Glass Waste Form Research Articles

Glass and ceramic nuclear waste forms: The scientific battle

ABSTRACT The scientific battle between two major classes of materials as to the best option for immobilization and disposal of high-level nuclear wastes (HLW) has endured since the 1970s. Both glass and ceramic materials afford different characteristics and properties that make them specifically suitable for certain types of radioactive wastes; both material types have been previously selected as preferred solutions for different waste treatment programs. During the 1980s, glass predominated following its selection for US defense wastes, and this choice is reflected in the available industrial waste treatment facilities to treat HLW. France, the United Kingdom, and the United States are each pursuing huge industrial programs of HLW treatment that isolate radioactive elements in glass forms for ultimate geologic disposal. However, several existing legacy wastes and emerging wastes from future nuclear technologies and fuel cycles will require innovative waste forms, including ceramics and glass-ceramics. This article tells the story of the history of ceramic and glass-ceramic waste forms for the immobilization of HLW and their latest developments. It also highlights some of the essential contributions of professor Rod Ewing (1946–2024) in the areas of radiation damage, natural analogues, and nuclear waste forms.

Vitrification of Cerium(Ⅲ) fluoride in iron phosphoborate glasses: A study on redox, structure, and stability

Cerium fluoride was vitrified to simulate actinide fluoride waste from high-level waste salt produced through dry reprocessing in a Molten Salt Reactor. Iron phosphoborate and phosphoborate glass were used to immobilize CeF3 and investigate the relationship between redox, structure and chemical stability of these glass wasteforms. The findings indicate that CePO4 crystals precipitated when the molar content of CeF3 exceeded 28.9 wt% in iron phosphoborate glass, as observed by X-ray Diffraction (XRD). X-ray Photoelectron Spectroscopy (XPS) was used to investigate the changes in the valence of Ce and Fe in both types of glass. Results showed that adding of CeF3 increased the fraction of Fe2+. In iron phosphoborate glass, Ce was mainly trivalent, whereas in phosphoborate glass, most Ce was oxidized to Ce4+. Raman analysis indicates that cerium-containing iron phosphoborate glass exhibits a pyrophosphate structure, while cerium-containing phosphoborate glass displays a metaphosphate structure. Results from Mössbauer Spectroscopy and X-ray Absorption Fine Structure (XAFS) indicate that iron predominately exists as Fe–O tetrahedral clusters within the glass, showcasing excellent stability. This is evident from the strong resistance to elemental leaching in the leaching test of the iron phosphoborate glass with cerium content.

Innovative approach for efficient extraction of Nd from electrorefining waste salt into nuclear waste forms

Electrorefining of the high-burnup spent nuclear fuels is critical to the closed fuel cycle in fast reactors for the sustainable development of nuclear energy. However, a remaining challenge in the electrorefining process is the separation of radioactive rare earth fission products to reduce the geological waste volume. In this work, the extraction of a major radionuclide Nd from the molten waste salt into durable waste forms was investigated through a two-step approach consisting of electrolytic reduction and oxidative extraction. Liquid Bi was found to be effective in the electrolytic extraction of Nd, leading to an extraction efficiency of up to 98.6% and a current efficiency of 93.5%. The Nd target in the resulting Nd–Bi cathode alloy was then selectively extracted using specific oxidizers in a glass melting step. High loadings of Nd2O3 in the converted zinc-in-titania glass ceramic and aluminoborosilicate glass waste forms were achieved, demonstrating the prospect of the proposed method in terms of minimizing radioactive waste generation and recycling of the electrorefining waste salt.

