Borosilicate Nuclear Waste Glass Research Articles

Formulation and testing of a high-tin borosilicate nuclear waste glass for in-can melting

Borosilicate waste glasses were successfully developed to immobilize two high-level waste raffinate streams via an in-can melter process with an Inconel 601 canister at 1050 °C. Measured viscosity and crystallinity thermal profiles were within the targeted processing constraints for the in-can melter process. Measured chemical durability of the glass by ASTM C1285–21 (Method A), ranged from normalized loss of boron, NL(B) = 1.44 – 2.65 g·m−2, and NL(B) decreased with increased waste loading, accompanied by increased SnO2 crystallinity. Measured corrosion of the in-can melter canister by a glass melt showed that Inconel 601 performed well at 1100 °C for up to 500 hr. Resistance polarization measurements versus time revealed that Inconel 601 corrosion rates in (and by) glass melts decreased from an initial rate of 63 mm·y−1 down to 10.2 mm·y−1 after 137 h with increased duration, which was attributed to formation of an oxide passivation layer (mainly Cr2O3) at the alloy-glass interface.

Predicting iodine solubility at high pressure in borosilicate nuclear waste glasses using optical basicity: an experimental study

Constraining the systematic evolution of iodine solubility in borosilicate glasses is required for the formulation of adequate glass matrices able to immobilize 129I representing a major troublesome radioisotopes produced by nuclear anthropic activities. We investigated experimentally the change in iodine solubility in a large series of borosilicate glasses synthesized under high-pressure conditions (1.5 GPa) and at 1350 °C. The nature of network modifying cation (Na, K, Rb, Ca, Ba and Sr) and the concentration of B2O3 (~ < 10 mol.% and ~ > 10 mol.%) have been tested. The XPS measurements showed that iodine speciation in glasses is mostly represented by iodide (I−) for a range of iodine solubility from 0.3 to 3.1 mol.% as determined by SEM EDS and LA-ICP-MS analyses. The iodine solubility is enhanced in glasses with lower B2O3 content and with a higher concentration in network modifying cation. Regardless of the cation nature: alkali or alkaline-earth; increasing the cation size appears to induce a decrease in iodine solubility. We used the optical basicity (ΛGlass) and iono-covalent parameter (ICPGlass) to express the large variety of investigated glass compositions; both relating to the electron donor capability of the glass. We show that iodine solubility is positively correlated with ΛGlass and accordingly negatively correlated with ICPGlass. It implies that more iodine will be dissolved in glass compositions having a stronger electron-donating capability. Evidence of the relationship between iodine solubility and oxygen network is shown by the iodine solubility positive trend with equilibrium constant of the oxygen speciation. Future work should take these parameters into consideration for modeling iodine solubility in borosilicate glasses providing additional information are collected such as the boron, aluminum and silicon speciation in glasses.

Speciation and solubility of rhenium in borosilicate waste glasses

Rhenium is frequently used as a non-radioactive surrogate for technetium. Speciation and solubility of rhenium in borosilicate nuclear waste glasses are investigated, where Raman spectroscopy shows rhenium in perrhenate (ReO4−) environments. Perrhenate solubility varied considerably at 1150 °C in ambient air in five different composition glasses, from 10,000 ppm Re for lithium alumino-borosilicate glass to near 4000 ppm for complex U.S. LAW and Japanese HLW glasses. Network modifying cation composition, especially lithium, strongly affect perrhenate solubility. Re-O symmetrical stretch mode (ν1Re-O) frequencies shift depending on the alkali cation mix in the glasses, while ν1Re-O widths are similar for all glasses studied, indicating random mixing of network modifying cations, especially Li+and Na+, around ReO4−. Perrhenate ν1Re-O areas track with Re concentrations, indicating that most, if not all, Re in the glass is in perrhenate environments. Mo added to these glasses did not significantly affect Re solubility, nor ν1Re-O behavior.

Alteration of 29Si-doped SON68 borosilicate nuclear waste glass in the presence of near field materials

The leaching of 29Si-doped SON68 French nuclear waste glass was investigated in Callovo-Oxfordian (COx) water in the presence of P235GH stainless steel and COx argillite. Experiments were carried out to understand the effect of near-field materials on the rate of glass dissolution and the migration of 29Si in the claystone core. Tests were performed by using percolation cells containing glass and stainless steel sandwiched between two clay cores in two different combinations: glass/clay at 50 and 90 °C, and glass/steel/clay/at 90 °C. COx water was injected through the percolation cells at 120 bars (solution flow rate 0.05–0.38 mL d−1). Results showed that the presence of steel did not enhance the dissolution rate of the glass despite the retention of 29Si released from the glass on the corrosion products. The glass dissolution rate up to 17 months tends towards 10−4 and 10−3 g m−2 d−1 at 50 and 90 °C, respectively. The glass alteration in the presence of clay leads to the formation of calcite and Mg-silicates at 50 °C in addition to apatite and powellite at 90 °C. The main steel corrosion products were siderite, magnetite, ferrosilicates, pyrite, pyrrhotite, troilite and mackinawite. MgSO4 and CaSO4 were also identified in the presence of steel.

