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Nuclear Graphite Waste Management

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Nuclear Graphite Waste Management Technical Committee meeting held in Manchester, United Kingdom, o 18-20 October 1999 o ro 00 CD O Nuclear Graphite Waste Management Foreword Copyright/Editorial note Contents Summary List of participants Related IAEA priced publications INTERNATIONAL ATOMIC ENERGY AGENCY Copyright © IAEA, 2001 Published by the IAEA in Austria May 2001 Individual papers may be downloaded for personal use; single printed copies may be made for use in research and teaching. Redistribution or sale of any material on the CD-ROM in machine readable or any other form is prohibited. EDITORIAL NOTE This publication has been prepared from the original material as submitted by the authors. The views expressed do not necessarily reflect those of the IAEA, the governments of the nominating Member States or the nominating organizations. The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA. The authors are responsible for having obtained the necessary permission for the IAEA to reproduce, translate or use material from sources already protected by copyrights. FOREWORD Graphite and carbon have been used as a moderator and reflector of neutrons in more than one hundred nuclear power plants, mostly in the United Kingdom (Magnox and AGRs), France (UNGGs), the former USSR (RBMK), the United States of America (HTR) and Spain. In addition many experimental and plutonium production reactors also use graphite as a neutron reflector or neutron moderator. Less frequently graphite is used for manufacturing fuel sleeves and other reactor core components. Many graphite reactors are relatively old, some of them have already been shut down. Therefore, dismantling, characterization and management of radioactive graphite waste are a problem for many IAEA Member States. Radioactive graphite management is not yet covered in IAEA publications. Only very general and limited information is available in publications dealing with decommissioning of nuclear reactors. Recognising this fact and reflecting the needs of Member States, the IAEA decided to initiate a project on characterization, treatment and conditioning of radioactive graphite from nuclear power plants. The aim is to prepare a publication dealing with all aspects of radioactive graphite waste handling, conditioning and disposal. The most significant scientific and technical information on the subject has been collected in a few countries operating nuclear power reactors with graphite components in the reactor core. Some other countries have recently been involved in an investigation of particular problems corresponding to their specific requirements. With the aim of collecting information available in various countries and reflecting it in a technical publication, the IAEA, in association with the British Nuclear Energy Society (BNES), convened a Seminar on Nuclear Graphite Waste Management. This publication contains presentations made during the seminar as well as a summary of the participants' discussions. The IAEA wishes to convey its thanks to the North West Branch of the British Nuclear Energy Society for hosting the seminar and to BJ. Marsden of AEA Technology for his principal contribution to the organization of the seminar and chairing of several principal sessions. The responsible IAEA officer was R. Burcl of the Division of Nuclear Fuel Cycle and Waste Technology. CONTENTS The uncertain future for nuclear graphite disposal: Crisis or opportunity? A.J. Wickham, G.B. Neighbour, M. Dubourg Technical assessment of the significance of Wigner energy for disposal of graphite wastes from the Windscale piles R.M. Guppy, J. McCarthy, S.J. Wisbey The description of Wigner energy and its release from Windscale pile graphite for application to waste packaging and disposal P.C. Minshall, A.J. Wickham Heat treatment of graphite and resulting tritium emissions J. Worner, W. Botzem, S.D. Preston Pyrolysis and its potential use in nuclear graphite disposal J.B. Mason, D. Bradbury Technical development of graphite waste treatment in NUPEC S. Saishu, T. Inoue Inert annealing of irradiated graphite by inductive heating W. Botzem, J. Worner Study on efficient methods for removal and treatment of graphite blocks in a gas cooled reactor S. Fujii, M. Shirakawa, T. Murakami A German research project about applicable graphite cutting techniques D. Holland, U. Quade, F. W. Bach, P. Wilk Thermal strain measurements in graphite using electronic speckle pattern interferometry S. Tamulevicius, L. Augulis, R. Augulis, V. Zabarskas, R. Levinskas, P. Poskas