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,Q6LWX6LQWHULQJRI:DVWH)RUPVLQDQ8QGHUJURXQG'LVSRVDO(QYLURQPHQW Michael I. Ojovan, Fergus G.F. Gibb, William E. Lee Immobilisation Science Laboratory, Department of Engineering Materials, The University of Sheffield, Mappin Street, Sheffield, S1 3JD, UK. $%675$&7 A waste management scheme is described, which aims to utilise the ambient pressure of a disposal environment, its radiation shielding and extended time of storage to ensure reliable immobilisation of radioactive waste in a glass composite or polycrystalline matrix form. The conditions required for natural sintering of the waste form in the repository are assessed for viscous flow and grain boundary diffusion mechanisms. ,QVLWX sintering of materials in the repository creates geochemically stable materials in equilibrium with the disposal environment ensuring a higher degree of safety compared to existing approaches. &21&(37 The immobilisation and disposal of current and future nuclear waste material is essential both to assure public safety and for the future development of nuclear power programmes. Immobilising materials must ensure reliable retention of hazardous radionuclides and be capable of hosting considerable amounts of nuclear waste while preserving their retention properties for an extended period of time. Highly durable glasses, crystalline, polycrystalline and glass composite matrices are potentially suitable host matrix materials for immobilisation of nuclear waste. The most dangerous components of nuclear waste are long-lived radionuclides (actinides, rare earth elements) that require millions of years to reach natural background level of radiation. Hence the only suitable disposal option is the emplacement of waste into a deep and stable geological formation. Current nuclear waste management schemes treat immobilisation and disposal separately. The first stage comprises immobilisation of nuclear waste in a durable matrix material. For high level radioactive waste (HLW) this is a vitreous glass-like material, which utilizes borosilicate glasses (France, UK, USA, Japan) or alumina-phosphate glass (Russia). Vitrification of HLW is a costly and complex technology, but is required to ensure safe storage conditions for waste until disposal facilities become available. The disposal stage comprises emplacement of canisters with vitrified HLW into a deep underground repository. The depth of currently developed repositories (e.g. in USA, France, Japan) is within 500 – 1000 m, despite the fact that deeper repositories would provide a much higher degree of safety [1, 2]. Moreover carrying out immobilisation directly in the disposal facility (LQVLWX) can significantly diminish costs and hazards associated with predisposal immobilisation [2]. A practical example of successful employment of an LQVLWX approach is the metal matrix immobilisation of spent sealed radiation sources in borehole repositories in Russia [3]. This efficiently resolved the problem of highly radioactive sources in Russia but spent sealed sources remain an unresolved issue