy such that if a barrier fails prematurely, other barriers will compensate. One concept not yet fully explored is the issue of remediation, that is, a form of geotechnical correction if there is a premature failure of the barriers. Industry could be loath to discuss this option as it, like reversibility and retrievability, might infer from the outset that the system could be flawed. It could also infer that if there is a premature failure then industry proposes a set of band-aids that hopefully will fix the problem: a poor image. However we should consider for example the environmental corrections taking place in and around uranium mines and processing plants in the Former Soviet Union, particularly in the former East Germany and Czechoslovakia where there has been significant 'remediation'. Mine effluents have been cleaned and are now safely allowed to flow into the river systems, underground acidity is being corrected and surface tailings consolidated and covered with clean material. Perhaps we should think in terms of what would we do today faced with a radioactive leak from an underground repository. If such a leak is water born then technologies such as pressure grouting of incoming aquifers and the repository, bypass hydraulics, water quality modification, retrievability of source, down stream environmental and health monitoring could all be applicable. If the contaminants are gas born, then a system of management also needs to be prescribed. Perhaps this could be based upon underground collection manifolds, filtering, conditioning of gases, dilution and release. Any recovered gas born particles should be treated as for any other solid waste. There are geo-technical systems available for the remediation of closed coal, salt and metal mines. The proposed sequestration of carbon dioxide and other gases also illustrates some of the technologies available for remediation. There is probably a strongly held public view that all manmade artefacts will need repair sooner or later, therefore why not prescribe that from the outset. It can be shown that with today's technology remediation of a premature failure in a repository can be addressed. This is not dependent upon new and as yet to be invented technologies but of course there will be evolution in these geo-technical areas. Would the public fear less the prospect of remediation if required than they do of irreversible and irreparable placement?

5. A GENERAL HYPOTHESIS FOR WASTE DISPOSAL It has been established that: • There is a clear discontinuity between the time to the forward horizon for the public and the temporal needs for the isolation of radioactive wastes.

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•

• •

Waste disposal is a dynamic process that incorporates chemical and physical change, allowance for exhaustion of the primary containment, probable interaction with the nearfield and finally the concept that buried wastes will alter and become part of the earth. The issue of radioactive waste disposal needs to be resolved, and Reversibility, retrievability and remediability are candidate ancillary concepts.

It falls to scientists and engineers to take these parameters into consideration when designing a viable system for the disposal of radioactive waste and to win the confidence of the public. Schemes that are technically valid but cannot engender public confidence will become the deLoreans of this industry: expensive, glitzy, technically smart but unsaleable. To stimulate discussion on the evolution of an acceptable system, a general hypothesis is now proposed as shown in Figure 2. The General Hypothesis for the Geological Disposal of Waste: the Hazard-Time Relationship, where the relative hazard (to the biosphere) of a waste is plotted against time. Line 'A-E' represents the Hazard-Time (H-T) Relationship. The yaxis is divided about a line above which there is increasing hazard and below which it is proposed that there is no discernible hazard. 'A' represents the opening of the repository to receive wastes and period 'A-B' on the H-T line corresponds to the period that a repository receives waste. 'B' represents the closure of the repository, 'C' the limit of the design life of the primary containment and 'D' the minimum lifetime of the artificial and natural barriers. After 'D' it is assumed that the waste has become an inert part of the earth and constitutes no risk to the biosphere. 'E' represents the full absorption of the waste into the earth. In the event that there is an unlikely premature failure of the system, then the suggested remedies are: • ' A-B': reversibility with recovery of the waste in its primary container. • 'B-C': retrievability of the waste in its primary container or remediation • 'C-D': remediation • 'D-E': no action required as the waste is now altered and possibly absorbed into the substrate materials. This model hopes to demonstrate a technical resolution for geological disposal that engenders public acceptance. For all future generations there is a prescribed remedy to any possible default of the system during the designated isolation period. Additionally there would be a scientific and engineering certainty that the risk of a fatality in the population would never exceed 1x10"6 p. a. although this is not an expression that inspires confidence in the public. Reversibility, retrievability and remediability however may give confidence to all concerned. 6. THE HYPOTHESIS APPLIED TO ILW Figure 3. The Hazard-Time Relationship for the Geological Disposal of ILW is based on the general hypothesis but ascribes time values to the x-axis. The time value for each point is: 'A' = present, 'B' = 100 years, 'C' = 1000 years and 'D' = 5000 years although these are indicative only and can be adjusted to suit a specific classification of ILW. It follows that for HLW, the values would be in the order of A = present, B = 100 years, C = 3-5000 years and D = 100,000 years. Conceptionally, the case for public acceptance of a remedial system to contain ILW for up to 5000 years seems achievable. Public acceptance of a system to isolate

