Treatment of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal: Problem and Solutions
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Treatment of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal: Problem and Solutions 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 2 Nuclear Technology Consultancy, PO Box 50, Builth Wells, LD2 3XA, UK, and School of Mechanical, Aerospace and Civil engineering, The University of Manchester, Manchester M13 9PL, UK ABSTRACT An overview is given of an International Atomic Energy Agency Coordinated Research Project (CRP) on the treatment of irradiated graphite (i-graphite) to meet acceptance criteria for waste disposal. Graphite is a unique radioactive waste stream, with some quarter-million metric tons worldwide eventually needing to be disposed of. The CRP has involved 24 organizations from 10 Member States. Innovative and conventional methods for i-graphite characterization, retrieval, treatment and conditioning technologies have been explored in the course of this work, and offer a range of options for competent authorities in individual Member States to deploy according to local requirements and regulatory conditions. INTRODUCTION Graphite is a porous, chemically inert material, highly conductive and resistant to corrosion, in general retaining its properties after exposure to an intense radiation field and at high temperatures. It does, however, undergo structural changes as a result of exposure to fast neutrons, resulting in dimensional change of components and significant changes in their mechanical and physical properties. In addition, certain impurities, together with the 1.1% of 13C naturally present in the graphite, become activated through interactions with slow neutrons and thus present a significant radiation hazard post-exposure which must be accommodated in subsequent dismantling and disposal. Finally, in reactors where the graphite has been exposed to an oxidising coolant, some degree of oxidation of the material induced by the ionizing radiation field will have taken place, potentially affecting its strength. Graphite is used in reactors as a neutron moderator and reflector, a structural material, and a fuel-element matrix material. It has been deployed in about 250 uranium- (or UO2)graphite 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. Most of those reactors are now quite old, with many already shutdown. The ability to dismantle and remove graphite stacks has already been demonstrated in a small number of different reactor designs e.g. at Fort St. Vrain (prototype HTR) in the USA, the air-cooled Brookhaven research reactor, also in the USA, and the GLEEP research reactor and Windscale prototype AGR in the UK. The resulting irradiated graphite waste (often referred to within the industry as i-gra
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