High-level Radioactive Waste Glass Research Articles

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.

Recovery of Se, Zr, Pd, and Cs from simulated high‐level radioactive waste glass through phase separation

AbstractIn this study, elemental recovery was performed using phase separation from simulated high‐level radioactive waste (HLW) glass. To cause phase separation, SiO2 and B2O3 were added to the simulated HLW glass and adjusted the ratio of SiO2: B2O3: other oxides to 40:50:10. The phase separated glass was immersed in aqueous solutions of 0–3 mol/L of HNO3, H2SO4, and a 1:1 mixture of HCl–HNO3 at 363 K for 20 h, and the dissolution behavior of 17 elements was examined. The relationship between the dissolved mass fraction of each element and the acid concentration in the immersion liquid could be approximated by the modified sigmoid function. The recovery of stable nuclei Se, Zr, Pd, and Cs instead of long‐lived radioactive nuclei was tested using a four‐stage leaching process in which the sample was immersed sequentially in four aqueous solutions at 363 K of distilled water, HNO3, H2SO4, and a 1:1 mixture of HCl–HNO3 for 20 h. It was confirmed that Se, Zr, Pd, and Cs could be recovered selectively. Furthermore, the recovery result could be predicted based on the individual dissolution results described above.

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.

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

Thermodynamic assessment of the Na2O-Al2O3-SiO2-B2O3 pseudo-binary and -ternary systems

Thermodynamic assessments of the pseudo-binary and -ternary systems formed by B2O3 with Na2O, Al2O3, and SiO2 were conducted according to the CALPHAD methodology. The compound energy formalism and two-sublattice partially ionic liquid models were used to thermodynamically represent the solid solutions malinkoite and mullite and the liquid phase of each system, respectively. A comprehensive literature review of available phase equilibria and thermodynamic experimental data as well as a detailed discussion of the modeling approaches implemented to optimize each system is provided. Assessment results are then presented and discussed, and the future path forward is outlined. The addition of B2O3 to the Na2O-Al2O3-SiO2 system contributes to the development of a thermodynamic database that will ultimately predict the equilibrium behavior of nepheline formation in high-level radioactive waste glass.

Effect of clayey groundwater on the dissolution rate of SON68 simulated nuclear waste glass at 70 °C

To predict the long-term behavior of high-level radioactive waste glass, it is necessary to study aqueous dissolution of the glass matrix under geological repository conditions. The present article focuses on SON68 (an inactive surrogate of the R7T7 glass) glass alteration in synthetic clayey groundwater at 70 °C. Experiments in deionized water as reference were also performed in the same conditions. Results are in agreement with those of previous studies showing that magnesium present in the solution is responsible for higher glass alteration. This effect is transient and pH-dependent: Once all the magnesium is consumed, the glass alteration rate diminishes. Precipitation of magnesium silicate of the smectite group seems to be the main factor for the increased glass alteration. A pH threshold of 7.5–7.8 was found, above which precipitation of these magnesium silicates at 70 °C is possible. TEM observations reveal that magnesium silicates grow at the expense of the passivating gel, which partly dissolves, forming large pores which increase mass transfer between the reacting glass surface and the bulk solution.

Real‐time monitoring of crystal accumulation in the high‐level waste glass melters using an electrical conductivity method

AbstractDuring the vitrification of high‐level radioactive waste (HLW) in HLW melters in the Waste Treatment and Immobilization Plant located in Washington State, spinel crystals may precipitate from glass and accumulate in the melter riser, preventing the discharge of molten glass into canisters. Therefore, an effort is being made to develop an electrical conductivity method to monitor crystal buildup in the melter riser. A vertically configured electrical conductivity (EC) probe with an alumina shaft and Pt‐10%Rh‐electrodes was designed and tested in standard conductivity solutions and glass melts both with and without spinel crystals. The EC probe measured conductivity in conductivity solutions within 10% of their certified values and showed a linear relationship with increased spinel layer thicknesses. Testing in silicate glass containing spinel crystals allowed for the determination of spinel conductivity as a function of temperature. The conductivities of spinel crystals of 20‐24 S/m at 800°C were in excellent agreement with the conductivity of trevorite (NiFe2O4) crystals at 800°C reported in the literature. The conductivities of spinel crystals and measured changes in the conductivity across the accumulated layer allowed for a successful measurement of spinel crystal accumulation in simulated HLW glass.