Interfacial structures of spinel crystals with borosilicate nuclear waste glasses from molecular dynamics simulations

AbstractSpinel crystal formation presents a critical issue and glass formulation in nuclear waste glass processing. In this paper, the interfacial structures of the model borosilicate nuclear waste glasses, the international simple glass (ISG), with two types of spinel crystals, namely the MgAl2O4 and NiFe2O4, were studied using classical molecular dynamics simulations with effective partial charge potentials and recently developed composition‐dependent boron‐related parameters. The simulation results revealed the structural features of the borosilicate nuclear waste glasses and their interfaces with the two types of spinel crystals. It was found that there exist notable structural changes of glasses close to the interfacial region, affected by the adjacent crystal structures, terms of preferential segregation and ordering of cations, as well as ctaion coordination numbers. Specifically, the fraction of fourfold coordinated boron (B3) in glass near the interface decreases as compared to the bulk glass. In addition, the amount of fourfold coordinated Al decreases while fivefold Al increases in the glass region close to the glass‐crystal interface, which suggests indication of initial stage of crystal growth as Al adopts higher (sixfold) coordination like in the crystal as compared to majority of fourfold coordination in the glass. These interfacial structure changes obtained from MD simulations provide evidence of the influence of the precipitated crystals on the surrounding melt and glass and the initial stage of crystal growth.

Structural features of ISG borosilicate nuclear waste glasses revealed from high-energy X-ray diffraction and molecular dynamics simulations

Vitrification is a widely accepted method to immobilize nuclear waste. Detailed structural information is critical to understand the physical and chemical behaviors, including the long-term chemical durability of these glasses that are to be stored in geological storage sites. High-energy X-ray diffraction studies are used to obtain accurate structural information of the International Simple Glass (ISG), a six-component borosilicate model nuclear waste glass with a composition (mol%) of 60.2SiO2-16.0B2O3-12.6Na2O-3.8Al2O3-5.7CaO-1.7ZrO2. Classical molecular dynamics (MD) simulations with recently developed effective potentials are utilized to generate structural models of the ISG glasses to provide interpretation of the high-energy X-ray structural data and additional short and medium range structural details. The atomic structure model generated from MD simulations shows an excellent agreement with high-energy X-ray diffraction, measured for the first time for ISG, with an Rx value of 5.5%. Systematic investigations of the medium-range structural features of ISG are performed, built upon recent study of the short-range structure. In particular, the glass former cation oxygen polyhedral analysis show that [SiO4]-[SiO4] and [BO4]-[SiO4] connections are the most abundant while [ZrO6] octohedra having higher probability to connect to another [SiO4], [BO4] and [BO3]. Moreover, we have studied the ISG surface structure using MD simulations. Noticeable differences in terms of both chemical compositions and structural features between the bulk and surface are observed. There exist increased small-sized rings, under coordinated Si species, high 3-coordinated boron content (86.5%) on the surface. Very importantly, sodium ions are found to be enriched on the glass surface, consistent with recent experimental results. Both bulk and surface structural information are discussed in terms of the dissolution and corrosion mechanisms of ISG and related nuclear waste glasses.

Effects of optical dopants and laser wavelength on atom probe tomography analyses of borosilicate glasses

AbstractAtom probe tomography (APT) is a novel analytical microscopy method that provides three dimensional elemental mapping with sub‐nanometer spatial resolution and has only recently been applied to insulating glass and ceramic samples. In this paper, we have studied the influence of the optical absorption in glass samples on APT characterization by introducing different transition metal optical dopants to a model borosilicate nuclear waste glass. A systematic comparison is presented of the glass optical properties and the resulting APT data quality in terms of compositional accuracy and the mass spectra quality for two APT systems: one with a green laser (532 nm, LEAP 3000X HR) and one with a UV laser (355 nm, LEAP 4000X HR). These data were also compared to the study of a more complex borosilicate glass (SON68). The results show that the analysis data quality, particularly the compositional accuracy and sample yield, was clearly linked to optical absorption when using a green laser, while for the UV laser optical doping aided in improving data yield but did not have a significant effect on compositional accuracy. Comparisons of data between the LEAP systems suggest that the smaller laser spot size of the LEAP 4000X HR played a more critical role for optimum performance than the optical dopants themselves. The smaller spot size resulted in more accurate composition measurements due to a reduced background level independent of the material's optical properties.

