High-level Waste Glass Research Articles

Influence of temperature and flow rate on alteration layer structure and leaching properties of simulated sulfur-containing borosilicate HLW glass under multi-field coupling effects

During long-term deep geological disposal process, high-level liquid waste (HLW) glass may be exposed to “Thermodynamics-Hydrodynamics-Mechanics-Chemistry” (THMC)multi-field coupling effects. The leaching properties of sulfur-containing borosilicate HLW glass at different temperatures (40–150 °C) and different flow rates (0.01–0.10 mL/min) under THMC coupling effects were studied in this work. The alteration layer structure of waste glass surface was also characterized. The results show that honeycomb-shaped phyllosilicate alteration layer forms after leaching in simulated groundwater and white block-shaped BaSO4 crystals appear on the glass surface at 90 °C after 28 d. The amount of BaSO4 crystals increases obviously at 150 °C. The leaching rates of B, Si, Cs and U elements increase with increasing temperature and flow rate. Particularly, the leaching rates of B, Si, Cs, and U under THMC coupling effects (90 °C, 0.01 mL/min, 10 MPa, pH=8.05) after 28 d in the simulated groundwater are 1.10 × 10–1, 1.73 × 10–1, 2.91 × 10–1, 3.79 × 10–1 g·m-2·d-1, respectively.

Evidence of nuclide migration from high-level radioactive waste glass via geogas

Characteristics of high-level radioactive waste transport in glass as gases were studied. A transmission electron microscope was used to observe the morphology of the substance after penetrating the high-level radioactive waste glass and the bentonite soil column, and the composition of the substance was analyzed using electron probe energy spectroscopy. The results show that the radionuclides in the high-level radioactive waste glass were transported out of the glass, the bentonite was penetrated by the geogas, and the transported substance occurred as nanometer-size particles and particle aggregates.

Statistical glass structure gene modeling on liquidus temperature of high level waste glasses

Statistical glass structure gene modeling (GSgM) on liquidus temperature (TL) of simulated high level waste (HLW) glass is detailed in this article. TL values of a set of HLW borosilicate glasses were experimentally determined by x-ray diffraction (XRD) method with 24 h heat-treated glass in gradient temperature furnace, and the primary crystalline phase was CaMoO4. Integral peak areas obtained from the FTIR peak fitting were used as the structural data, and the statistical TL-structure model was first established. The S-P (TL) model was shown to have a high accuracy, i.e., R2 0.978, P value 0.0009 and RMSE 7.6 °C. Model validation showed the predicted and measured TL values were 960.4 °C and 965 °C, respectively. Model iteration further reduced P value to 0.0001 and RMSE to 7 °C. Relationship between composition (C) and the glass network structure (S) was then simulated, and composition-FTIR structure-TL analysis indicated that SiO2, ZnO, rB3+, MoO3, ZrO2, B2O3 and rare earth affect TL by influence the structural units Si–O–Si rocking vibrations, νs (Si–O–Si) in [SiO4] and νs ([BO3] or BO2O−). Results showed that GSgM offers an alternative approach to simulate TL of complex glass with limited experimental data, providing a plausible linkage between composition, structure and property.

Volatilization of sodium and boron from nuclear waste glass and associated effects on glass structure and thermal stability

During nuclear waste vitrification, loss of sodium (Na) and boron (B) occurs, as these elements are highly volatile at high temperatures, which affects the properties of the glass products. Moreover, the evaporation of volatile nuclides (Cs and Tc) occurs together with that of Na and B, which may cause environmental hazards. Thus far, only a few studies have focused on the volatilization of Na and B from nuclear waste glass and loss-related effects on glass properties. In this study, we investigated the volatilization behaviors of Na and B from a simulated high-level waste glass as functions of heating temperature and dwelling duration, and evaluated the effects of Na and B loss on the glass structure and thermal stability. The results showed that volatilization of Na occurs predominantly through the diffusion-controlled process at a temperature range of 950 °C to 1100 °C, and owing to its higher frequency factor, Na diffuses and volatilizes at a faster rate than that of B. Volatilization of B occurs by both a diffusion-controlled process from bulk to the surface and a chemical reaction process on the surface. Based on the data obtained regarding the composition of Na and B and structure of the glass, a hypothetical model was proposed to explain the volatilization behaviors of Na and B from a structural viewpoint. As the loss of Na and B increases, the amount of BO4− reduces, whereas the amount of BO3 increases; finally, the Si-O network becomes more polymerized and has less thermal stability, resulting in the crystallization of the glass at elevated temperatures. These results would serve as the basis to estimate the amount of Na and B loss and the properties of the glass melt during nuclear waste vitrification, especially in melter idling.

