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.

It is pertinent to the safety case for geological disposal in the UK that the behaviour of vitrified wastes in proximity to cementitious materials is understood. In this study, vitrified simulant intermediate level nuclear waste (ILW) was subject to dissolution in a synthetic cement water solution to simulate disposal conditions. Results show that the presence of alkali / alkaline earth elements in the cementitious solution can be favourable, at least in the short-term, leading to lower dissolution rates associated with incorporation of these elements into the altered layer of the glass.

The dissolution of the United Kingdom’s vitrified high-level-waste simulant, CaZn MW28, was investigated following the Product Consistency Test-B protocol for 112 d at 90 °C and in ultra-high-quality water. Residual rate dissolution (stage II) and rate resumption (stage III), after 28 d, was observed. Thermodynamic modelling suggested that solutions were saturated with respect to Mg- and Zn-bearing phases, and the presence of Mg- and Zn-smectite clays was tentatively observed. The formation of these phases was concurrent with a significant increase in the dissolution rate, similar to Stage III behavior seen in other nuclear waste simulant glass materials, indicating that the addition of Mg and Zn to high-level-waste glass (7.3 wt. % combined) significantly influences the dissolution rate.

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.

Nuclear wastes generated from reprocessing of used nuclear fuel tend to contain a large fraction of rare earth (RE, e.g., Nd3+), transition (TM, e.g., Mo6+, Zr4+), alkali (A, e.g., Cs+), and alkaline earth cations (AE, e.g., Ba2+, Sr2+). Various strategies have been considered for immobilizing such waste streams, varying from nominally crystal-free glass to glass-ceramic to multi-phase ceramic waste forms. For glass and glass-ceramic waste forms, the added glass-forming system is generally alkali-alkaline earth-aluminoborosilicate (i.e., Na-Ca-Al-B-Si oxide). In a US-UK collaborative project, summarized here, we investigated the glass structure and crystallization dependence on compositional changes in simulated nuclear waste glasses and glass-ceramics. Compositions ranged in complexity from five – to – eight oxides. Specifically, the roles of Mo and rare earths are investigated, since a proposed glass-ceramic waste form contains crystalline phases such as powellite [(AE,A,RE)MoO4] and oxyapatite [(RE,AE,A)10Si6O26], and the precipitation of molybdenum phases is known to be affected by the rare earth concentration in the glass. Additionally, the effects of other chemical additions have been systematically investigated, including Zr, Ru, P, and Ti. A series of studies were also undertaken to ascertain the effect of the RE size on glass structure and on partitioning to crystal phases, investigating similarities and differences in glasses containing single RE oxides of Sc, Y, La, Ce, Nd, Sm, Er, Yb, or Lu. Finally, the effect of charge compensation was investigated by considering not only the commonly assessed peralkaline glass but also metaluminous and peraluminous compositions. Glass structure and crystallization studies were conducted by spectroscopic methods (i.e., Raman, X-ray absorption, nuclear magnetic resonance (NMR), optical absorption, photoluminescence, photoluminescence excitation, X-ray photoelectron spectroscopy), microscopy (i.e., scanning electron microscopy, transmission electron microscopy, electron probe microanalysis), scattering (i.e., X-ray and neutron diffraction, small angle measurements), and physical characterization (i.e., differential thermal analysis, liquidus, viscosity, density). This paper will give an overview of the research program and some example unpublished results on glass-ceramic crystallization kinetics, microstructure, and Raman spectra, as well as some examples of the effects of rare earths on the absorption, luminescence, and NMR spectra of starting glasses. The formal collaboration described here has resulted in the generation of a large number of results, some of which are still in the process of being published as separate studies.

The forward rate of dissolution of the International Simple Glass (ISG) was determined under alkaline conditions at 40 °C using the Single Pass Flow Through (SPFT) method. Forward rates were consistent with those obtained in the literature for this glass composition. The formation of altered gel layers and surface pits was observed on the surface of glass particles, especially at the very highest pH values, despite the application of high flow rates to prevent the build-up of solubility limiting phases. These features could be attributed to preferential localized dissolution at sites with a higher alkali concentration or from a separate, less durable, vitreous phase. These results may indicate that surface pit and altered gel formation occurs under the forward rate of dissolution as imposed by the SPFT method, particularly for simplified borosilicate glass materials.

