High-level Nuclear Waste Glass Research Articles

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.

Surface interfacial analysis of simulant high level nuclear waste glass dissolved in synthetic cement solutions

The corrosion mechanisms and kinetics of a Mg-rich alkali aluminoborosilicate glass simulating UK high-level waste (CaZn28) were investigated upon dissolution in synthetic cement solutions. Dissolution varied as a function the different pH and alkali/alkaline earth content of each cement solution. High resolution microscopy and spectroscopy techniques ascertained the nature of the interface between the glass and the cement solutions. TEM-EDS revealed alkali- and alkaline earth-rich silica gels, into which K, Ca and Mg were incorporated. TEM-SAED, combined with synchrotron micro-focus XRD, identified the ubiquitous precipitation of the Mg-aluminate layered double hydroxide phase, meixnerite (Mg6Al2(OH)18·4H2O), in addition to goethite (FeOOH) and crystalline silica. The C-S-H phase, tobermorite (Ca5Si6O16(OH)2·4H2O), was identified in the most Ca-rich solution only. These data give insight to the role of alkali/alkaline earth-rich solutions in the dissolution or radioactive waste glasses, of importance to the final disposition in a geological disposal facility.

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.

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.

Influence of young cement water on the corrosion of the International Simple Glass

Understanding the corrosion of nuclear waste glass is critical to predicting its safe disposal within a geological facility. The corrosion mechanisms and kinetics of the International Simple Glass, a simplified version of high-level nuclear waste glass, was shown to be significantly influenced by a high pH cement solution representative of disposal conditions. We provide the first microscopic characterisation of the porous, Zr-rich aluminoalkali-silica gel corrosion layer that was observed. Ca, Na and K from the cement solution were incorporated into the corrosion layer to charge compensate Si, Al and Zr species; the incorporation of Al was postulated to result in precipitation of an aluminosilicate-rich gel with large voids, facilitating rapid transport of species through the gel layer and significantly enhancing the corrosion rate. Precipitation of Al-containing zeolite and phyllosilicate phases was also observed, indicating that cementitious solutions may promote the detrimental ‘rate resumption’ stage of glass corrosion.

Risk-informed assessment of radionuclide release from dissolution of spent nuclear fuel and high-level waste glass

This paper aims to detail the different parameters to be considered for use in an assessment of radionuclide release. The dissolution of spent nuclear fuel and high-level nuclear waste glass was considered for risk and performance insights in a generic disposal system for more than 100,000years. The probabilistic performance assessment includes the waste form, container, geology, and hydrology. Based on the author’s previous extended work and data from the literature, this paper presents more detailed specific cases of (1) the time dependence of radionuclide release, (2) radionuclide release coupled with container failure (rate-limiting process), (3) radionuclide release through the opening area of the container and cladding, and (4) sequestration of radionuclides in the near field after container failure. These cases are better understood for risk and performance insights. The dissolved amount of waste form is not linear with time but is higher at first. The radionuclide release rate from waste form dissolution can be constrained by container failure time. The partial opening area of the container surface may decrease radionuclide release. Radionuclides sequestered by various chemical reactions in the near field of a failed container may become stable with time as the radiation level decreases with time.

Accelerated Leach Testing of Glass (ALTGLASS): II. Mineralization of hydrogels by leachate strong bases

AbstractThe durability of high level nuclear waste glasses must be predicted on geological time scales. Waste glass surfaces form hydrogels when in contact with water for varying test durations. As the glass hydrogels age, some exhibit an undesirable resumption of dissolution at long times while others exhibit near steady‐state dissolution, that is, nonresumption of dissolution. Resumption of dissolution is associated with the formation of zeolitic phases while nonresumption of dissolution is associated with the formation of clay minerals. Hydrogels with a stoichiometry close to that of imogolite, (Al2O3·Si(OH)4), with ferrihydrite (Fe2O3·0.5H2O), have been shown to be associated with waste glasses that resume dissolution. Aluminosilicate hydrogels with a stoichiometry of allophane‐hisingerite ((Al,Fe)2O3·1.3‐2Si(OH)4) have been shown to be associated with waste glasses that exhibit near steady‐state dissolution at long times. These phases are all amorphous and poorly crystalline and are also found on natural weathered basalt glasses. Interaction of these hydrogels with excess alkali and OH− (strong base) in the leachates, causes the Al2O3·nSiO2 (where n=1‐2) hydrogels to mineralize to zeolites. Excess alkali in the leachate is generated by alkali in the glass. Preliminary rate‐determining leach layer forming exchange reactions are hypothesized based on these findings.

