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.

The long-term behavior of nuclear glass subjected to alpha radiation by minor actinides must be investigated with a view to geological disposal. This study focuses on the effect of alpha radiation on the chemical reactivity of R7T7 glass with pure water, mainly on the residual alteration rate regime. A glass specimen doped with 0.85 wt% 239PuO2 (α emitter) is leached under static conditions in argon atmosphere at 90°C and at a high surface-area-to-volume ratio (S/V = 20 cm−1). The alteration rate is monitored by the release of glass alteration tracer elements (B, Na and Li). Radiation effects on the leached glass and its gel network are characterized by SEM and TEM analyses. Plutonium release is also measured by radiometry and its chemical oxidation state is assessed by measuring the pH and redox potential of the leachates. The results do not highlight any significant effect of alpha radiation on the residual alteration of this doped glass. This observation is consistent with SEM and TEM characterizations, which show that a protective layer can be formed under alpha radiation. Very low concentrations of soluble plutonium are measured in the leachate. These Pu releases are three orders of magnitude lower than the boron release, indicating strong plutonium retention.

The development of an iodine immobilization technique that can fix radioactive iodine in waste form for a long period and constrain its leaching into pore water is necessary in order to secure the long-term safety of geological disposal of transuranic (TRU) waste. Lead borate glass vitrified at a low temperature is regarded as a promising material for immobilizing the Iodine-129 that is recovered from spent AgI filters generated by reprocessing plants in Japan and which may have a significant effect on the long-term safety of geological disposal.

Batch leaching tests were conducted to understand glass dissolution behavior in various solutions that account for geological disposal conditions. Boron dissolved at the highest rate in all types of solutions to be used as an index element for measuring the glass dissolution rate. On the other hand, lead dissolved in these solutions at a much lower rate. These results are consistent with an electron micro-probe analysis (EPMA) of the altered glass surfaces that indicated the depletion of boron and enrichment of lead near the surfaces.

The altered glass surfaces were further examined by scanning and transmission electron microscopy (SEM/TEM) and X-ray diffraction (XRD). SEM/TEM observation showed formation of a porous altered layer consisting of fine crystallites on the pristine glass and euhedral crystals on the altered layer. XRD analysis indicated that the fine crystallites and euhedral crystals are hydrocerussite, Pb3 (CO3)2(OH) 2, which was predicted by geochemical calculation as the precipitate for the experimental system.

At Sellafield, the Post Operational Clean Out (POCO) of solids from the base of the highly active waste storage tanks, in preparation for decommissioning, will result in a high molybdenum stream which will be vitrified using the current Waste Vitrification Plant (WVP). In order to minimise the number of containers required for POCO, the high molybdenum feed could be co-vitrified by addition to reprocessing waste, using the borosilicate glass formulation currently utilised on WVP. Co-vitrification of high molybdenum feeds has been carried out using non-active simulants, both in the laboratory and on the Vitrification Test Rig (VTR) which is a full scale working replica of a WVP processing line.

In addition, a new borosilicate glass formulation containing calcium has been developed by NNL which allows a higher incorporation of molybdenum through the formation of a durable CaMoO4 phase, after the solubility limit of molybdenum in the glass has been reached. Vitrification of the high molybdenum feed in the presence of varying quantities of reprocessing waste liquor using the new glass formulation has been carried out in the laboratory. Up to ∼10 wt% MoO3 could be incorporated without any detrimental phase separation in the product glass, but increasing the fraction of reprocessing waste was found to decrease the MoO3 incorporation. Soxhlet and static powder leach tests have been performed to assess the durability of the glass products. This paper discusses the results of the vitrification of high molybdenum feeds in the presence of reprocessing liquor in both the borosilicate glass formulation currently utilised on WVP and the modified formulation which contain calcium.

The effect of γ-radiation on the mechanical properties of model UK intermediate and high level nuclear waste glasses was studied up to a dose of 8 MGy. It was determined that γ-irradiation up to this dose had no measurable effect upon the Young’s modulus, shear modulus, Poisson’s ratio, indentation hardness, or indentation fracture toughness. The absence of measurable radiation induced changes in mechanical properties was attributed to redox mediated healing of electron-hole pairs generated by γ-irradiation by multivalent transition metal ions, in particular the Fe3+ - Fe2+ couple.

Glasses in the Al2O3-B2O3-Fe2O3-Na2O-SiO2 system were produced at a temperature of 1150 °C, annealed, and examined using XRD and SEM/EDX. Surfaces of same samples were additionally heat-treated and etched with HCl. The pristine samples were X-ray amorphous and rather homogeneous except the B1 sample that contained trace crystalline phases of carnegieite/nepheline and spinel. Corrosion of these glasses via an etching treatment proceeds by a conventional mechanism with damage of their surface layers, however, the B2 glass exhibits a “drop-type” microstructure after etching that suggests occurrence of liquid-liquid phase separation.