On the effect of Al on alumino-borosilicate glass chemical durability

The chemical durability of borosilicate glass used to confine nuclear wastes is known to vary nonlinearly with their composition, making glass dissolution rate predictions difficult. Here, we focus on the effect of Al, an important oxide of these materials. The initial and residual glass dissolution rates were investigated through experiments conducted at 90 °C and pH 9. Our results show that low Al content glasses dissolve initially faster than glasses with higher Al content, but quickly achieve a lower residual rate due to the rapid formation of a more passivating gel. Nevertheless, the high Al content gel will eventually achieve passivation with a slower reorganization. The strengthening effect of Al on the Si-O bond hydrolysis will result in more stable gels, suggesting that slightly higher Al content in the gel will improve nuclear high-level waste glass durability in geological disposal conditions.

Study on the benefit of HLLW partitioning on the high-level waste glasses from the viewpoint of waste management

The volume and characteristics of heat-generating waste glasses produced through spent fuel reprocessing PUREX process + HLLW partitioning TRPO process are evaluated as a function of burn-up and cooling period of spent fuel in this study. The waste loading of glass is assumed to be restricted by the heat generation rate, MoO3 content, and noble metal content. The results show that: 1) For spent fuel of 8–20 years cooling, the volume of waste glasses after HLLW partitioning is 39%–25% of that before partitioning. 2) After removing >99.967% of α-nuclides in HLLW through TRPO process, the TRPO raffinate stream can be immobilized into waste glasses. After 95 years of decay storage, intermediate-depth disposal rather than geological disposal of TRPO raffinate waste glasses could be engineered. 3) Separation of Sr and Cs from TRPO raffinate hardly reduces the volume of waste glasses. SrCs separation before vitrification is unnecessary. 4) The insoluble residues generated during the shearing and dissolution of spent fuel may not be suitable for mixing with the TRPO raffinate stream, nor for immobilizing into a borosilicate glass matrix. Alternative solutions rather than vitrification are encouraged.

Effects of vapor hydration and radiation on the leaching behavior of nuclear glass

During the long-term disposal of high-level nuclear waste glass in the geological disposal facility, the nuclear glass could be exposed to an unsaturated vapor environment prior to the direct contact with groundwater. Additionally, glass will experience damages due to self-irradiation. In this study, experiments were performed to investigate the chemical durability of glass after being hydrated and irradiated. This study proposes an attempt to investigate the impact of ionizing effect by applying alpha or gamma radiations to the pre-hydrated glass. The impact on chemical durability is evaluated by leaching the hydrated and irradiated glass samples. With the studied doses, 8 Giga Gy of alpha radiation or 100 kGy of gamma radiation, no significant modification of chemical durability is observed upon irradiation. On the other hand, leaching of pre-hydrated glass shows the increase of the initial release of Si, which is considered to be related to the dissolution of secondary phases. Furthermore, the total dissolution of hydrated layer is observed, which would result in the release of radionuclides at the same rate as Si does and thus needs further investigation.