Investigation of graphite column bricks of Leningrad NPP unit 3 RBMK-1000 reactor after 18 years of operation P.A. Platonov, O.K. Chugunov, Ya.I. Shtrombakh, V.M. Alekseev, N.S. Lobanov, V.N. Manevski, V.IKarpukhin, IF. Novobratskaya, LA. Frolov, V.M. Semochkin, V.A. Denisov, A. V Borodin, A.A. Kalinnikov, D.E. Furkasov, Yu. V Garusov, M.A. Pavlov, A.N. Ananiev, A.V. Makushkin, V.N. Rogozin, V.P. Aleksandrov, V.D. Baldin Derivation of a radionuclide inventory for irradiated graphite-chlorine-36 inventory determination F.J. Brown, J.D. Palmer, P. Wood Radiochemical characterisation of graphite from Julich experimental reactor (AVR) B. Bisplinghoff, M. Lochny, J. Fachinger, H. Brucher Radionuclide characterization of graphite stacks from plutonium production reactors of the Siberian group of chemical enterprises A.V. Bushuev, Yu.M. Verzilov, V.N. Zubarev, A.E. Kachanovsky, I.M. Proshin, E.V. Petrova, T.B. Aleeva, A.A. Samarkin, B.G. Silin, A.M. Dmitriev, E.V Zakharova, S.I. Ushakov, I.I. Baranov, Yu.I Kabanov, E.N. Kolobova A.G Nikolaev Investigation of morphology and impurity of nuclear-grade graphite, and leaching mechanism of carbon-14 R. Takahashi, M. Toyahara, S. Maruki, H. Ueda, T. Yamamoto Management of UKAEA graphite liabilities M. Wise Radioactive graphite management at UK Magnox nuclear power stations G. Holt SUMMARY At the invitation of the British Nuclear Energy Society (BNES) the IAEA convened the seminar/workshop on nuclear graphite waste management as a part of the Technical Committee meeting (TCM) on Characterization, Treatment and Conditioning of Radioactive Graphite from Nuclear Power Plants. The seminar was held on 18-20 October 1999 in Manchester, United Kingdom and was attended by more than 55 participants from 11 countries (Austria, Japan, Russia, Ukraine, France, Germany, Spain, the United Kingdom, Lithuania, Indonesia, and the United States of America). Twenty technical papers were presented and discussed in five sessions covering various aspects of radioactive graphite waste management. Practical approaches to the solution of particular technical problems were demonstrated during the technical visit to the Sellafield site on 20 October. The seminar was followed up by a technical meeting attended by 11 participants nominated by the Member States. Information presented at the seminar was further reviewed and selected parts were incorporated in a draft technical document on characterization, treatment and conditioning of radioactive graphite from nuclear power plants. BACKGROUND Graphite has been used as a moderator and reflector of neutrons in more than 100 nuclear power plants as well as many experimental reactors and plutonium production reactors in various countries. Graphite is also used for fuel sleeves and other components, which are disposed of during operation and in some cases the volume of the graphite involved is of the same order as the core itself. Considering that the core of a typical graphite moderated reactor may contain 2000 tonnes of graphite, the volumes involved are considerable. Most of the older graphite moderated reactors are already shut down and therefore decommissioning planning and preparation represents a very serious matter. Radioactive graphite dismantling, handling, conditioning and disposal are a common part of the decommissioning activities. The radioactive graphite coming from nuclear installations has different characteristics than other radioactive waste due to its physical and chemical properties and also because of the presence of tritium and carbon-14. Even after many years of irradiation, graphite retains most of the good mechanical properties and is relatively insoluble and not otherwise particularly chemically reactive. It appears therefore to fulfil most of the general requirements for a solid radioactive waste form suitable for disposal. However, the evaluation of the radioactivity inventory of graphite moderators and other details of graphite used in nuclear reactors show that this graphite cannot be accepted by existing disposal sites without particular conditioning. Depending on the graphite source (moderator, fuel compartments details-sleeves), various radionuclides are present — mainly tritium, 14C, but also corrosion/activation products (57Co, 60Co; 54Mn; 59Ni, 63Ni; 22Na), fission products (134Cs, 137Cs; 90Sr; 152Eu, 154Eu, 155Eu; 144Ce) and small amounts of uranium and transuranium elements (238Pu, 239Pu; 241Am, 243Am). The graphite in some of the experimental and plutonium production low temperature reactors contains a considerable amount of stored Wigner energy. Unexpected
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