Multinuclear Magnetic Resonance Investigation of Crystalline Alkali Molybdates.

A variety of crystalline alkali molybdate phases are characterized by (23)Na, (133)Cs, and (95)Mo magic-angle-spinning nuclear magnetic resonance (MAS NMR) to provide spectroscopic handles for studies of devitrification products in borosilicate nuclear waste glasses. The NMR parameters obtained from line-shape simulations are plotted as a function of various structural parameters to discern trends that may prove useful in the determination of unknown phases. These are applied to Cs3Na(MoO4)2, the most common precipitate found in cesium- and molybdenum-bearing model nuclear waste glasses, the crystal structure of which has not yet been determined, to provide structural constraints that may guide the refinement of powder X-ray diffraction data.

Vapor hydration of a simulated borosilicate nuclear waste glass in unsaturated conditions at 50 °C and 90 °C

Vapor hydration of a simulated typical French nuclear intermediate-level waste (ILW) glass in unsaturated conditions has been studied in order to simulate its behaviour under repository conditions before complete saturation of the disposal site.

Monte Carlo simulations of coupled diffusion and surface reactions during the aqueous corrosion of borosilicate glasses

Borosilicate nuclear waste glasses develop complex altered layers as a result of coupled processes such as hydrolysis of network species, condensation of Si species, and diffusion. However, diffusion has often been overlooked in Monte Carlo models of the aqueous corrosion of borosilicate glasses. Therefore, three different models for dissolved Si diffusion in the altered layer were implemented in a Monte Carlo model and evaluated for glasses in the compositional range (75−x) mol% SiO2 (12.5+x/2) mol% B2O3 and (12.5+x/2) mol% Na2O, where 0≤x≤20%, and corroded in static conditions at a surface-area-to-volume ratio of 1000m−1. The three models considered instantaneous homogenization (M1), linear concentration gradients (M2), and concentration profiles determined by solving Fick's 2nd law using a finite difference method (M3). Model M3 revealed that concentration profiles in the altered layer are not linear and show changes in shape and magnitude as corrosion progresses, unlike those assumed in model M2. Furthermore, model M3 showed that, for borosilicate glasses with a high forward dissolution rate compared to the diffusion rate, the gradual polymerization and densification of the altered layer is significantly delayed compared to models M1 and M2. Models M1 and M2 were found to be appropriate models only for glasses with high release rates such as simple borosilicate glasses with low ZrO2 content.

Current Understanding and Remaining Challenges in Modeling Long‐Term Degradation of Borosilicate Nuclear Waste Glasses

Chemical durability is not a single material property that can be uniquely measured. Instead, it is the response to a host of coupled material and environmental processes whose rates are estimated by a combination of theory, experiment and modeling. High‐level nuclear waste (HLW) glass is perhaps the most studied of any material yet there remain significant technical gaps regarding their chemical durability. The phenomena affecting the long‐term performance of HLW glasses in their disposal environment include surface reactions, transport properties to and from the reacting glass surface, and ion exchange between the solid glass and the surrounding solution and alteration products. The rates of these processes are strongly influenced and are coupled through the solution chemistry, which is in turn influenced by the reacting glass and also by reaction with the near‐field materials and precipitation of alteration products. Therefore, those processes must be understood sufficiently well to estimate or bound the performance of HLW glass in its disposal environment over geologic time scales. This article summarizes the current state of understanding of surface reactions, transport properties and ion exchange along with the near‐field materials and alteration products influences on solution chemistry and glass reaction rates. Also summarized are the remaining technical gaps along with recommended approaches to fill those technical gaps.

Rhenium Solubility in Borosilicate Nuclear Waste Glass: Implications for the Processing and Immobilization of Technetium-99

The immobilization of technetium-99 ((99)Tc) in a suitable host matrix has proven to be a challenging task for researchers in the nuclear waste community around the world. In this context, the present work reports on the solubility and retention of rhenium, a nonradioactive surrogate for (99)Tc, in a sodium borosilicate glass. Glasses containing target Re concentrations from 0 to 10,000 ppm [by mass, added as KReO(4) (Re(7+))] were synthesized in vacuum-sealed quartz ampules to minimize the loss of Re from volatilization during melting at 1000 °C. The rhenium was found as Re(7+) in all of the glasses as observed by X-ray absorption near-edge structure. The solubility of Re in borosilicate glasses was determined to be ~3000 ppm (by mass) using inductively coupled plasma optical emission spectroscopy. At higher rhenium concentrations, additional rhenium was retained in the glasses as crystalline inclusions of alkali perrhenates detected with X-ray diffraction. Since (99)Tc concentrations in a glass waste form are predicted to be <10 ppm (by mass), these Re results implied that the solubility should not be a limiting factor in processing radioactive wastes, assuming Tc as Tc(7+) and similarities between Re(7+) and Tc(7+) behavior in this glass system.

