Garnets, A3B2C3O12, are considered to be potential host phases for the immobilization of high-level nuclear waste as they can accommodate a number of elements of interest, including Zr, Ti and Fe. The naturally occurring garnet, kimzeyite, Ca3(Zr,Ti)2(Si,Al,Fe)3O12, can contain ˜30wt% Zr. An understanding of the radiation tolerance of these materials is crucial to their potential use in nuclear waste immobilization. In this study two synthetic analogues of kimzeyite of composition Ca3Zr2FeAlSiO12 and Ca3Hf2FeAlSiO12 were monitored in situ during irradiation with 1.0 MeV Kr ions using the intermediate voltage electron microscope-Tandem User Facility (IVEM) at Argonne National Laboratory. The structure of these materials was previously determined by neutron diffraction and 57Fe Mössbauer spectroscopy. Ca3Zr2FeAlSiO12 and Ca3Hf2FeAlSiO12 have very similar structural properties with cubic Ia3d symmetry, the only significant difference being the presence of Zr and Hf, respectively, on the 6 coordinated B sites.

The structural changes induced by the changes of the uranium oxidation state in the ideal fluorite lattice of pure UO2 are discussed. Experimental results evidence strong distortions of the oxygen sub-lattice due to dynamic (at high temperature) or static (at low temperature) fluctuations of the local charges in the cationic sublattice. These changes in the oxidation state are often described using the Vegard’s law because a linear dependence of lattice parameter is observed over a wide range of compositions. Nevertheless, an ideal solid solution model cannot explain this behavior where the elastic effects are directly related to the ionic radii of the cations. Strong evidence is provided that enthalpy effects are relevant in these systems and that they are directly associated with the local structural changes observed during neutron scattering experiments.

Adsorption of uranium onto a Spanish smectite was studied, analyzing the effects of the most important parameters such as pH, ionic strength, radionuclide concentration and solid to liquid ratio. Batch sorption studies, in anoxic conditions under N2 atmosphere, were carried out on the bentonite previously purified and converted into the homoionic Na-form. In the sorption edges, two regions could be clearly distinguished. At pH lower than 5, sorption depended strongly on the ionic strength, possibly indicating the predominance of the uranyl ionic exchange process. At higher pH, sorption did not depend on the ionic strength but only on pH. The sorption behavior in this region suggested the predominance of a surface complexation mechanism. Sorption isotherms showed a non linear behavior in the concentration range used. Sorption data were interpreted using a non electrostatic standard model combining surface complexation, with the weak and strong SOH sites of the clay, and ionic exchange. The acid – base properties of the weak SOH sites were determined by potentiometric titrations. The model used was able to reproduce, in a very satisfactory way, all the data in a wide range of experimental conditions.

Sellafield Ltd operates 3 vitrification lines to convert highly active concentrate liquor arising as a waste product of reprocessing operations into glass for safe interim storage in the Vitrified Product Store (VPS) prior to long term disposal. Highly Active Liquor (HAL) is stored in Highly Active Storage Tanks and transferred to WVP in batches to the liquid stock tank. It is metered in a semi-continuous batch operation to a calciner (rotating tube furnace) where it is converted into an oxide powder (calcine). Glass frit is fed at the lower end of the calciner where it discharges into an Inconel melter vessel controlled at approximately 1100 C. The glass and calcine are melted together and then poured into a container as a batch operation. After two pours the container is allowed to cool, a lid is then fitted to the container, which after further cooling is welded to the container. This container is then cleaned and transferred to the VPS. Platinoid species containing ruthenium, rhodium and palladium present in the HAL form insoluble oxide phases in the glass product. The platinoid concentration in the glass will increase with increasing waste oxide loading to an extent that settling of the platinoids in the glass may occur, leading to heel enrichment, poor melter performance and difficulty in draining the melter. The viscosity will also increase, which may require higher melter temperatures to mix and pour the molten glass and could result in enhanced corrosion of the melter. Inactive laboratory scale experiments with different glass frit formulations have been performed to determine whether product quality could be maintained with higher platinoid concentrations. Operational envelopes with existing formulations were expanded to observe laboratory trial performance and determine any changes to resulting glass qualities. Also, glasses with high waste incorporations have been produced to test process capability and to ascertain any potential phase separation or devitrification issues that could affect either the process or product performance. Physical properties of the different glass formulations were performed to measure changes in viscosity, density and the rates of settling to examine the amount of phase separation that can occur. The results have shown that ruthenium, palladium or rhodium were insoluble in the melt and were not evenly distributed throughout the glass but clustered together. These results will be used as a basis for further development work. This paper presents some findings of these experiments.

Thermodynamic calculations of the reaction between hydrated OPC phases and saline groundwater indicate an elevated pH > 13, which is not associated with the well known initial release of the alkalis. Instead, the pH elevation is attributed to the generation of OH− accompanied with the precipitation of Friedel's salt (Ca3Al2O6·CaCl2·10H2O; AFm-Cl2) from the reaction of portlandite (Ca(OH)2;CH) and hydrogarnet (Ca3Al2O6·6H2O; C3AH6) with chloride ions from the saline groundwater. If such a reaction mechanism were to occur in the context of the geological disposal of radioactive wastes, the impact of a hyper alkaline plume on other barrier components, such as a bentonite buffer, could be significant. Experimental investigations were therefore conducted using only portlandite and hydrogarnet to represent hydrated OPC and NaCl solution to represent a saline groundwater. The pH elevation was confirmed and showed good agreement with the thermodynamic calculations. The experiments were repeated using hardened OPC paste to confirm this reaction mechanism in the presence of other hydrated OPC phases. In this case, however, the pH elevation was not as high as expected. This deviation can be explained by the residual aluminum, after being partially consumed by AFt and/or AFm, not being wholly assigned to hydrogarnet and a better agreement between the thermodynamic calculations and the experimentally measured results can be made assuming a fraction of aluminum is incorporated into the C-S-H gel phase.