Platinum group metal particles aggregation in nuclear glass melts under the effect of temperature

The viscosity of simulated high level radioactive waste glasses containing platinum group metal particles is studied over a wide range of shear stress, as a function of the particles content and the temperature, thanks to a stress imposed rheometer, coupled to a high-temperature furnace. The system shows a very shear thinning behavior. At high shear rate, the system behaves as a suspension of small clusters and individual particles and is entirely controlled by the viscosity of the glass matrix as classical suspensions. At low shear rate, above a certain fraction in platinum group metal particles, the apparition of macroscopic aggregates made up of chains of RuO2 particles separated by thin layers of glass matrix strongly influences the viscosity of the nuclear glass and leads, in particular, to the apparition of yield stress and thixotropic effects. The maximum size of these clusters as well as their effective volume fraction have been estimated by a balance between Van der Waals attractive forces and hydrodynamic forces due to shear flow. We showed experimentally and theoretically that this aggregation phenomenon is favored by an increase of the temperature, owing to the viscosity decrease of the glass matrix, leading to an unusual increase of the suspension viscosity.

REDOX state analysis of platinoid elements in simulated high-level radioactive waste glass by synchrotron radiation based EXAFS

Extended X-ray Absorption Fine Structure (EXAFS) analyses were performed to evaluate REDOX (REDuction and OXidation) state of platinoid elements in simulated high-level nuclear waste glass samples prepared under different conditions of temperature and atmosphere. At first, EXAFS functions were compared with those of standard materials such as RuO2. Then structural parameters were obtained from a curve fitting analysis. In addition, a fitting analysis used a linear combination of the two standard EXAFS functions of a given elements metal and oxide was applied to determine ratio of metal/oxide in the simulated glass. The redox state of Ru was successfully evaluated from the linear combination fitting results of EXAFS functions. The ratio of metal increased at more reducing atmosphere and at higher temperatures. Chemical form of rhodium oxide in the simulated glass samples was RhO2 unlike expected Rh2O3. It can be estimated rhodium behaves according with ruthenium when the chemical form is oxide.

Effect of natural and synthetic iron corrosion products on silicate glass alteration processes

Glass long term alteration in the context of high-level radioactive waste (HLW) storage is influenced by near-field materials and environmental context. As previous studies have shown, the extent of glass alteration is strongly related to the presence of iron in the system, mainly provided by the steel overpack around surrounding the HLW glass package. A key to understanding what will happen to the glass-borne elements in the geological disposal lies in the relationship between the iron-bearing phases and the glass alteration products formed.In this study, we focus on the influence of the formation conditions (synthetized or in-situ) and the age of different iron corrosion products on SON68 glass alteration. Corrosion products obtained from archaeological iron artifacts are considered here to be true analogues of the corrosion products in a waste disposal system due to the similarities in formation conditions and physical properties. These representative corrosion products (RCP) are used in the experiment along with synthetized iron anoxic corrosion products and pristine metallic iron. The model-cracks of SON68 glass were altered in cell reactors, with one of the different iron-sources inserted in the crack each time. The study was successful in reproducing most of the processes observed in the long term archaeological system.Between the different systems, alteration variations were noted both in nature and intensity, confirming the influence of the iron-source on glass alteration. Results seem to point to a lesser effect of long term iron corrosion products (RCP) on the glass alteration than that of the more recent products (SCP), both in terms of general glass alteration and of iron transport.