Glasses in the Al2O3-B2O3-Fe2O3-Na2O-SiO2 system were produced at a temperature of 1150 °C, poured onto a metal plate and annealed. The nature of the structural units and their bonding in the structure were studied by infrared and Raman spectroscopic techniques. The structural network of all the glasses studied is built from major [SiO4] tetrahedra with 2-3 non-bridging oxygens (NBO). Incorporation of Fe2O3 offers a destructive effect in the glass network.

Valence state and local environment of Fe in complex glasses related to the system Al2O3-B2O3-Fe2O3-Na2O-SiO2 were studied. In all the glasses, the major fraction of Fe exists as Fe3+ ions but a minor fraction of Fe2+ ions especially in the glass with the lowest K=[SiO2]/[B2O3] ratio was also present. Average Fe—O distance in the first shell is 1.80-1.85 Å and coordination number is 4-6. The intensity due to the second sphere is rather weak demonstrating homogeneous distribution of Fe ions in the glass.

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.

A calcium phosphate ceramic waste-form has been developed at AWE for the immobilisation of chloride containing wastes arising from the pyrochemical reprocessing of plutonium. In order to determine the long term durability of the waste-form, aging trials have been carried out at PNNL. Ceramics were prepared using Pu-239 and -238, these were characterised by PXRD at regular intervals and Single Pass Flow Through (SPFT) tests after approximately 5 yrs.

While XRD indicated some loss of crystallinity in the Pu-238 samples after exposure to 2.8 x 1018 α decays, SPFT tests indicated that accelerated aging had not had a detrimental effect on the durability of Pu-238 samples compared to Pu-239 waste-forms.

Iodine filters expended after nuclear fuel reprocessing contain radioactive iodine (I-129), almost all of which exists as silver iodide (AgI). The synthetic rock technique is a solidification treatment technique using hot isostatic press (HIP), in which the alumina adsorbent base material is synthesized to form a dense solidified material (synthetic rock), and I-129 is physically confined in the form of AgI in the alumina matrix. Thus, it is necessary to understand the matrix dissolution behavior to evaluate the iodine release behavior.

Experiments involving the dissolution of the matrix were carried out under various temperatures (35–70 °C) and pH values (10–12.5) that reflect the disposal conditions. The results of the experiments showed that the dissolution rate of Al visibly increases with temperature and pH. The dissolution rate constant was calculated from the initial data assuming the dissolution of the matrix as a primary reaction. The logarithmic rate constant showed a good linear correlation with the pH and the reciprocal of temperature. The 27Al-NMR analysis of the solutions of the dissolved matrix showed that the major chemical species present in the solutions was Al(OH)4-. This indicated that the dissolution of the matrix can be described by the following equation: Al2O3 + 2OH- + 3H2O → 2Al(OH)4-. Subsequently, the empirical equation of the rate of dissolution of the matrix as a function of the temperature and pH was derived. It will be used to evaluate the iodine release behavior from the synthetic rock.

An iodine-immobilizing cement solidification process using calcium aluminate cement with gypsum additive was developed. Powdered cement solid was repeatedly immersed in ion-exchanged water with varying liquid-to-solid ratios (L/S) in accelerated dissolution tests simulating interaction with groundwater at waste disposal sites. The measured concentrations of iodine in the water were on the order of 10−5 to 10−3 mol⋅dm−3 in the entire L/S range. These concentration levels are extremely low compared with those in the case of ordinary Portland cement. Calculations with a solution equilibrium model for the cement immersed in ion-exchanged water showed that the observed iodine release profile versus integrated L/S ratio from the immersion test was explained by a dissolution model of minerals in the cement.

Single-phase calcium chlorosilicate and sodalite, two potential ceramic waste-forms for the immobilisation of CaCl2-based pyroprocessing wastes, have been fabricated at temperatures below the volatilisation point of CaCl2. Solid solutions doped with Sm3+ as an inactive analogue for trivalent actinides have been fabricated and characterised. XRD analysis shows both phases will successfully accommodate Sm3+, with the sodalite in particular remaining single-phase. Fabrication of Sm-doped calcium chlorosilicate in air results in the formation of SmOCl and Ca(Si2O5) secondary phases, however, calcination in an inert atmosphere is shown to successfully retard the formation of SmOCl allowing for higher levels of doping.

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.