Bentonite colloids released into groundwater from the buffer in a geological repository for high-level radioactive waste may influence the migration of radioactive elements in fractures of the host rock. In the present study, column experiments were carried out to investigate the migration behavior of bentonite colloids in artificially fractured granite. The objective of the study is to determine whether the colloids are filtered by the fracture, considering the effects of the fracture length and the ionic strength of transporting solutions. Results indicate that bentonite colloids are not filtered if the transporting solution is dilute (i.e., distilled water or 10-4 M NaCl solution). This is evidenced by the observed tendency for normalized colloid concentrations in the column effluent (C/C0) to rapidly approach 1. An initial increase in the breakthrough curves is also observed at early stages of migration experiments involving more concentrated solutions (10-3 M NaCl solution), but the curves later evolve to an approximate steady state with C/C0 < 1. The amount of filtered colloids tends to increase with increasing fracture length. These results suggest that bentonite colloids can be filtered by interactions between the colloids and fracture surfaces, and that these interactions are affected by the ionic strength of the transporting solution. This migration behavior is simulated by a model of colloid migration, taking into consideration the effects of colloid filtration by fracture surfaces. The information on the filtration of bentonite colloids would contribute to a more realistic assessment of colloid effects on radionuclide migration.

The diffusion of radionuclides in cementitious materials used as an engineered barrier is an important parameter in the performance assessment of the sub-surface repository system used for low-level radioactive waste disposal in Japan. In particular, organic carbon-14 and iodine-129 would provide large contributions to the dose evaluation, because of their low ability to be adsorbed on cementitious materials. In this study, the diffusion of acetate and iodide in hardened cement pastes was examined by through-diffusion experiments. Low-heat Portland cement containing 30 wt% fly ash (FAC), which is a candidate cement material for the construction of the sub-surface repository, was prepared for the diffusion experiments. The effective diffusion coefficients, De, of the trace ions for hardened FAC cement pastes were estimated to be on the order of 10-13 m2 s-1 at the beginning of the diffusion experiments. Then, the rate of diffusion of the trace ions decreased over the experimental period of 1-15 months. This is probably due to the change in the microstructure of the FAC as the result of a pozzolanic reaction. After a few months, the values of De were estimated to be on the order of 10-14 m2 s-1. These results suggest that an engineered barrier made of FAC can act as an effective barrier inhibiting the diffusion of trace ions such as organic carbon and iodine.

Thin crystals of rutile, brookite, and anatase were irradiated in-situ with 1.0 MeV Kr using the IVEM-TANDEM facility. Synthetic rutile and cassisterite (SnO2, rutile structure) remained crystalline up to 5 × 1015 ion cm-2 at 50 K. Natural brookite and anatase with low impurity levels became amorphous at 8.1 ± 1.8 × 1014 and 2.3 ± 0.2 × 1014 ions cm-2, respectively, at 50 K. Irradiation at higher temperature revealed Tc = 170 K for brookite and 242 K for anatase. Natural rutile with about 2 wt% impurities became amorphous at 9.4 ± 1.8 × 1014 ions cm-2 at 50 K and has a Tc = 207 K. The available data reveal both a structural effect in the polymorphs with low levels of chemical impurities and a chemical effect in natural rutile specimens containing up to about 1.7 wt% impurities.

The background and planned research activities are outlined for a new UK research consortium focused on Decommissioning, Immobilisation And Management Of Nuclear-wastes for Disposal (DIAMOND). This consortium is the first integrated trans-disciplinary and multi-institution academic research network in the UK, focused on nuclear waste management.

Hot isostatically pressed (HIPed) glass-ceramics for the immobilization of uranium-rich intermediate-level wastes and Hanford K-basin sludges were designed. These were based on pyrochlore-structured Ca(1-x)U(1+y)Ti2O7 in glass, together with minor crystalline phases. Detailed microstructural, diffraction and spectroscopic characterization of selected glass-ceramic samples has been performed, and chemical durability is adequate, as measured by both MCC-1 and PCT-B leach tests.

95Mo magic-angle spinning nuclear magnetic resonance (MAS NMR) spectroscopy is surprisingly sensitive to the local environment of tetrahedral molybdate species. A series of compounds related to expected crystallization products in nuclear waste glasses are probed to calibrate their spectral characteristics. Glasses formed with fast and slow quenching show a glassy peak corresponding to tetrahedral molybdate species. With slow quenching, a prominent sharp peak is observed, representing crystallinity. In sodium-borosilicate glasses with 2.5 mol% MoO3, the sharp peak corresponds to pure crystalline sodium molybdate. Cesium-sodium and lithium-sodium borosilicate glasses with Mo show crystalline peaks as well, and suggest that NMR may potentially be used to characterize mixed-cation molydates and more complex phase assemblages. While precise quantification of Mo in different phases is likely to be time-consuming, reasonable estimates can be obtained routinely, making 95Mo MAS NMR a useful tool for investigating phase separation and crystallization in model nuclear waste materials.