Partitioning of rhodium and ruthenium between Pd–Rh–Ru and (Ru,Rh)O2 solid solutions in high-level radioactive waste glass

The partitioning of rhodium and ruthenium between Pd–Rh–Ru alloy with a face-centered cubic (FCC) structure and (Ru,Rh)O2 solid solution has been investigated between 1273 and 1573 K at atmospheric oxygen fugacity. The rhodium and ruthenium contents in FCC increase, while the RhO2 content in (Ru,Rh)O2 decreases with increasing temperature due to progressive reduction of the system. Based on the experimental results and previously reported thermodynamic data, the thermodynamic mixing properties of FCC phase and (Ru,Rh)O2 have been calibrated in an internally consistent manner. Phase equilibrium of platinum grope metals in an HLW glass was calculated by using the obtained thermodynamic parameters.

Viscoplasticity of simulated high-level radioactive waste glass containing platinum group metal particles

The shear rate dependency of the viscosity of three simulated high-level radioactive waste glasses containing 0, 1.2 and 4.5wt% platinum group metals (PGMs) was examined at a temperature range of 1173–1473K by a rotating viscometer. Shear stress when the shear rate equals zero, i.e. yield stress, was also measured by capillary method. The viscosity of the glass containing no PGM was shear rate-independent Newtonian fluid. On the other hand, the apparent viscosity of the glasses containing PGMs increased with decreasing shear rate, and nonzero amount of yield stresses were detected from both glasses. The viscosity and yield stress of the glass containing 4.5wt% PGMs was roughly one to two orders of magnitude greater than the glass containing 1.2wt% PGMs. These viscoplastic properties were numerically expressed by Casson equation.

Oxidation state and local environment of selenium in alkali borosilicate glasses for radioactive waste immobilisation

The oxidation state and local environment of selenium in alkali borosilicate glasses for high level radioactive waste (HLW) immobilisation were studied by Se K-edge XANES and EXAFS spectroscopy. An inactive surrogate of the UK's “MW” HLW-loaded glass and two waste-free glasses were investigated. Results confirm that the predominant Se oxidation state in all air-melted glasses is Se 4+. Low levels (< 10% each of Se 0 and Se 6+) were also detected in the simulated HLW-loaded glass. The presence of Se 6+ is consistent with moderately oxidising melting conditions arising from decomposition of nitrates in the waste. Results also suggest small but measurable Fe–Se redox interactions. Imposed atmospheres of air, N 2, or H 2 during melting resulted in increasingly reduced average Se valence as expected. Linear combination XANES fitting to quantify the relative abundances of these species was restricted by the limited availability of appropriate XANES standards for reduced Se species. EXAFS of all air-melted glasses provided robust fits indicating Se–O bond lengths of 1.71 ± 0.1 Å and CN = 3 ± 0.3, consistent with Se 4+ being present in SeO 3 2− selenite groups. Therefore, 79Se in UK alkali borosilicate HLW glasses is expected to occur predominantly as Se 4+ in SeO 3 2− selenite groups.

Mechanical properties of nuclear waste glasses

The mechanical properties of nuclear waste glasses are important as they will determine the degree of cracking that may occur either on cooling or following a handling accident. Recent interest in the vitrification of intermediate level radioactive waste (ILW) as well as high level radioactive waste (HLW) has led to the development of new waste glass compositions that have not previously been characterised. Therefore the mechanical properties, including Young’s modulus, Poisson’s ratio, hardness, indentation fracture toughness and brittleness of a series of glasses designed to safely incorporate wet ILW have been investigated. The results are presented and compared with the equivalent properties of an inactive simulant of the current UK HLW glass and other nuclear waste glasses from the literature. The higher density glasses tend to have slightly lower hardness and indentation fracture toughness values and slightly higher brittleness values, however, it is shown that the variations in mechanical properties between these different glasses are limited, are well within the range of published values for nuclear waste glasses, and that the surveyed data for all radioactive waste glasses fall within relatively narrow range.

