Nuclear fuel debris generated at the Fukushima Daiichi nuclear power plant during the loss of coolant accident in 2011, still resides within the reactor units, constantly cooled by water. Until it is retrieved, the fuel debris will corrode, releasing radioactive elements into the coolant water and the ground surrounding the reactors. To predict the corrosion behaviour of these materials, and to establish parameters for experiments with U-containing and real fuel debris, the corrosion of two surrogate fuel debris materials, with a composition of Ce(1-x)ZrxO2 (x = 0.2 and 0.4), was investigated. Materials were synthesised by a wet chemistry route and pellets were sintered at 1700°C in air atmosphere. Due to the slow corrosion kinetics, aggressive conditions were applied, and corrosion experiments were performed in 9 mol.L-1 HNO3 under static conditions. The incorporation of Zr into the structure of Ce reduced the normalised dissolution rate; from (3.75 ± 0.15) × 10-6 g.m-2.d-1 to (4.96 ± 0.28) × 10-6 g.m-2.d-1 for RL(Ce) of Ce0.8Zr0.2O2 and Ce0.6Zr0.4O2, respectively.

The addition of Cr2O3 to modern UO2 fuel modifies the microstructure so that, through the generation of larger grains during fission, a higher proportion of fission gases can be accommodated. This reduces the pellet-cladding mechanical interaction of the fuel rods, allowing the fuels to be “burned” for longer than traditional UO2 fuel, thus maximising the energy obtained. We here describe the preparation of UO2 and Cr-doped UO2 using Hot Isostatic Pressing (HIP), as a potential method for fuel fabrication, and for development of analogue materials for spent nuclear fuel research. Characterization of the synthesised materials confirmed that high density UO2 was successfully formed, and that Cr was present as particles at grain boundaries and also within the UO2 matrix, possibly in a reduced form due to the processing conditions. In contrast to studies of Cr-doped UO2 synthesised by other methods, no significant changes to the grain size were observed in the presence of Cr.

A range of stoichiometric and mixed A-site cation brannerite glass-ceramics have been synthesised and characterised. The formation of UTi2O6 in glass is reliant on ensuring all uranium remains tetravalent by processing in an inert atmosphere. ThTi2O6 forms in glass under both inert and oxidising atmospheres due to the lack of other easily available oxidation states. CeTi2O6 could not be made to form within this glass system. The formation of A0.5B0.5Ti2O6 phases depends strongly on the oxidation states of the A and B cations available in the process atmosphere, with the most successful compositions having an average final oxidation state of (A,B)4+. Mixed cation brannerite compositions that formed in argon include U0.75Th0.25Ti2O6 and U0.71Ce0.29Ti2O6. Those forming in air include U0.23Th0.77Ti2O6, Th0.37Ce0.63Ti2O6, and U0.41Ce0.59Ti2O6.

A zirconolite glass-ceramic material is a candidate wasteform for immobilisation of chlorine contaminated plutonium residues, in which plutonium and chlorine are partitioned to the zirconolite and aluminosilicate glass phase, respectively. A preliminary investigation of chlorine speciation was undertaken by analysis of Cl K-edge X-ray Absorption Near Edge Spectroscopy (XANES), to understand the incorporation mechanism. Cl was found to be speciated as the Cl- anion within the glass phase, according to the characteristic chemical shift of the X-ray absorption edge. By comparison with Cl K-edge XANES data acquired from reference compounds, the local environment of the Cl- anion is most closely approximated by the mineral marialite, in which Cl is co-ordinate to 4 x Na and/or Ca atoms.