Vanadium and chromium redox behavior in borosilicate nuclear waste glasses

X-ray absorption spectroscopy (XAS) was used to characterize vanadium (V) and chromium (Cr) environments in low activity nuclear waste (LAW) glasses synthesized under a variety of redox conditions. V 2O 5 was added to the melt to improve sulfur incorporation from the waste; however, at sufficiently high concentrations, V increased melt foaming, which lowered melt processing rates. Foaming may be reduced by varying the redox conditions of the melt, while small amounts of Cr are added to reduce melter refractory corrosion. Three parent glasses were studied, where CO–CO 2 mixtures were bubbled through the corresponding melt for increasing time intervals so that a series of redox-adjusted-glasses was synthesized from each parent glass. XAS data indicated that V and Cr behaviors are significantly different in these glasses with respect to the cumulative gas bubbling times: V 4+/V total ranges from 8 to 35%, while Cr 3+/Cr total can range from 15 to 100% and even to population distributions including Cr 2+. As Na-content decreased, V, and especially, Cr became more reduced, when comparing equivalent glasses within a series. The Na-poor glass series show possible redox coupling between V and Cr, where V 4+ populations increase after initial bubbling, but as bubbling time increases, V 4+ populations drop to near the level of the parent glass, while Cr becomes more reduced to the point of having increasing Cr 2+ populations.

Helium diffusion coefficient measurements in R7T7 nuclear glass by 3He(d,α)1H nuclear reaction analysis

Abstract The immobilization of fission products and minor actinides by vitrification is the reference process for industrial management of high-level radioactive wastes generated by spent fuel reprocessing. Radiation damage and radiogenic helium accumulation must be specifically studied to evaluate the effects of minor actinide alpha decay on the glass long-term behavior under repository conditions. A specific experimental study was conducted for a comprehensive evaluation of the behavior of helium and its diffusion mechanisms in borosilicate nuclear waste glass. Helium production was simulated by external implantation with 3He ions at a concentration (≈1 at.%) 30 times higher than obtained after 10,000 years of storage. Helium diffusion coefficients as a function of temperature were extracted from the depth profiles after annealing. The 3He(d,α)1H nuclear reaction analysis (NRA) technique was successfully adopted for low-temperature in situ measurements of depth profiles. Its high depth resolution revealed helium mobility at temperatures as low as 253 K and the presence of a trapped helium fraction. The diffusion coefficients of un-trapped helium atoms follow an Arrhenius law between 253 K and 323 K. An activation energy of 0.55 ± 0.03 eV was determined, which is consistent with a process controlled by diffusion in the glass free volume.

Corrosion of Glass-Bonded Sodalite as a Function of pH and Temperature

This paper reports the results of corrosion tests with glass-bonded sodalite, a ceramic waste form (CWF) that is being developed to immobilize radioactive electrorefiner salt used to condition spent sodium-bonded nuclear fuel, and with sodalite and binder glass, the two major components of the CWF. These tests were performed with dilute pH-buffered solutions in the pH range of 5-10 at temperatures of 70 and 90°C to determine the pH dependences of the forward dissolution rates of the CWF and its components. The tests show that the pH dependences of the dissolution rates of sodalite, binder glass, and glass-bonded sodalite are similar to the pH dependence of dissolution rate of borosilicate nuclear waste glasses, with a negative pH dependence in the acidic region and a positive pH dependence in the basic region. The dissolution rates are higher at 90°C than at 70°C. Our results on the forward dissolution rates and their temperature and pH dependences will be used as components of a waste form degradation model to predict the long-term behavior of the CWF in a nuclear waste repository.