Bibliography of the Published Papers, Reports, Articles and Abstracts on Water-Rock, Mineral, Glass (Natural Glass, Engineering Glass, High-Level Radioactive Waste Glass), and Ceramics Interaction

Bibliography of the Published Papers, Reports, Articles and Abstracts on Water-Rock, Mineral, Glass (Natural Glass, Engineering Glass, High-Level Radioactive Waste Glass), and Ceramics Interaction

Corrosion Behavior of Simulated LLW Glass in Deionized Water

AbstractStatic leach tests were conducted for simulated low-level radioactive waste (LLW) glass in deionized water at 90 °C for up to one year to investigate the dissolution mechanism of LLW glass. Widely studied leaching behavior of high-level radioactive waste (HLW) glass is referred in discussing the dissolution mechanism. LLW glass is characterized by higher sodium (Na) and aluminum (Al) contents than HLW glass, about twice as high as R7T7, with its SiO2 content close to HLW glass. Powdered simulated LLW glass of three different chemical compositions was tested with the glass-surface-to-water-volume ratio of 2,000 m−1. The release rates of boron (B), widely used as an indicator of dissolution for HLW glass, decreased with time during leaching, as commonly observed in similar tests for HLW glass. The pH of the leachate was stable around 11.3 - 11.6, which is higher than those in similar tests for HLW glass by one pH unit or more. The concentrations of Al in the leachates were higher compared to data for HLW glass by two orders of magnitude. The high concentration seems to be caused by higher pH. In the leachate condition of the present tests, a zeolitic mineral (analcime) is thermodynamically more stable than amorphous silica (SiO2(am)) which is known to control the concentration of dissolved silica (Si) with respect to HLW glass. The present results imply that dissolution of the LLW glass is accompanied with formation of analcime under virtually closed systems such as geological repository where the groundwater flow rate is quite low.

Recovery of Metals from Simulated High-Level Radioactive Waste Glass through Phase Separation

The recovery of metals from simulated high-level radioactive waste (HLW) in a glass form using the phase separation of borosilicate glass was studied in order to satisfy possible future demands such as the adoption of some new treatments for the waste in the glass or use of the vitrified elements as resources. The simulated HLW glass was separated into SiO2-rich and B2O3-rich phases at 973 and 873K when the ternary mass ratio of SiO2:B2O3:Na2O was adjusted to 68:27:5 by the addition of SiO2 and B2O3 to the simulated HLW glass. Annealing at the lower temperature of 878K promoted the distribution of the elements in the B2O3-rich phase. Approximately 90% of the Ni, Zn, Fe, Nd, Te, Zr, and Mo were distributed in the B2O3-rich phase, and these elements, except Zr, were almost completely leached into 1 mol dm−3 nitric acid at 363 K. The leaching of Zr was also achieved using 1.5 mol dm−3 sulfuric acid after the nitric acid leaching. The leached fractions of the glass-network components, such as B and Al, were lower than the other elements at approximately 70–80% due to the existence in the SiO2-rich phase. An increase in the concentration of CaO in the glass in the range of 2–5 wt% inhibited the distribution of elements into the B2O3-rich phase. The increase in the concentration of CaO also changed the structure of the B2O3-rich phase from the continuous unit shape to the discontinuous spherical shape. Consequently, the leaching fraction of every element dramatically decreased. The structure of the B2O3-rich phase returned to continuous when the concentration of CaO was more than 10 wt%. From this glass, the leaching of 69% Zr was possible using only nitric acid without any sulfuric acid.