Effect of composition on peraluminous glass properties: An application to HLW containment

Part of the Research and Development program concerning high level nuclear waste (HLW) glasses aims to assess new glass formulations able to incorporate a high waste content with enhanced properties in terms of thermal stability, chemical durability, and process ability. This study focuses on peraluminous glasses of the SiO2 – Al2O3 – B2O3 – Na2O – Li2O – CaO – La2O3 system, defined by an excess of aluminum ions Al3+ in comparison with modifier elements such as Na+, Li+ or Ca2+. To understand the effect of composition on physical properties of glasses (viscosity, density, Tg), a Design Of Experiments (DOE) approach was applied to investigate the peraluminous glass domain. The influence of each oxide was quantified to build predictive models for each property. Lanthanum and lithium oxides appear to be the most influential factors on peraluminous glass properties.

Alteration layer formation of Ca- and Zn-oxide bearing alkali borosilicate glasses for immobilisation of UK high level waste: A vapour hydration study

The UK high level nuclear waste glass modified with CaO/ZnO was investigated using the vapour phase hydration test, performed at 200 °C, with the aim of understanding the impact of the modification on the chemical composition and microstructure of the alteration layer. Experiments were undertaken on non-modified and CaO/ZnO-modified base glass, with or without 25 wt% of simulant Magnox waste calcine. The modification resulted in a dramatic reduction in gel layer thickness and also a reduction in the reaction rate, from 3.4 ± 0.3 g m−2 d−1 without CaO/ZnO modification to 0.9 ± 0.1 g m−2 d−1 with CaO/ZnO. The precipitated phase assemblage for the CaO/ZnO-modified compositions was identified as hydrated Ca- and Zn-bearing silicate phases, which were absent from the non-modified counterpart. These results are in agreement with other recent studies showing the beneficial effects of ZnO additions on glass durability.

Elucidating the Effect of Iron Speciation (Fe2+/Fe3+) on Crystallization Kinetics of Sodium Aluminosilicate Glasses

We report on the influence of Fe2O3 on the crystallization kinetics of nepheline (Na2O·Al2O3·2SiO2)‐based sodium aluminosilicate glasses. A series of glasses with varying Al2O3/Fe2O3 content were synthesized in the system 25Na2O–(25–x) Al2O3–xFe2O3–50SiO2 (x varies between 0 and 5 mol%) through melt‐quench technique. A systematic set of experiments were performed to elucidate the influence of iron speciation (Fe2+/Fe3+) on the crystallization kinetics of these glasses including: (1) obtaining the details of nonisothermal crystallization kinetics by differential scanning calorimetry, (2) determining the influence of heat treatment on the structure and iron coordination in glasses by X‐ray photoelectron spectroscopy and wet chemistry, and (3) following the crystalline phase evolution in glasses in air and inert environments by X‐ray diffraction and scanning electron microscopy. The crystallization of two polymorphs of NaAlSiO4—carnegieite (orthorhombic) and nepheline (hexagonal)—was observed in all the glasses, wherein the incorporation of iron promotes the formation of nepheline over carnegieite while shifting the crystallization mechanism from surface to volume. The influence of environment (air versus inert) and iron content on the crystallization kinetics of these glasses is contextualized from the perspective of the devitrification problem usually observed in sodium‐ and alumina‐rich high level nuclear waste glasses.

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 Zn- and Ca-oxides on the structure and chemical durability of simulant alkali borosilicate glasses for immobilisation of UK high level wastes

