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ABSTRACTZirconolite (CaZrTi2O7) has been identified as a candidate ceramic wasteform for the immobilisation and disposal of Pu inventories, for which there is no foreseen future use. Here, we provide an overview of relevant zirconolite solid solution chemistry with respect to Ce, U and Pu incorporation, alongside a summary of the available literature on zirconolite aqueous durability. The zirconolite phase may accommodate a wide variety of tri- and tetravalent actinide and rare-earth dopants through isovalent and heterovalent solid solution, e.g.CaZr1–xPuxTi2O7or Ca1–xPuxZrTi2–2xFe2xO7. The progressive incorporation of actinides within the zirconolite-2M parent structure is accommodated through the formation of zirconolite polytypoids, such as zirconolite-4M or 3T, depending on the choice of substitution regime and processing route. A variety of standardised durability tests have demonstrated that the zirconolite phase exhibits exceptional chemical durability, with release rates of constituent elements typically <10−5gm−2·d−1. Further work is required to understand the extent to which polytype formation and surrogate choice influence the dissolution behaviour of zirconolite wasteforms.

Zirconolite (CaZrTi2O7) has been identified as a candidate ceramic wasteform for the immobilisation and disposal of Pu inventories, for which there is no foreseen future use. Here, we provide an overview of relevant zirconolite solid solution chemistry with respect to Ce, U and Pu incorporation, alongside a summary of the available literature on zirconolite aqueous durability. The zirconolite phase may accommodate a wide variety of tri- and tetravalent actinide and rare-earth dopants through isovalent and heterovalent solid solution, e.g. CaZr1–xPuxTi2O7or Ca1–xPuxZrTi2–2xFe2xO7. The progressive incorporation of actinides within the zirconolite-2M parent structure is accommodated through the formation of zirconolite polytypoids, such as zirconolite-4M or 3T, depending on the choice of substitution regime and processing route. A variety of standardised durability tests have demonstrated that the zirconolite phase exhibits exceptional chemical durability, with release rates of constituent elements typically <10-5gm-2·d-1. Further work is required to understand the extent to which polytype formation and surrogate choice influence the dissolution behaviour of zirconolite wasteforms.

Technetium-99 immobilization in low-temperature nuclear waste forms often relies on additives that reduce environmentally mobile pertechnetate (TcO4–) to insoluble Tc(IV) species. However, this is a short-lived solution unless reducing conditions are maintained over the hazardous life cycle of radioactive wastes (some ∼10,000 years). Considering recent experimental observations, this work explores how rapid formation of ettringite [Ca6Al2(SO4)3(OH)12·26(H2O)], a common mineral formed in cementitious waste forms, may be used to directly immobilize TcO4–. Results from ab initio molecular dynamics (AIMD) simulations and solid-phase characterization techniques, including synchrotron X-ray absorption, fluorescence, and diffraction methods, support successful incorporation of TcO4–into the ettringite crystal structure via sulfate substitution when synthesized by aqueous precipitation methods. One sulfate and one water are replaced with one TcO4–and one OH–during substitution, where Ca2+-coordinated water near the substitution site is deprotonated to form OH–for charge compensation upon TcO4–substitution. Furthermore, AIMD calculations support favorable TcO4–substitution at the SO42–site in ettringite rather than gypsum (CaSO4·2H2O, formed as a secondary mineral phase) by at least 0.76 eV at 298 K. These results are the first of their kind to suggest that ettringite may contribute to TcO4–immobilization and the overall lifetime performance of cementitious waste forms.

The release of radionuclides from spent fuel in a geological disposal facility is controlled by the surface mediated dissolution of UO2in groundwater. In this study we investigate the influence of reactive surface sites on the dissolution of a synthesized CeO2analogue for UO2fuel. Dissolution was performed on the following: CeO2annealed at high temperature, which eliminated intrinsic surface defects (point defects and dislocations); CeO2-xannealed in inert and reducing atmospheres to induce oxygen vacancy defects and on crushed CeO2particles 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 CeO2dissolution rate; the greater the proportion of intrinsic defects and oxygen vacancies, the higher the dissolution rate. Dissolution of oxygen vacancy-containing CeO2-xgave rise to rates that were an order of magnitude greater than for CeO2with fewer oxygen vacancies. While enhanced solubility of Ce3+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 UO2fuel analogues and reduce uncertainty in the long term performance of spent fuel in geological disposal.

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 CeO2and ThO2, spent nuclear fuel analogues that approximate as closely as possible the microstructure characteristics of fuel-grade UO2but 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.