Use this URL to cite or link to this record in EThOS: http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.618014
Title: Options for treatment of legacy and advanced nuclear fuels
Author: Maher, Christopher John
ISNI:       0000 0004 5352 9258
Awarding Body: University of Manchester
Current Institution: University of Manchester
Date of Award: 2014
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Abstract:
The treatment of advanced nuclear fuels is relevant to the stabilisation of legacy spent fuels or nuclear materials and fuels from future nuclear reactors. Historically, spent fuel reprocessing has been driven to recover uranium and plutonium for reuse. Future fuel cycles may also recover the minor actinides neptunium, americium and perhaps curium. These actinides would be fabricated into new reactor fuel to produce energy and for transmutation of the minor actinides. This has the potential to reduce the long lived radioactivity of the spent fuel and reprocessing high level waste, whilst also maximising energy production. To achieve these aims there are a range of materials that could be used as advanced nuclear fuels, these include metals, oxides, carbides, nitrides and composite materials, and these fuels may also be alloyed. These advanced fuels may need to be reprocessed, and as head end is the first chemical treatment step in a reprocessing plant, the issues caused by treating these advanced fuels are faced primarily by head end. Changes to the overall reprocessing specification, such as reduction in discharge authorisations for volatile radionuclides, will have the greatest impact upon head end. All these factors may lead to the introduction of pre-treatment technologies (e.g. Voloxidation) or enhanced dissolution technologies, e.g. mediated dissolution using silver(II).Literature and experimental studies show that uranium dioxide and low plutonium content MOx dissolves in nitric acid via direct and indirect nitrate reduction. The indirect nitrous acid catalysed route is kinetically most significant. The kinetics for the dissolution of uranium dioxide and 5 % plutonium MOx have been derived experimentally. Studies of the dissolution of MOx pellets in concentrated nitric acid and near boiling conditions indicate that dissolution shows a degree of mass transfer limitation. Thermodynamic studies show that the pronounced reduction in the MOx dissolution extent at 30-40% plutonium is due to the thermodynamics of the key dissolution reactions. One technology that could be used to dissolve plutonium-rich residues that are generated from the reprocessing of MOx fuels is mediated dissolution. Inactive studies using linear staircase voltammetry (LSCV) and constant current bulk electrolysis (BE) have been used to optimise a 100 ml dissolution cell. The generation of silver(II) is dependent upon silver concentration, agitation and the size of the separator membrane. Whilst the stability of silver(II) is defined by the kinetics of water oxidation, this is dependent upon a number of factors including nitric acid concentration, silver(I):(II) ratio, temperature and the rate of migration from the catholyte into the anolyte. LSCV experiments have shown that Tafel analysis confirms there is a good relationship between potential and anode current density assuming oxygen evolution and silver(I) oxidation. Kinetic modelling of the BE experiments can be used to model the silver(II) generation, steady state and decomposition due to reaction with water. The dissolution cell has been demonstrated to be capable of dissolving plutonium dioxide to 200 g.l-1 in less than 2 hours with good faradaic efficiency.
Supervisor: Schroeder, Sven; Livens, Francis Sponsor: National Nuclear Laboratory ; Sellafield Ltd ; Nuclear Decommissioning Agency ; European Comission FPVII ACSEPT project (contract number 2007-211267) ; European Space Adgency
Qualification Name: Thesis (Ph.D.) Qualification Level: Doctoral
EThOS ID: uk.bl.ethos.618014  DOI: Not available
Keywords: Uranium oxide ; Plutonium oxide ; MIxed uranium plutonium oxide ; dissolution ; head-end ; Mediated electrochemical oxidation
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