Abstract
The release of radionuclides from spent fuel in a geological disposal facility is controlled by the surface mediated dissolution of UO2 in groundwater. In this study we investigate the influence of reactive surface sites on the dissolution of a synthesized CeO2 analogue for UO2 fuel. Dissolution was performed on the following: CeO2 annealed at high temperature, which eliminated intrinsic surface defects (point defects and dislocations); CeO2-x annealed in inert and reducing atmospheres to induce oxygen vacancy defects and on crushed CeO2 particles of different size fractions. BET surface area measurements were used as an indicator of reactive surface site concentration. Cerium stoichiometry, determined using X-ray Photoelectron Spectroscopy (XPS) and supported by X-ray Diffraction (XRD) analysis, was used to determine oxygen vacancy concentration. Upon dissolution in nitric acid medium at 90 °C, a quantifiable relationship was established between the concentration of high energy surface sites and CeO2 dissolution rate; the greater the proportion of intrinsic defects and oxygen vacancies, the higher the dissolution rate. Dissolution of oxygen vacancy-containing CeO2-x gave rise to rates that were an order of magnitude greater than for CeO2 with fewer oxygen vacancies. While enhanced solubility of Ce(3+) influenced the dissolution, it was shown that replacement of vacancy sites by oxygen significantly affected the dissolution mechanism due to changes in the lattice volume and strain upon dissolution and concurrent grain boundary decohesion. These results highlight the significant influence of defect sites and grain boundaries on the dissolution kinetics of UO2 fuel analogues and reduce uncertainty in the long term performance of spent fuel in geological disposal.
Highlights
Spent nuclear fuel is a heterogeneous ceramic material composed primarily of UO2, with a minor component of actinides and fission products
Dissolution rate investigation was performed on samples annealed at 600 and 1000 °C; both were representative of the removal of “sharp edges” and intrinsic surface defects, but samples annealed at 1000 °C represented particles with a significantly lower surface area
Expanding eq 4 to account for the factors that are expected to contribute to the dissolution kinetics a first order dissolution rate expression can be derived for the CeO2 samples investigated in this study α)A4(1
Summary
Spent nuclear fuel is a heterogeneous ceramic material composed primarily of UO2, with a minor component of actinides and fission products. The internationally supported strategy for the safe disposal of this nuclear waste material is within a geological disposal facility,[1] where the release of radionuclides to the environment will be dominated by the interaction of the UO2 matrix with groundwater. To assess the long-term performance of the geological disposal site toward the containment of radionuclides, a safety case is being prepared, which assesses the mechanism and kinetics of fuel dissolution; a large number of laboratory experiments have been focused on determining the dissolution kinetics of UO2 under a wide variety of temperature and redox conditions.[2,3] dissolution rates determined from natural uraninite (UO2) ore weathering indicate that rates measured in the laboratory are orders of magnitude greater than in nature.[4,5] This phenomenon has been widely considered and is attributed to a number of factors including artifacts of specimen preparation, solution saturation state, and changes in surface area with time.[6] Clearly, it is important to the development of a robust postclosure safety case for spent fuel disposal that such uncertainty is reduced
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