We revise the data fitting in our original paper [The dissolution rates of simulated UK Magnox - ThORP blend nuclear waste glass as a function of pH, temperature and waste loading, Miner. Mag.79 (2015) 1529–1542]. The intrinsic rate constant data were calculated incorrectly, the corrected data are presented herein. To support the corrected analysis we have also taken the opportunity to report some additional 90°C data. The conclusions of the original paper remain sound.

For the UK geological disposal facility (GDF) concept in a high strength crystalline rock, Nirex Reference Vault Backfill (NRVB) has been considered to use as a cementitious backfill material. The lifetime performance of this high pH material is required to be extremely long, and as a consequence of the interaction with the geological environment (e.g. groundwater), the backfill material will evolve and age with time. In this paper, we present the results of a leaching experiment, where NRVB is placed in contact with three different groundwater compositions representative of granitic, clay and saline groundwater for 35 days. Some differences were observed related with the mineralogy. X-ray Diffraction results showed the formation of more ettringite, mainly in the NRVB samples leached in clay and saline groundwater. An increase of the pH was also observed in the NRVB samples leached with the three different ground waters, being these results in accordance with the reactive transport modelling performed.

The dissolution of a soda-lime silicate glass in two cement leachate compositions, Young Cement Water (YCW) and Ca(OH)2, was investigated, as an analogue for dissolution of vitrified nuclear waste in a cementitious geological disposal facility. Dissolution was performed at repository temperatures (50°C) and under CO2-exclusion. Dissolution rates were observed to be a factor of 20 times higher in YCW than in Ca(OH)2, as result of the high potassium content of YCW solutions. The precipitation of the zeolite phase, K-phillipsite (K(Si,Al)8O16·6H2O), is thought to be responsible for elevated dissolution rates. Conversely, in Ca(OH)2 solutions, the precipitation of calcium- and silica-containing phases, such as tobermorite (Ca5Si6O16(OH)·4H2O), acted to reduce rates of dissolution by forming a barrier to diffusion. These results show that dissolution of vitrified nuclear waste materials in a cementitous repository may be significant during the early stages of cement leaching in groundwater.

This study details the characterization of a glass sample exposed to hyperalkaline water and calcium-rich sediment for an extended time period (estimated as 2 - 70 years) at a lime (CaO) waste site in the UK. We introduce this site, known as Peak Dale, in reference to its use as a natural analogue for nuclear waste glass dissolution in the high pH environment of a cementitious engineered barrier of a geological disposal facility. In particular, a preliminary assessment of alteration layer chemistry and morphology is described and the initiation of a long-term durability assessment is outlined.

This study investigates the use of ‘simple’ glasses, comprising six components, to represent the structure of complex LAW glasses proposed for Immobilized Low Activity Wastes from the Hanford site in the USA. The 18 elements present in ILAW glasses LAW A44, ORP LB2, and LAW A23 were represented by Al2O3, B2O3, CaO, Na2O, SiO2, and ZrO2 according to their coordination chemistry and their roles as network formers and modifiers. The dissolution behavior of each ‘simple’ glass was compared to its corresponding candidate “complex” LAW glass through PCT-B tests. Significant differences were observed; the durability of complex glasses was concluded to be LAW A44 > ORP LB2 ≥ LAW A23 whereas in their simplified versions the order was LAW A44 > LAW A23 > ORP LB2. These results are discussed in relation to compositional differences and highlight the importance of minor glass components in controlling glass durability. The implications of these results for the use of simplified glass compositions are discussed.

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.

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.

Technetium-99, a β-emitting radioactive fission product of 235U, formed in nuclear reactors, presents a major challenge to nuclear waste disposal strategies. Its long half-life (2.1 x 105 years) and high solubility under oxic conditions as the pertechnetate anion [Tc(VII)O4] is particularly problematic for long-term disposal of radioactive waste in geological repositories. In this study, we demonstrate a novel technique for quantifying the transport and immobilisation of technetium-99m, a γ-emitting metastable isomer of technetium-99 commonly used in medical imaging. A standard medical gamma camera was used for non-invasive quantitative imaging of technetium-99m during co-advection through quartz sand and various cementitious materials commonly used in nuclear waste disposal strategies. Spatial moments analysis of the resulting 99mTc plume provided information about the relative changes in mass distribution of the radionuclide in the various test materials. 99mTc advected through quartz sand demonstrated typical conservative behaviour, while transport through the cementitious materials produced a significant reduction in radionuclide centre of mass transport velocity over time. Gamma camera imaging has proven an effective tool for helping to understand the factors which control the migration of radionuclides for surface, near-surface and deep geological disposal of nuclear waste.

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.