Studies on the physicochemical properties of borosilicate glass developed for radioactive waste

To simultaneously vitrify Ion Exchange Resin, Zeolite, and Dry Active Waste generated from Korean Nuclear Power Plant, a borosilicate glass system was formulated. Viscosity and electrical conductivity of the glass were measured and within the desired ranges at the processing temperature. Those activation energies were evaluated as 152 and 70.46 kJ/mol within a temperature range of 1223 to 1623 K, respectively. Time-Temperature-Transformation study was performed using data from heat treatment. The hematite crystal was found within a temperature range of 823 to 1123 K. Mossbauer spectroscopy showed about 42% Fe 2+ state existed in the glass produced from operation of the pilot-scale plant. Product Consistency Test performed from 7 to 120 d in the glass showed the leach rates of B, Na, Li and Si were much less than those of the benchmark glass. International Organization for Standardization test was performed at 363 K for 1022 d and shown that Cumulative Fraction Leached values of Na, Li, and Si were saturated below the fraction of 0.4 except that the leaching of B increases continuously. About 50 μm thickness layer was observed to be as a protective layer against continuous corrosion. According to Vapor Hydration Test, the corrosion rate of the glass was 2 g/m 2 /d and met the specification (50 g/m 2 /d). Based on Soxhlet leaching accomplished at 371 K for 30 d, the weight loss of the glass was determined as 106.8 g/m 2 which was lower than those of other HLW (High-Level radioactive Waste) glasses.

Recovery of Metals from Simulated High-Level Radioactive Waste Glass through Phase Separation

The recovery of metals from simulated high-level radioactive waste (HLW) in a glass form using the phase separation of borosilicate glass was studied in order to satisfy possible future demands such as the adoption of some new treatments for the waste in the glass or use of the vitrified elements as resources. The simulated HLW glass was separated into SiO2-rich and B2O3-rich phases at 973 and 873K when the ternary mass ratio of SiO2:B2O3:Na2O was adjusted to 68:27:5 by the addition of SiO2 and B2O3 to the simulated HLW glass. Annealing at the lower temperature of 878K promoted the distribution of the elements in the B2O3-rich phase. Approximately 90% of the Ni, Zn, Fe, Nd, Te, Zr, and Mo were distributed in the B2O3-rich phase, and these elements, except Zr, were almost completely leached into 1 mol dm−3 nitric acid at 363 K. The leaching of Zr was also achieved using 1.5 mol dm−3 sulfuric acid after the nitric acid leaching. The leached fractions of the glass-network components, such as B and Al, were lower than the other elements at approximately 70–80% due to the existence in the SiO2-rich phase. An increase in the concentration of CaO in the glass in the range of 2–5 wt% inhibited the distribution of elements into the B2O3-rich phase. The increase in the concentration of CaO also changed the structure of the B2O3-rich phase from the continuous unit shape to the discontinuous spherical shape. Consequently, the leaching fraction of every element dramatically decreased. The structure of the B2O3-rich phase returned to continuous when the concentration of CaO was more than 10 wt%. From this glass, the leaching of 69% Zr was possible using only nitric acid without any sulfuric acid.

Aqueous Corrosion Behavior of Glass Phase of Simulated Low Level Waste Form Produced by In-can Type Induction-Heated Melting

Static corrosion tests were performed for the glass phase of a simulated waste form of non-combustible radioactive low-level waste to study a basic aqueous corrosion behavior. The waste form, which was fabricated from simulated waste sample by use of in-can type induction-heated melting, consists of two separated phases; a glass phase and a metal phase. Tests were performed for the glass phase from two types of the waste form with different chemical composition at 35°C and S/V ratio of 2,600 m−1. The glass phase with both types showed an incongruent dissolution similar to conventional high-level radioactive waste (HLW) glasses, i.e., the normalized elemental mass loss (NLi) for soluble elements such as B and Na continued to increase after the saturation of insoluble elements such as Si, A1 and Ca. The NLi for B increased in proportion to the square root of time except for early stage, which suggests that the rate of the long-term dissolution or alteration may be controlled by a diffusion process. Potential secondary phases forming as the results of incongruent dissolution were estimated to be kaolinite and calcite by comparison of the measured solution data with the thermodynamically calculated phase stability relationships. These results suggest that the glass phase has a potential chemical durability not so different from conventional HLW glasses.