Molten salt flameless oxidation of graphite is one of the prospective methods of irradiated graphite waste processing. Molten salts are capable to retain a considerable part of radionuclides, to neutralize acidic off gases, moreover spent salts could be vitrified on completion of the process. We have used thermodynamic modelling to assess the efficiency of molten salt oxidation of graphite. Equilibrium compositions of both the melt and the off gas were calculated depending on graphite content and temperature. The feasibility of decontaminating the irradiated graphite of its near-surface layers using complete molten salt oxidation was investigated in a series of laboratory experiments. As the molten salt medium used to oxidize irradiated graphite we have investigated lithium, potassium and sodium carbonates. Sodium sulphate, boron oxide, barium and potassium chromates were also used as oxidizers. Tests were carried out at 870–1270 К. The efficiency of decontamination of graphite blocks has been assessed based on the activity of 137Cs and 60Со in the samples before and after molten salt oxidation. Data obtained demonstrated the feasibility of decontamination by molten salt removal of near surface layers on irradiated graphite blocks. Decontamination rate and efficiency depend on oxidizers used and temperature of process.

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).

Laves-type intermetallic phases have been observed to be the dominant phases in a series of alloy compositions being designed for the immobilization of technetium in a metallic waste form. The dominant metals in the alloy compositions were Fe-Mo and Fe-Mo-Zr. The alloy composition, Fe-Mo-Zr, also contained Pd, Zr, Cr, and Ni. Both non-radioactive rhenium-containing and radioactive technetium-bearing alloy compositions were investigated. In the Fe-Mo series, the phases observed were Fe2Mo (C14 Laves phase) and ferrite in agreement with predictions. Both Tc and Re resided predominantly in the Laves phases. In the Fe-Mo-Zr system, the phases included hexagonal C14 with the composition (Fe,Cr)2Mo, cubic C15 phase with a (Fe,Ni)2Zr composition, and the hcp phase Pd2Zr. The observation of these phases was in agreement with predictions. Re was found in the C14 intermetallic, (Fe,Cr)2Mo. Technetium was also observed to be partitioned preferentially into the (Fe,Cr)2Mo phase; however, this phase exhibited a cubic structure consistent with the C15 structural type. The composition of Laves phases is influenced by both the atomic size and electro-negativity of the constituent elements. The long-term release behavior of technetium under nuclear waste disposal conditions may be more dependent on the corrosion characteristics of these individual Laves phases containing Tc than the other metallic phases.

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.

The dissolution of spent nuclear fuel is defined in two different time steps, i) the Instant Release Fraction (IRF) occurring shortly after water contacts the solid spent fuel and responsible of the fast release of those radionuclides that have been accumulated in the zones of the spent fuel pellet with low confinement, such as gap and grain boundaries and ii) the long term release of radionuclides confined in the spent fuel matrix, much slower and dependent on the conditions of the water that contacts the spent fuel.

Several models have been developed to date to explain the dissolution behavior of spent nuclear fuel under disposal conditions. The Matrix Alteration Model (MAM) is one of the most evolved radiolytic models describing the dissolution mechanism in which an Alteration/Dissolution source term model is based on the oxidative dissolution of spent fuel. Under deep repository conditions and at the expected of water contacting time (after 1000 years of spent fuel storage), α radiation will be the main contributor to water radiolysis. In the current study, simulations evaluating the effect of surface area on the alteration/dissolution of spent fuel matrix are performed considering different particle sizes of spent fuel and simulations integrating the actinides dissolution have been performed considering the precipitation of secondary phases.

The corrosion behavior of simulated spent nuclear fuel (SIMFUEL) was investigated using electrochemical impedance spectroscopy and solution chemistry analyses. The SIMFUEL was exposed to aerated solutions of NaCl+NaHCO3 with and without calcium (Ca) and silicate. Two SIMFUEL compositions were studied, representing spent nuclear fuel (SNF) corresponding to 3 or 6 at % burnup in terms of fission product equivalents of surrogate elements. For all tested cases, the polarization resistance increased with increased immersion time, indicating possible blocking effects due to accumulation of corrosion products on the SIMFUEL surface. The potential-pH diagram suggests formation of schoepite that may cause the increase in the polarization resistance. The addition of Ca and silicate produced no measureable change in the polarization resistance measured at the corrosion potential. The dissolution rate ranged from 1 to 3 mg/m2-day, which is similar to the range of dissolution rates for SIMFUEL and SNF reported in the literature for comparable conditions. SIMFUEL burnup did not have a major effect on the dissolution rate. Analysis of the solution chemistry shows that uranium is the dominant element dissolved in the posttest solutions, and the dissolution rates calculated from uranium (U) concentrations are consistent with the dissolution rates obtained from impedance measurements. Simulated-fission product elements (i.e., barium, molybdenum, strontium, and zirconium) dissolved from the SIMFUEL electrode at a relatively high rate. Sorption test results indicated significant sorption of U onto the oxide formed on stainless steel. Electrochemical methods were found to be effective for measuring the uranium dissolution rate in real time.