Potential methods to reduce the thermal impact of vitrified waste from spent mixed oxide fuel on geological repositories

Compared to the vitrified, high-level radioactive waste (HLW) from spent UO2 fuel (UO2-HLW glass), that of mixed oxide (MOX) fuel (MOX-HLW glass) exhibits high heat generation over a long period of time. Thus, to reduce the associated waste volume and repository footprint, it is essential to prevent bentonite buffer alteration by mitigating heat generation. In this study, we evaluated five potential methods of decreasing the bentonite buffer temperature in a geological repository for MOX-HLW glass. Of these five methods, the separation of minor actinides (MA) from high-level liquid waste (HLLW) was the most effective. This is because the main nuclide involved in heat generation in MOX-HLW glass is Am-241. An MA separation rate of 81% was required to maintain the bentonite buffer temperature below the Japanese upper disposal limit of 100 °C. Thus, it is important to measure the amount of Am-241 when disposing of MOX-HLW glass. Assuming that only Am-241 was disposed of, the maximum Am-241 waste loading under a bentonite buffer temperature of 100 °C was 0.33 wt%, equating to the disposal of only 1.32 kg of Am-241 per 400 kg of MOX-HLW glass. Thus, there is a need for new methods of treating and disposing of Am-241.Graphical abstractRelationship between bentonite buffer temperature and (a) minor actinide (MA) separation ratio from high-level liquid waste (HLLW), (b) mixed oxide (MOX) high-level radioactive waste (HLW) glass and UO2 HLLW mixing ratio, (c) waste loading in MOX-HLW glass, (d) MOX-HLW glass storage time, and (e) canister footprint, representing five methods of mitigating heat generation from disposed MOX-HLW glass.

Crystallization of molybdenum oxide phase from simulated high-level waste glass under slow cooling

Crystallization of molybdenum oxide phase from simulated high-level waste glass under slow cooling

Optimal waste loading in high-level nuclear waste glass from high-burnup spent fuel for waste volume and geological disposal footprint reduction

The effects of waste loading in vitrified high-level nuclear waste (HLW) from the reprocessing of high-burnup spent fuel (56 GWd/tHM) on waste volume reduction (i.e., number of HLW canisters produced) and decay heat generation were investigated to minimize the repository footprint. As the waste loading increases, the number of canisters produced decreases; however, the decay heat and subsequent footprint per canister increase. The best estimate waste loading observed was 23 wt% (including 10 wt% Na2O), wherein the repository footprint minimizes to 60.0 m2/tHM. These results are slightly higher than those for standard HLW (55.5 m2/tHM). However, upon comparing the repository footprint with electric power generation, the benefit of optimization was that the footprint for high-burnup HLW (131 m2/TWh) was 13% smaller than that of standard HLW (151 m2/TWh).Graphical abstractRepository footprint of the vitrified HLW as a function of waste loading for different cooling times. a, b Footprint per tons of heavy metal (tHM), c, d footprint per electricity generation (tera watt-hour, TWh). a, c 45 GWd/tHM, b, d 56 GWd/tHM.

Optimization of waste management for mixed oxide fuel in light-water reactors: Evaluation of the effectiveness of blending spent uo2 and mixed oxide fuels in reprocessing

This study evaluated the impact of using mixed oxide fuel in light- water reactors (MOX-LWRs) from the viewpoint of waste management. As one of the challenges of MOX-LWR fuel is the high decay heat, one reasonable option for waste management may be to reduce the heat generation of high-level radioactive waste (HLW) glass by blending spent MOX and UO2 fuels. Both the thermal properties of the HLW glass and the thermal impact on the repository are examined. The results indicate that introducing blended UO2–MOX HLW glass (blended HLW) as a technical option can control the heat generation of the HLW glass and the thermal impact on the repository, which then lead to a reduction of 5–30% in the emplacement area. The obtained results demonstrate the viability of a future alternative approach for the management of MOX-LWR.