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HLW for up to 100,000 years is more problematic but is perhaps achievable with increases in geological isolation and in the design life of the primary containment. 7. CONCLUSION This paper establishes that the disposal of radioactive waste is a dynamic process encompassed in the four dimensions of space and time. It demonstrates that public perception of future time horizons is significantly shorter than the periods required for the isolation of radioactive waste. A General Hypothesis has been developed that should give the public a greater confidence in geological disposal systems. This hypothesis is not based on the traditional stated objectives that the repository will not fail or if it does then the consequences to the biosphere will be minimal, but rather that should it fail, then the placement is reversible, the wastes retrievable or the system remediable. It is hoped that these views will stimulate a fresh approach to the development of disposal systems taking into account public perceptions of space, time, risk and trust. Without public acceptance, the most elite of technical systems will not be adopted.

ACKNOWLEDGEMENTS The author's doctoral study at the School of Geography, University of Oxford is supervised by Professor Gordon Clark, Halford Mackinder Professor of Geography and is materially supported by British Nuclear Fuels plc, Tokyo Electric Power Company, Steering Committee on High-Level-Radioactive-Waste Project, Japan and Nationale Genossenschaft for die Lagerung radioaktiver Abfalle (NAGRA), Switzerland. Comments on drafts of this paper by Emma Cornish, Gerald Clark, Brian Eyre and Chris Ealing are greatly appreciated.

REFERENCES (1) Parliamentary Office of Science and Technology (November 1997), Radioactive Waste Where Next? London. (2) The House of Lords, Session 1997-98. Select Committee on Science and Technology, Management of Nuclear Waste: Written Evidence, (April 1998). London: The Stationery Office. (3) The House of Lords, Session 1998-99. Select Committee on Science and Technology, 3"" Report: Management of Nuclear Waste, (March 1999). London: The Stationery Office. (4) National Radiological Protection Board, (1989) Living With Radiation. Didcot, Oxon. (5) Duncan, I. J. 1999a. A Community that Accepts Risk Should be Rewarded. Decision Risk and Policy 4 (3): 191-199. (6) Duncan, I. J. 1999b. Some Aspects of the Relationship between Society and the Disposal of Radioactive Waste. Paper read at The Uranium Institute: Twenty-Fourth Annual Symposium, September 1999, at London.

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Figure 1. Waste Disposal Concepts

Figure 2. The General Hypothesis for the Geological Disposal of Waste: the Hazard - Time Relationship

Figure 3. The Hazard - Time Relationship for the Geological Disposal of lLW

Authors' Index B Balkey, J J Barlow, S V Borisov, V V Bradbury, D

N 57-66 105-116 183-190 191-200

P

Cave, L

Nakajima, M Netecha, M E Newstead, S

85-94 43-46, 183-190 117-128

o Orlov, Yu V

43-46, 183-190

141-150 P

Palmer, J D Daish, S R Duncan, I J

117-128 201-210

E

Edler, G R Ellis, AT

191-200 73-82

105-116

S Seddon, W Shishkin, V A Stanislavski, G A

95-104 43-46, 67-72 183-190

T

G Gabaraev, B A Green, T H

Thomson, P F G 67-72 161-170

H

Harrison, M Hey,R I

21-30 73-82

161-170 161-170

L Langley, KF Leech, N A

47-56 117-128

M Mazokin, V A McCracken, G McTagget, L Miller, T J

V Van Velzen, L P M Vasiliev, G A Vasiljev, A P

129-140 183-190 43-46

W

K

Kornyeyev, A A Krinitsyn, A P

3-20

43-46, 67-72, 183-190 31-40 73-82 153-160

Waker, C H Wieneke, R E Wilding, C R Williams, J Wood, C J

171-182 57-66 161-170 47-56 191-200