Compositional modification of United Kingdom high level nuclear waste (HLW) glasses was investigated with the aim of understanding the impact of adopting a ZnO/CaO modified base glass on the vitrified product phase assemblage, glass structure, processing characteristics and dissolution kinetics. Crystalline spinel phases were identified in the vitrified products derived from the Na2O/Li2O and the ZnO/CaO modified base glass compositions; the volume fraction of the spinel crystallites increased with increasing waste loading from 15 to 20wt%. The spinel composition was influenced by the base glass components; in the vitrified product obtained with the ZnO/CaO modified base glass, the spinel phase contained a greater proportion of Zn, with a nominal composition of (Zn0.60Ni0.20Mg0.20)(Cr1.37Fe0.63)O4. The addition of ZnO and CaO to the base glass was also found to significantly alter the glass structure, with changes identified in both borate and silicate glass networks using Raman spectroscopy. In particular, these glasses were characterised by a significantly higher Q3 species, which we attribute to Si–O–Zn linkages; addition of ZnO and CaO to the glass composition therefore enhanced glass network polymerisation. The increase in network polymerisation, and the presence of spinel crystallites, were found to increase the glass viscosity of the ZnO/CaO modified base glass; however, the viscosities were within the accepted range for nuclear waste glass processing. The ZnO/CaO modified glass compositions were observed to be significantly more durable than the Na2O/Li2O base glass up to 28days, due to a combination of the enhanced network polymerisation and the formation of Ca/Si containing alteration layers.

Effect of Callovo-Oxfordian clay rock on the dissolution rate of the SON68 simulated nuclear waste glass

Long-term storage of high-level nuclear waste glass in France is expected to occur in an engineered barrier system (EBS) located in a subsurface Callovo-Oxfordian (COx) clay rock formation in the Paris Basin in northeastern France. Understanding the behavior of glass dissolution in the complex system is critical to be able to reliably model the performance of the glass in this complex environment. To simulate this multi-barrier repository scenario in the laboratory, several tests have been performed to measure glass dissolution rates of the simulated high-level nuclear waste glass, SON68, in the presence of COx claystone at 90°C. Experiments utilized a High-Performance Liquid Chromatography (HPLC) pump to pass simulated Bure site COx pore water through a reaction cell containing SON68 placed between two COx claystone cores for durations up to 200days. Silicon concentrations at the outlet were similar in all experiments, even the blank experiment with only the COx claystone (∼4mg/L at 25°C and ∼15mg/L at 90°C). The steady-state pH of the effluent, measured at room temperature, was roughly 7.1 for the blank and 7.3–7.6 for the glass-containing experiments demonstrating the pH buffering capacity of the COx claystone. Dissolution rates for SON68 in the presence of the claystone were elevated compared to those obtained from flow-through experiments conducted with SON68 without claystone in silica-saturated solutions at the same temperature and similar pH values. Additionally, through surface examination of the monoliths, the side of the monolith in direct contact with the claystone was seen to have a corrosion thickness 2.5× greater than the side in contact with the bulk glass powder. Results from one experiment containing 32Si-doped SON68 also suggest that the movement of Si through the claystone is controlled by a chemically coupled transport with a Si retention factor, Kd, of 900mL/g.

Hierarchical Modeling of HLW Glass-Gel-Solution Systems for Stage 3 Glass Degradation

ABSTRACTThe necessity to a priori predict the durability of high level nuclear waste (HLW) glasses on extended time scales has led to a variety of modeling approaches based primarily on solution (leachate) concentrations. The glass composition and structure control the leachate and the gel compositions which in turn control what reaction products form: the leached layer is a hydrogel and reacts with the solution (leachate) to form secondary phases some of which cause accelerated glass dissolution which is undesirable. Glasses with molar excess alkali that is not bound to glass forming (Al,Fe,B)O4 structural groups in the glass resume accelerated leaching. The hydrogels of the glasses that resume accelerated leaching at long times contain excess alkali and the leachates contain excess strong base, [SB]ex. The [SB]ex further accelerates aluminosilicate gel aging into analcime with time. Glasses with no excess molar structural alkali do not resume accelerated leaching: the glass generates weak acids, [WA], in the leachate favoring hydrogel aging into clays. These data indicate that the gel layer transforms to secondary phases in situ in response to interactions with the chemistry of a continuously evolving leachate.

Nepheline Crystallization in High-Alumina High-Level Waste Glass

ABSTRACTThe understanding of the crystallization of aluminosilicate phases in nuclear waste glasses is a major challenge for nuclear waste vitrification. Robust studies on the compositional dependence of nepheline formation have focused on large compositional spaces with hundreds of glass compositions. However, there are clear benefits to obtaining complete descriptions of the conditions under which crystallization occurs for specific glasses, adding to the understanding of nucleation and growth kinetics and interfacial conditions. The focus of this work was the investigation of the microstructure and composition of one simulant high-level nuclear waste glass crystallized under isothermal and continuous cooling schedules. It was observed that conditions of low undercooling, nepheline was the most abundant aluminosilicate phase. Further undercooling led to the formation of additional phases such as calcium phosphate. Nepheline composition was independent of thermal history.