The effects of mixing multi-component HLW glasses on spinel crystal size

The Hanford Waste Treatment and Immobilization Plant will vitrify radioactive waste into borosilicate glass. The high-level waste (HLW) glass formulations are constrained by processing and property requirements, including restrictions aimed at avoiding detrimental impacts of spinel crystallization in the melter. To understand the impact of glass chemistry on crystallization, two HLW glasses precipitating small (∼5 μm) spinel crystals were individually mixed and melted with a glass that precipitated large (∼45 μm) spinel crystals in ratios of 25, 50, and 75 wt%. The size of spinel crystals in the mixed glasses varied from 5 to 20 μm. Small crystal size was attributed to: (1) high concentrations of nuclei due to the presence of ruthenium oxide and (2) chromium oxide aiding high rates of nucleation. Results from this study indicate that the spinel crystal size can be controlled using chromium oxide and/or noble metal concentrations in the melt, even in complex mixtures like HLW glasses. Smaller crystals tend to settle more slowly than larger crystals, therefore smaller crystals would be more acceptable in the melter without a risk of failure. Allowing higher concentrations of spinel-forming waste components in the waste glass enables glass compositions with higher waste loading, thus increasing plant operational flexibility. An additional benefit to the presence of chromium oxide in the glass composition is the potential for the oxide to protect melter walls against corrosion.

Machine Learning Enabled Models to Predict Sulfur Solubility in Nuclear Waste Glasses.

The U.S. Department of Energy is considering implementing the direct feed approach for the vitrification of low-activity waste (LAW) and high-level waste (HLW) at the Hanford site in Washington state. If implemented, the nuclear waste with a higher concentration of alkali/alkaline-earth sulfates (than expected under the previously proposed vitrification scheme) will be sent to the vitrification facility. It will be difficult for the existing empirical models to predict sulfate solubility in these glasses or design glass formulations with enhanced sulfate loadings in such a scenario. Further, the existing models are unable to produce reliable predictions when applied to HLW glasses whose composition falls outside of the range encompassed by the database used to develop/calibrate the models. Accordingly, this study harnesses the power of artificial intelligence (machine learning, ML) with a goal to address the limitations of the existing models. Toward this, three ML models have been trained using a large database; comprising >1000 LAW and HLW glasses and encompassing a wide range of glass compositions and processing variables. Next, the ML model with the best prediction performance has been used to quantitatively assess and rank the influence (i.e., importance) of glasses' compositional/processing variables on the SO3 solubility in the glasses. Finally, on the premise of such understanding of influential and inconsequential variables, two closed-form analytical models─with disparate degrees of complexity (one highly parametrized and one with fewer input variables)─have been developed. Results show that both analytical models produce predictions of SO3 solubility in LAW and HLW glasses with an accuracy analogous to ML models and substantially higher than the analytical models that represent the current state-of-the-art. Overall, this study's outcomes present a roadmap─informed by data and channeled by artificial intelligence─that can be leveraged in the future to design nuclear waste glasses with unprecedented sulfur loadings.

Thermal treatment of nuclear fuel-containing Magnox sludge radioactive waste

Magnesium aluminosilicate and magnesium borosilicate glass formulations were developed and evaluated for the immobilisation of the radioactive waste known as Magnox sludge. Glass compositions were synthesised using two simplified bounding waste simulants, including corroded and metallic uranium and magnesium at waste loadings of up to 50 wt.%. The glasses immobilising corroded simulant waste formed generally homogenous and fully amorphous glasses, while those immobilising metallic wastes contained crystallites of UO 2 and U 3 O 8 . Uranium speciation within the glass was investigated by micro-focus X-ray absorption near edge spectroscopy and it was shown that the borosilicate glass compositions were characterised by a slightly lower mean uranium oxidation state than the aluminosilicate counterparts. This had an impact upon the durability, and uranium within glasses of higher mean oxidation states was dissolved more readily. All material showed dissolution rates that were comparable with simulant high level radioactive waste glasses, while the borosilicate-based formulations melted at a temperature suitable for modern vitrification technologies used in radioactive waste applications. These data highlight the potential for vitrification of hazardous radioactive Magnox sludge waste in borosilicate or aluminosilicate glass formulations, with the potential to achieve >95 % reduction in conditioned waste volume over the current baseline plan. • Magnesium aluminosilicate and magnesium borosilicate glass systems were synthesised to incorporate metallic uranium and magnesium, U 3 O 8 and Mg(OH) 2 • Uranium speciation within the glass was investigated by micro-focus X-ray absorption near edge spectroscopy • Borosilicate glasses possessed greater chemically durable than their aluminosilicate counterparts • All glass compositions demonstrated durability that is comparable to UK and French high level waste glass

