Processing of Irradiated Graphite: The Outcomes of an IAEA Coordinated Research Project
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Processing of Irradiated Graphite: The Outcomes of an IAEA Coordinated Research Project Michael I. Ojovan1 and Anthony J. Wickham2 1
Waste Technology Section, Division of Nuclear Fuel Cycle and Waste Technology, Department of Nuclear Energy, International Atomic Energy Agency, PO Box 100, Wagramerstraße 5, Vienna, A-1400 Austria [email protected] 2 Nuclear Technology Consultancy, Cwmchwefru, Llanafanfawr, Builth Wells, LD2 3PW, UK, and School of Mechanical, Aerospace and Civil Engineering, The University of Manchester, Manchester M13 9PL, UK [email protected] ABSTRACT Dismantling of old reactors and the management of radioactive graphite wastes are becoming increasingly important issues for a number of IAEA Member States. Exchange of information and research cooperation in resolving identical problems between different institutions contributes towards improving waste-management practices, their efficiency, and general safety. The IAEA Coordinated Research Project (CRP) under the title 'Treatment of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal' was conducted during 20102014 and has involved 24 organisations from ten Member States [1]. The CRP has explored both innovative and conventional methods for graphite characterisation, retrieval, treatment, and conditioning technologies and produced an IAEA technical document [2] which has identified a number of unresolved scientific and technical issues such as the need to: 1. Improve the scientific understanding required on creation, chemical form, location and release behaviour (transport models) of radionuclides; 2. Improve predictive models of radioisotope behaviour; 3. Ensure that sampling programmes are statistically representative of the totality of the graphite to be disposed of; 4. Establish an accurate radionuclide inventory; 5. Consider novel alternative dismantling and treatment strategies. The CRP promoted the exchange of technical information on R & D activities and will facilitate practical application for treatment and conditioning of graphite waste. The collaboration continues under the IAEA International Decommissioning and Predisposal Networks (IDN and IPN). INTRODUCTION Graphite is used in nuclear reactors primarily as a neutron moderator and reflector, a structural material, and a fuel-element matrix material. It has been deployed in about 250 commercial reactors such as the United Kingdom Magnox and Advanced Gas-Cooled Reactors (AGR), the French UNGG, a small number of high-temperature reactors (HTRs), the Soviet-era RBMKs, and in numerous ‘production’ reactors and materials-testing reactors. Many of those reactors have seen more than 40 years’ service, and about one half of them are shut down.
4117 Downloaded from https:/www.cambridge.org/core. Cornell University Library, on 16 May 2017 at 01:30:11, subject to the Cambridge Core terms of use, available at https:/www.cambridge.org/core/terms. https://doi.org/10.1557/adv.2017.198
Graphite is a porous, chemically inert material, resistant to corrosion, in general retaining it
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