Zirconolite glass-ceramics are being developed as potential wasteforms for the disposition of Pu wastes in the UK. Previous studies utilised a variety of surrogates whilst this work uses both cold-press and sinter and hot isostatic press methods to validate the wasteform with PuO2. A cold press and sinter sample was fabricated as part of a validation study for plutonium incorporation in hot isostatically pressed (HIPed) wasteforms. The results confirmed the cold-press and sinter, achieved successful waste incorporation and a microstructure and phase assemblage that was in agreement with those expected of a HIPed equivalent. A HIP sample was fabricated of the same composition and characterised by SEM and XRD. Results were in agreement with the sintered sample and achieved complete waste incorporation into the glass-ceramic wasteform. These samples have demonstrated successful incorporation of PuO2 into glass-ceramic HIPed wasteforms proposed for processing Pu-based waste-streams in the UK.

Cerium incorporation in zirconolite glass-ceramic systems, consolidated by hot isostatic pressing, was investigated. Samples were formulated to target Ce incorporation on the Ca and / or Zr sites. Results show successful incorporation of Ce into the ceramic phase although complete waste digestion was not achieved. The formation of a Ce-bearing perovskite phase when targeting Zr substitution, was associated with reduction of Ce4+ to Ce3+. This study is part of preliminary work towards fabricating Pu-bearing glass-ceramic HIP samples.

Long half-life biologically active fission products, such as technetium-99, present particular problems for the disposal of spent nuclear fuel (SNF). Technetium is present in relatively high concentrations in fuel (approx. 1kg tonne-1 SNF) and has very high mobility in oxidizing environments. Technetium is therefore generally removed from SNF either by solvent extraction and reduction, during the PUREX process, or by sorption via ion exchange processes. Historically technetium has been disposed of via dilution and dispersion in the sea but stringent regulations now mean that the preferred long term option is immobilization in a highly stable and durable matrix. In this contribution we have looked at the synthesis of fluorite derivative crystalline host phases based on the zirconolite structure. Samples have been characterised by X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive spectroscopy (EDX), thermo-gravimetric analysis (TG), and mass spectroscopy (MS). We have used Mo as an inactive surrogate for Tc.

A preliminary investigation of the synthesis and characterization of simulant ‘lava-like’ fuel containing materials (LFCM), as low activity analogues of LFCM produced by the melt down of Chernobyl Unit 4. Simulant materials were synthesized by melting batched reagents in a tube furnace at 1500 °C, under reducing atmosphere with controlled cooling to room temperature, to simulate conditions of lava formation. Characterization using XRD and SEM-EDX identified several crystalline phases including ZrO2, UOx and solid solutions with spherical metal particles encapsulated by a glassy matrix. The UOX and ZrO2 phase morphology was very diverse comprising of fused crystals to dendritic crystallites from the crystallization of uranium initially dissolved in the glass phase. This project aims to develop simulant LFCM to assess the durability of Chernobyl lavas and to determine the rate of dissolution, behavior and evolution of these materials under shelter conditions.

Glass-ceramics were developed initially for the immobilization of miscellaneous Pu-residues at the UK’s Sellafield site from which it was uneconomic to recover Pu for reuse. Renewed interest in the immobilization of a portion of the UK PuO2 stockpile has led to glass-ceramics being evaluated for bulk Pu immobilization. The Nuclear Decommissioning Authority (NDA) in the UK have proposed hot isostatic pressing (HIP) as a potential consolidation technique for the processing of these wasteforms. In this study, zirconolite based glass-ceramics were investigated to determine an optimum formulation. The yield of zirconolite is shown to vary with glass composition and glass fraction, such that a higher Al content favours zirconolite formation. The sample preparation process is discussed to highlight the importance of a high temperature heat-treatment during sample preparation to achieve high quality HIPed wasteforms.

Fe K-edge X-ray absorption spectroscopy (XAS) was applied to study the structural response of iron phosphate glasses to atomic displacements arising from ion beam irradiation, as an analogue of α-recoil damage arising from actinide immobilization. Analysis of XAS spectra demonstrated reduction of Fe3+ to Fe2+ as a consequence of 2 MeV Kr+ and 2 MeV Au+ implantation to a fluence of 2 x 1016 ions / cm2 and 5 x 1015 ions / cm2, respectively.