Structure and crystallization behavior of phosphorus-containing nepheline (NaAlSiO4) based sodium aluminosilicate glasses

The study aims at understanding the structural role of P2O5 in meta- and per-aluminous sodium aluminosilicate glasses, followed by its correlation with their crystallization behavior. Experimental results suggest an increase in the degree of polymerization of the phosphate network and a change in the aluminum environment in the glasses from Al(OSi)4 to Al(OP)4 with an increasing P2O5/SiO2 ratio. The substitution of SiO2–by–P2O5 in the amount varying between 10 – 20 mol.% suppresses the crystallization of nepheline. This has been attributed to the unavailability of Al3+ (to form Si–O–Al linkages) due to the formation of Al–O–P linkages in the glass structure. An increase in P2O5 content (>20 mol.%) in peraluminous glasses results in the crystallization of AlPO4 as the primary phase. The results have been discussed in the light of various models developed to predict the compositional dependence of nepheline crystallization in Hanford's high-level waste glasses.

Assessing the effect of radioactive waste glass dissolution on early-stage radionuclide migration using simplified geological repository Monte Carlo transport models

The vitrification of radioactive waste within glass and subsequent disposal within a geological disposal facility (GDF) requires a comprehensive understanding of the effect of glass dissolution on GDF performance. This paper aims to analyse the effect of both high-level and intermediate-level waste (HLW and ILW) glass dissolution source terms on radionuclide release into the geosphere just above the disposal vault (the ‘crown’). Radionuclide migration was simulated in GoldSim for HLW in either granite or clay host rocks with a bentonite buffer using carbon steel or copper canisters, whereas ILW simulations considered either granite or clay host rocks, in either bentonite buffer or cement backfill, using concrete or cast-iron canisters. Glass dissolution source terms were varied by coupling GoldSim and MATLAB to modify the initial, residual, and resumption dissolution rates of the glass or by applying the analytical GRAAL model to glass dissolution. HLW glass results indicate no preference of granite over clay host rocks for a given canister type but that a copper canister is preferable to steel. ILW results suggest that a granite–bentonite–cast-iron environment yields lowest crown activities with cast-iron preferable to concrete as the canister, bentonite preferable to cement as the buffer/backfill, and granite preferable to clay as the host rock. Varying glass dissolution source terms (initial, residual, and resumption dissolution rates) had an understood effect on radionuclide migration, although changes were arguably insignificant considering peak crown activity for both HLW and ILW.

Basic study of formation of acicular crystals by chemical reactions between RuO2 and NaNO3 in a waste vitrification process

ABSTRACT Aggregates of acicular RuO2 occur in high-level waste glass. Large aggregates may settle rapidly in the glass, providing a more efficient electrical path on the melter bottom and interfering with the Joule heating of the glass. To clarify the formation mechanism of aggregates and to develop countermeasures to such formation, chemical reactions between RuO2 and NaNO3 with/without borosilicate glass were observed. Sodium ruthenates were formed from a RuO2–NaNO3 mixture heated without the glass or on a flat glass. On the flat glass, aggregates of acicular crystals of RuO2 appeared with a decreasing amount of sodium ruthenates at high temperature. Neither sodium ruthenates nor acicular crystals of RuO2 were formed from the RuO2–NaNO3 mixture well mixed with powdery glass and heated. Results of a series of experiments suggest that acicular crystals of RuO2 and their aggregates formed through the formation and decomposition of sodium ruthenate when the dissolution of sodium into the glass was slower than its reaction with RuO2. Promoting the dissolution of sodium into a glass or reducing the amount of sodium in liquid waste is expected to suppress the formation of aggregates and their sedimentation.