Irradiation Impact on the Leaching Behavior of HLW Glasses

Fission products and minor actinides arising from spent fuel reprocessing are immobilized in a borosilicate matrix known as high level nuclear waste glass (i.e. HLW glass). Glass packages are intended eventually for disposal in a geological repository. The long-term behavior of HLW glass subjected to radiation by long-life radionuclides must thus be investigated with respect to geological disposal. The present article focuses on HLW glass alteration under irradiation conditions by providing a general description of the methodology and illustrating it through examples. The scientific approach adopted to study irradiation potential effects combines experiments performed on radionuclide-doped borosilicate glasses and non radioactive glasses externally irradiated, focusing on all the different leaching stages (initial, drop and residual steps). Different parameters, such as the cumulative dose, the dose rate, the electronic and the ballistic impacts, are taken into account. As under disposal conditions the glass radioactivity will be quickly governed by alpha decay (after 300 years), the results presented here to illustrate the methodology concern the behavior of alpha-doped glasses. Concerning the initial alteration rate, both the alpha activity and the alpha cumulative dose have been studied and no significant effects have been observed. The structural variations observed on the irradiated solid are thus not important enough to induce any changes in the initial chemical reactivity between glass and water. Concerning the drop and residual alteration steps, the behavior of a 239Pu-doped glass having an alpha dose rate corresponding to a HLW glass after around 1500 years has shown no effect on the kinetics, nor on the altered layer morphology. Moreover, the plutonium appears to be highly retained in the alteration layer (∼ 95%).

Structural Behavior of Tc and I Ions in Nuclear Waste Glass

Technetium-99 (Tc) and iodine-129 (I) are two long-lived fission products of high volatility, which makes their study in glass structure challenging. Both technetium and iodine have broad ranging multivalent chemistry and complex reactivity dependent on redox conditions; technetium and iodine redox may vary from Tc0 to Tc7+ and from I- to I7+. Relatively few studies have been done on their speciation in glass, in part because of their low retention at the temperatures required for glass melting. To better understand the redox and structural behavior of Tc and I in various nuclear waste glasses, a series of technetium- and iodine-containing borosilicate glasses of varied chemistry were prepared at scales ranging from a few grams to hundreds of kilograms. Technetium was included in both high-level and low-level nuclear waste glass formulations under a variety of redox conditions at concentrations ranging from 0.003 wt% to 0.06 wt%. Non-radioactive iodine glass samples were prepared in crucible melts using excess amounts of sodium or potassium iodide salts or ammonium iodate that resulted in concentrations ranging from 0.04 to 1.27 wt% iodine. These samples were also compared to glasses prepared in pilot-scale experiments in which the overall retentions reached 48% and 34% for technetium and iodine, respectively. Tc and I speciation in the resulting glasses were determined by X-ray absorption spectroscopy (XAS). While technetium was found as Tc0, Tc4+, and Tc7+, only I- was identified in these glasses. Previous studies of Tc local environment information inferred from K-edge XAS and Raman spectroscopy identified pertechnetate tetrahedra surrounded by network-modifying cations in oxidized glasses and octahedral TcO6 units in glasses prepared under reducing conditions. Conversely, iodine K-edge XAS of all glasses studied indicate iodide environments with lithium or sodium nearest-neighbors resembling disordered versions of octahedral sites in crystalline lithium or sodium iodide.

Phase Separation and Crystallisation in UK HLW Vitrified Products

Given the complex composition of the UK High Level nuclear Waste (HLW) streams and the single base glass frit composition used to vitrify these streams, some phase separation and/or crystallisation in the vitrified product is almost inevitable. Depending upon the specific composition(s), the presence of phase separated/crystalline material can be deleterious, inconsequential or even advantageous to the quality of the product. It can also determine (or be determined by) the operational plant parameters for HLW glass manufacture.This paper focuses on HLW product glasses manufactured on the Vitrification Test Rig, a full scale non-active vitrification plant based on the Atelier de Vitrification de La Hague process. The influence of processing conditions, waste stream composition and overall waste loading on the phase separation/crystallisation in historical and current vitrified products are discussed. The implications of future changes to waste stream constituents along with a potential move to an alternative base glass formula are also explored.

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.

The Effect of γ-radiation on Mechanical Properties of Model UK Nuclear Waste Glasses

ABSTRACTThe 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.