This study investigates the dissolution of CeO2, an isostructural analogue for UO2 and ThO2, which was synthesized to closely approximate the microstructure of a spent nuclear fuel matrix. Dissolution of CeO2 particles was performed in simplified solutions representative of saline, near-neutral and alkaline ground waters that may be encountered in geological disposal scenarios, and in acidic medium for comparison. The normalized mass loss of cerium was found to be significantly influenced by the formation of colloidal particles, especially in the near-neutral and alkaline solutions investigated. The normalized dissolution rate, RL(Ce), k (g m-2 d-1), in these two solutions was found to be similar, but significantly lower than in a nitric acid medium. The activation energies based on the normalized release rate of cerium, at 40°C, 70°C and 90°C in each solution, were in the range of 24 ± 3 kJ mol-1 to 27 ± 7 kJ mol-1, indicative of a surface-mediated dissolution mechanism. The mechanism of dissolution was postulated to be similar in each of the solutions investigated, and further work is proposed to investigate the role of carbonate on the CeO2 dissolution mechanism.

Hot Isostatic Pressing of Cs-exchanged IONSIV IE-911 samples is shown to produce a mixture of ceramic phases, the nature and mass fractions of these have been determined by Rietveld analysis of powder X-ray diffraction data. The main Cs phase that forms is Cs2TiNb6O18, after this reaches approximately 30% of the total crystalline content the remaining Cs is partitioned into Cs2ZrSi6O15. Durability tests using the PCT-B method for 7 days at 90 °C with deionised water lead to Cs leach rates of 0.032 and 0.038 g∙m−2∙day−1 for samples exchanged to 6 and 12 wt% Cs, respectively, indicating a durable wasteform is produced.

Pyrochemical reprocessing of nuclear fuels, in which electrochemical separation of actinides and fission products is mediated by a molten alkali chloride salt (typically a LiCl-KCl eutectic) is of interest for future nuclear energy cycles. A key challenge in the management of pyrochemical reprocessing wastes is decontamination and recycling of the molten salt medium to remove entrained actinides and radioactive lanthanide fission products. Since pyrochlore oxides are promising candidates for the immobilisation of lanthanides and actinides, we sought to use the “problematic” molten salt to our advantage as a reaction medium for low temperature synthesis of titanate pyrochlores. Through control of reaction time and temperature, we demonstrated the synthesis of lanthanide pyrochlores at temperatures as low as 700 °C in 1 h, compared to 1350 °C in 36 h for conventional solid state synthesis. The importance of this study is in demonstrating the potential feasibility for decontamination of pyrochemical reprocessing wastes by simple addition of TiO2 to form lanthanide and actinide pyrochlores by rapid molten salt assisted reaction at moderate temperature.

The development of suitable waste forms for waste produced by generation IV reactors is of critical concern for future operations. To date no accepted disposal route for Tri-Structural Isotropic (TRISO) High Temperature Reactor (HTR) fuel exists. Alumino-borosilicate glass has been studied for its ability to encapsulate TRISO particle fuels. This glass was selected for its high aqueous durability. Encapsulation was achieved by cold pressing and sintering of glass powders mixed with HTR fuel. Sintering profiles capable of eliminating interconnected porosity in the composites were developed. The chemical compatibility and wetting of the glass matrix with the fuel were analysed along with the aqueous durability of the sintered glass matrix. Composites sintered under a controlled atmosphere produced unfractured monoliths with minimal chemical interaction between the glass and the TRISO particles. The Product Consistency Test (PCT) durability assessment indicated the sintered alumino-borosilicate glass was approximately an order of magnitude more durable than an equivalent R7T7 borosilicate glass. These results suggest sintered alumino-borosilicate glass-TRISO particle composites may provide a potential disposal route for spent TRISO particle fuel.