Surface layers on a borosilicate nuclear waste glass corroded in MgCl 2 solution

Surface layers on the French borosilicate nuclear waste glass, R7T7, corroded in MgCl 2 solution were studied to determine the composition, structure and stability of crystalline phases. The characteristics of the phases constituting the surface layer varied with the parameter S V × t , the glass surface area ( S) to solution volume ( V) ratio, times time ( t). At low S V × t values (< 360 days/m; ≤ 36 d) th surface layer was thin and contained mainly iron hydroxide particles and hydrotalcite crystals. At an intermediate S V × t value (2800 d/m; 5.5 y) the surface layer contained hydrotalcite-, chlorite- and saponite-type phases. At the highest S V × t value (10 7 d/m; 463 d) the major phases were saponite, powellite, barite and cerianite solid solutions. About 95% of the uranium and > 98% of the neodymium released from the glass were precipitated in the surface layer. In the 463 day experiment, 86% of the neodymium in the surface layer was in solid solution with powellite, the rest with saponite. Uranium was contained exclusively in saponite. High S V ratios, typical of disposal conditions for vitrified high-level radioactive waste, favor retention of actinides in fairly insoluble corrosion products. Observation of similar corrosion products on natural glasses as on nuclear waste glasses lend support to the hypothesis that the host phases for actinides observed in the laboratory are stable over geological periods of time.

Effects of Water Content, Temperature and Pressure on Actinide Tracer Diffusion in Melts of Haplogranitic Composition

An understanding of the rates and mechanisms of U and Th diffusion in silicate melts and glasses is necessary to fully understand such diverse phenomena as U-Th decay series disequilibria, melt-crystal partitioning, and leaching of actinides from nuclear waste glasses. Uranium diffusivities have been determined in anhydrous borosilicate nuclear waste glasses (Dunn, 1987), and numerous investigators have studied release rates of actinides during leaching of glass by hydrothermal solutions (e.g. Schreiber et al., 1985). Dissolved water is known to enhance diffusion of other high field strength elements in silicate melts (e.g. Harrison and Watson, 1983). In light of the abundance of water in felsic silicate melts, and the prevalence of water-rich compositions in ancient weathered natural glasses (Mungall and Martin 1994), we have investigated the effects of water content and temperature on U and Th diffusion rates in a synthetic haplogranite composition at pressures ranging from one atmosphere to 10 kbars.

AEM study of reacted surface layer of borosilicate nuclear waste glasses

Analysis of the reacted surface layer of borosilicate glass is important to the understanding of the long term nuclear waste glass reaction process. The objective is to assess the simulated nuclear waste glass/water reaction pathway by identifying new crystalline phases that appear on the glass surfaces during the reaction. The results can be used to validate models generated to predict long-term performance of the nuclear waste glass under a range of conditions.In this study, extensive scanning electron microscopy (SEM) with qualitative energy dispersive x-ray spectroscopy (EDS) analysis, quantitative analytical and high resolution transmission electron microscopy (AEM and HRTEM) have been performed on two 202U glasses which were reacted in saturated water vapor for 14 and 23 days, respectively. In order to study the microchemical and microstructural changes with increasing depth from the reaction surface, TEM specimens were prepared in cross-section using the ultramicrotomy slicing technique similar to that described by Bates et al. In this process, small chunks containing the reacted surface layer and a thin layer of glass were first broken off from the sample surface and each of these chunks was then embedded in resin to form a block. Finally, thin sections, approximately 90 nm thick, were microtomed from these blocks and were transfered to copper mesh grids covered by holey carbon films for observation. AEM and HRTEM analyses were accomplished using a JEOL JEM-2000FX microscope attached with a Noran/TN-5500 EDS system at the University of New Mexico.

Properties and behavior of the platinum group metals in the glass resulting from the vitrification of simulated nuclear fuel reprocessing waste

Two types of platinum group metal particles were found in borosilicate nuclear waste glasses: needle-shaped RuO2 particles and spherical PdRhxTey alloys. They form a dense sediment of high electrical conductivity and relatively high viscosity at the bottom of the ceramic melting furnace. The sludge shows a non-Newtonian flow behavior. The viscosity and conductivity of the sludge depend not only on the platinum group metal content but also on the texture and morphology of the RuO2 particles. RuO2 forms long, needle-shaped crystals which are caused by alkalimolybdate salt melts that formed in the calcine layer. The salt melts oxidize the Ru present as small RuO2 particles after calcination to higher oxidation states. Ruthenium (VI) compounds are formed, presumably, which are not stable with respect to RuO2 under the melting conditions. RuO2 precipitates and crystallizes into long, needle-like particles.

Comparison of in Situ and Laboratory Corrosion Experiments with Borosilicate Nuclear Waste Glass

ABSTRACTThe comparison of laboratory data from the corrosion of borosilicate nuclear waste glass (German SM513LW11 and French R7T7) with data from the Materials Interface Interactions Test (MIIT) and Repository Systems Simulation Test (RSST) illustrates the inherent limitations of in situ tests. Although in situ tests may confirm the short term behavior of waste forms and identify phenomena associated with the repository system, they do not provide the fundamental basis for the extrapolation of long-term behavior.