Effects of Al:Si and (Al + Na):Si ratios on the properties of the international simple glass, part II: Structure

AbstractHigh‐alumina containing high‐level waste (HLW) will be vitrified at the Waste Treatment Plant at the Hanford Site. The resulting glasses, high in alumina, will have distinct composition‐structure‐property (C‐S‐P) relationships compared to previously studied HLW glasses. These C‐S‐P relationships determine the processability and product durability of glasses and therefore must be understood. The main purpose of this study is to understand the detailed structural changes caused by Al:Si and (Al + Na):Si substitutions in a simplified nuclear waste model glass (ISG, international simple glass) by combining experimental structural characterizations and molecular dynamics (MD) simulations. The structures of these two series of glasses were characterized by neutron total scattering and 27Al, 23Na, 29Si, and 11B solid‐state nuclear magnetic resonance (NMR) spectroscopy. Additionally, MD simulations were used to generate atomistic structural models of the borosilicate glasses and simulation results were validated by the experimental structural data. Short‐range (eg, bond distance, coordination number, etc) and medium‐range (eg, oxygen speciation, network connectivity, polyhedral linkages) structural features of the borosilicate glasses were systematically investigated as a function of the degree of substitution. The results show that bond distance and coordination number of the cation‐oxygen pairs are relatively insensitive to Al:Si and (Al + Na):Si substitutions with the exception of the B‐O pair. Additionally, the Al:Si substitution results in an increase in tri‐bridging oxygen species, whereas (Al + Na):Si substitution creates nonbridging oxygen species. Charge compensator preferences were found for Si‐[NBO] (Na+), [3]B‐[NBO] (Na+), [4]B (mostly Ca2+), [4]Al (nearly equally split Na+ and Ca2+), and [6]Zr (mostly Ca2+). The network former‐BO‐network former linkages preferences were also tabulated; Si‐O‐Al and Al‐O‐Al were preferred at the expense of lower Si‐O‐[3]B and [3]B‐O‐[3]B linkages. These results provide insights on the structural origins of property changes such as glass‐transition temperature caused by the substitutions, providing a basis for future improvements of theoretical and computer simulation models.

Ruthenium solubility and its impact on the crystallization behavior and electrical conductivity of MoO3-containing borosilicate-based model high-level nuclear waste glasses

The present study focuses on investigating the solubility of RuO2 in a borosilicate-based model high-level waste glass and understanding its impact on the crystallization behavior and electrical conductivity of the resulting vitrified waste forms. The solubility limit of RuO2 in the investigated glass composition has been determined to be 460 ppm by weight. The higher concentration of RuO2 results in sub-micron sized Ru-rich inclusions in the glassy matrix, which eventually agglomerate to form needle-like and polyhedral RuO2 crystals. It is observed that RuO2 selectively promotes the crystallization of the rare-earth apatite phase over the powellite phase. The as-synthesized RuO2-containing glasses exhibit semiconducting behavior with a similar level of electrical conductivity below the glass transition. The percolation of non-uniformly distributed RuO2 inclusions may result in a formation of short-range low-resistive conduction pathways in the host glass matrix leading to an apparent metallic-like behavior of selected thin samples with the highest ruthenium content.

Reevaluating the efficacy of moderate annealing in nuclear waste vitrification for sustainable high-level waste management

The reliable immobilization of nuclear waste, particularly high-level waste (HLW), is a key issue for sustainable utilization of nuclear energy. Annealing process is adopted in nuclear waste vitrification to improve mechanical properties of products; however, its efficacy on chemical stability of treated HLW glass remain uncertain. This study aims to reevaluate the effects of moderate annealing on chemical durability of sodium borosilicate glass matrix incorporated with simulated HLW. Two types of leaching tests, i.e. acid leaching and the modified Product Consistency Test (PCT), were performed to examine the leaching behavior of HLW glassy products controlled by corrosion mechanisms of ion exchange and hydrolysis, respectively. The acid leaching results showed that annealed glass samples released lanthanum (La) species at a faster rate than that of quenched samples, but the effects of annealing on the release behavior of alkaline and alkaline-earth elements were limited. In contrast, the PCT results showed minor difference between quenched and annealed samples, which suggests that the small structure modification caused by moderate annealing was not comparable with the dominant hydrolysis reaction occurred in the PCT process. According to the free-volume theory, as well as evidences derived from Raman characterization and thermodynamic measurements, La was inferred to exist in the free-volume region in the HLW glass, and the annihilation of free volume during annealing could mobilize La from its stable surrounding environment, and thus reducing the chemical stability of nuclear waste glass.