In this investigation, CeO2 analogues, which approximate as closely as possible the characteristics of fuel-grade UO2, were characterised after dissolution under a wide range of conditions. Powdered samples were subject to a range of aggressive and environmentally relevant alteration media with different solubility controls, and reacted at 70 °C and 90 °C. Dissolution kinetics were monitored through analysis of the coexisting aqueous solution. Monolith samples were monitored for development of surface defects such as pores and dissolution pits, in addition to morphological changes at grain boundaries and surface pores upon dissolution under aggressive conditions. The surfaces were analysed using confocal profilometry, vertical scanning interferometry and scanning electron microscopy. Dissolution rates were found to be greatest in low pH solutions and at higher temperatures. Preferential dissolution appears to occur at grain boundaries and on particular grains, suggesting a crystallographic control on dissolution.

Fe K edge X-ray absorption (XAS) and Fourier Transform Infra-Red (FT-IR) spectroscopies have been used to study potential structural modifications in sodium borosilicate glasses as a consequence of Kr+ irradiation. Glasses were doped with simulant waste elements and irradiated at room temperature with 450 keV Kr+ ions to a fluence of 2x1015 Kr+ ions cm-1. According to SRIM calculations, a damaged surface region approximately 400nm wide was produced. In order to probe only the damaged surface layer, XAS measurements were taken in total electron yield mode and FT-IR spectroscopy was conducted in reflectance off the glass surface. No change in Fe valence state was detected by XAS following irradiation. Reflectance FT-IR data revealed a shift to higher wavenumbers in the absorption bands located between 850 and 1100 cm-1 in the doped glasses, corresponding to bond stretching in the silicate network. Deconvolution of FT-IR spectra revealed the shift was due to polymerisation of the silicate network. Network connectivity was found to decrease in the un-doped glass, following irradiation. The results suggest an increase in silicate network connectivity by a cation mediated process, and demonstrates the successful application of surface sensitive XAS and FT-IR to the investigation of ion beam induced damage in amorphous materials.

X-ray powder diffraction data are presented for nine new phases with the tetragonal tungsten bronze structure. By comparison with the structure of the M=Nb,RE=La analogue, these contain Ba in the large pentagonal sites with coordination number 15, a range of large rare-earth cations and Bi in the twelve coordinate sites, and the smaller cations, Ti and either Nb or Ta, in the octahedral framework sites. The a lattice parameter data of the tetragonal unit cell show an approximately linear variation with trivalent ion radius, apart from Dy, whose value appears to be anomalous; data for c show little variation with trivalent ion radius.

Technetium-99 (99Tc) is a fission product produced during the burning of nuclear fuel and is particularly hazardous due to its long half life (210000 years), relatively high content in nuclear fuel (approx. 1 kg per ton of SNF), low sorption, and high mobility in aerobic environments. During spent nuclear fuel (SNF) reprocessing Tc is released either as a separate fraction or in complexes with actinides and zirconium. Although Tc has historically been discharged into the marine environment more stringent regulations mean that the preferred long term option is to immobilise Tc in a highly stable and durable matrix. This study investigated the feasibility of incorporating of Mo (as a Tc analogue) in a crystalline host matrix, synthesis by solid state synthesis under different atmospheres. Samples have been characterised with X-ray diffraction (XRD), scanning electron microscopy (SEM) and X-ray absorption spectroscopy (XAS).

Microwave dielectric heating was utilized to synthesize potential host phases for halide radioisotopes. Synthesis of Pb5(VO4)3I was successfully achieved, without the use of a sealed container, using a modified domestic microwave oven (DMO) operating at 2.45 GHz. Rapid synthesis of Pb5(VO4)3X (X = F, Cl, Br) was also achieved using a commercially available microwave muffle furnace, also operating at 2.45 GHz. The combination of rapid heating rate and inverse temperature profile characteristic of microwave heating was found to promote formation of the target phases whilst retarding the volatilization of the halide species. Pb5(VO4)3I ceramic bodies produced in the DMO exhibited a heterogeneous zoned microstructure, whereas Pb5(VO4)3X phases with X = F, Cl, and Br fabricated in the microwave muffle furnace were dimensionally uniform suggesting this could be a promising route to fabricating single phase, dense halide containing ceramics.