RESUMO
The immobilisation and disposal of fissile materials from civil and defence nuclear programmes requires compatible, passively safe and proliferation resistant wasteforms. In this study, we demonstrate the application of an albite glass-zirconolite ceramic material for immobilisation of chloride contaminated plutonium oxide residues in the United Kingdom. The chlorine solubility limit in the albite glass phase was determined to be 1.0 ± 0.1 wt%, above the maximum envisaged chorine inventory of 0.5 wt%, attainable at a 20 wt% PuO2 incorporation rate within the ceramic. Cl K-edge of X-ray Absorption Near Edge Spectroscopy (XANES) was exploited to confirm partitioning of Cl to the glass phase, speciated as the chloride anion, with exsolution of crystalline NaCl above the chlorine solubility limit. Combinatorial fitting of Cl XANES data, utilising a library of chemically plausible reference spectra, demonstrated the association of Cl with Na and Ca modifier cations, with environments characteristic of the aluminosilicate chloride minerals eudialyte, sodalite, chlorellestadite and afghanite. Adventitious incorporation of Ca, Zr and Ti within the albite glass phase apparently assists chlorine solubility, by templating a local chemical environment characteristic of the mineral reference compounds. The partitioning of Ce, as a Pu analogue, within the glass-ceramic was not adversely impacted by incorporation of Cl. The significance of this research is in demonstrating the compatibility of the glass-ceramic wasteform toward Cl solubility at the expected incorporation rate, below the determined solubility limit. Thus, an upstream heat treatment facility to remove chloride contamination, as specified in the current conceptual flowsheet, would not be required from the perspective of wasteform compatibility, thus providing scope to de-risk the technology roadmap and reduce the projected capital and operational plant costs.
RESUMO
In the safety case for the geological disposal of nuclear waste, the release of radioactivity from the repository is controlled by the dissolution of the spent fuel in groundwater. There remain several uncertainties associated with understanding spent fuel dissolution, including the contribution of energetically reactive surface sites to the dissolution rate. In this study, we investigate how surface features influence the dissolution rate of synthetic CeO2 and ThO2, spent nuclear fuel analogues that approximate as closely as possible the microstructure characteristics of fuel-grade UO2 but are not sensitive to changes in oxidation state of the cation. The morphology of grain boundaries (natural features) and surface facets (specimen preparation-induced features) was investigated during dissolution. The effects of surface polishing on dissolution rate were also investigated. We show that preferential dissolution occurs at grain boundaries, resulting in grain boundary decohesion and enhanced dissolution rates. A strong crystallographic control was exerted, with high misorientation angle grain boundaries retreating more rapidly than those with low misorientation angles, which may be due to the accommodation of defects in the grain boundary structure. The data from these simplified analogue systems support the hypothesis that grain boundaries play a role in the so-called "instant release fraction" of spent fuel, and should be carefully considered, in conjunction with other chemical effects, in safety performance assessements for the geological disposal of spent fuel. Surface facets formed during the sample annealing process also exhibited a strong crystallographic control and were found to dissolve rapidly on initial contact with dissolution medium. Defects and strain induced during sample polishing caused an overestimation of the dissolution rate, by up to 3 orders of magnitude.