Copper tellurite glass for radioactive iodine immobilization

Radioactive iodine-129 can easily diffuse into the environment and possesses a half-life of 1.57 × 107 years. Therefore, durable and stable waste forms are required for the treatment I-129. In this study, CuI·Cu2O·TeO2 glasses were fabricated by incorporating CuI into a copper tellurite glass (0.3Cu2O·0.7TeO2) matrix at different concentrations ranging from 0 to 30 mol%. The X-ray diffraction analysis confirmed the amorphous phase of all prepared samples, and X-ray fluorescence analysis verified iodine volatilization during the melting process. With the addition of 30 mol% CuI, the maximum iodine loading was approximately 13.79 wt%. X-ray photoelectron spectroscopy and Raman spectroscopy were used to understand the structure of the CuI·Cu2O·TeO2 glass waste forms. As the CuI content increased in the glass matrix, there was an increase of +2-valent Cu and OBO. It was further confirmed that there were no significant changes in the structure of Te, which serves as the glass network former. Furthermore, all samples satisfied the US regulation of 2 g/m2 for the normalized release of all elements, as confirmed through the product consistency test. The proposed glass waste forms have cost advantages over previously developed silver-based glasses, making them a promising avenue for further research.

Vitrification of lead–bismuth alloy nuclear waste into a glass waste form

AbstractLead–bismuth eutectic (LBE), a promising coolant in advanced nuclear systems, can be activated by neutrons during nuclear reactor operations. The decommissioning of nuclear facilities would generate lead–bismuth (Pb–Bi) alloy‐contaminated nuclear waste. The current metallic nuclear waste treatment approach involves remelting followed by cementitious solidification. This increases the waste volume and the risk of radionuclide migration in groundwater. Therefore, this study developed a method for vitrification of Pb–Bi alloy waste. Different amounts of SiO2 were added at 750–1100°C in the air to turn the simulated LBE waste into glass waste form. The values of the normalized elemental leaching rates (Pb, Bi, Si, Te, and Ni) determined using the 28‐day static leaching test were less than .2 g m−2 d−1 and varied with SiO2 addition. Furthermore, a three‐stage evolution of the glass structure with SiO2 addition was proposed according to the structural analysis performed using Raman and X‐ray photoelectron spectroscopies. The evolution stages were as follows: (i) the stage of heavy metal transition from covalent to ionic heavy metals (7.5 wt% < SiO2 < 15 wt%), (ii) the stage of increase in bridging oxygen (15 wt% < SiO2 < 20 wt%), and (iii) the stage of domination of the Si–O network (20 wt% < SiO2 < 25 wt%). The evolution of the glass structure resulted in varying glass chemical durability. Finally, the glass‐forming region of (20–48)PbO–(35–70)Bi2O3–(7.5–25)SiO2 (wt%) and the temperature needed to melt those glasses were determined through the melting test, where radionuclides and toxic heavy metals showed undetectable volatilization during vitrification. Hence, turning LBE waste into glass waste form will be a potential approach for treating Pb–Bi alloy nuclear waste.

Effects of irradiation behavior on physicochemical properties of glassy radioactive contaminated soil containing various contents of actinides nuclear waste

The evaluation of radiation resistance of the treated radioactive contaminated soil is crucial. The irradiation behavior of simulated radioactive soil waste irradiated with 1.5 MeV Xe20+ ions at fluences of 1 × 1012-1 × 1015 ions/cm2 was studied. Before the irradiation experiment, all the samples were sintered by microwave. The results showed that microwave sintering may be used to treat radioactive contaminated soil. In addition, the irradiation experiment results show that when the Nd2O3 content was low (<20 wt.%), the irradiation has little effect on the sample. When the Nd2O3 content was higher, the Vickers hardness of the sample (25 wt.%) decreased by 7 % at a fluence of 1 × 1015 ions/cm2, which may be due to the high Nd2O3 content that destroyed the overall stability of the glass waste form. The low normalized leaching rate of the irradiated sample (LRNd, ∼10-6 g·m-2·d-1) also proved that it had good aqueous durability. Moreover, the radiation resistance of the sample was illustrated by studying the influence mechanism of 1.5 MeV Xe20+ irradiation on radioactive contaminated soil. This work can help to study the environmental pollution problems of radioactive contaminated soil containing various contents of actinide nuclear waste.

Design and Application of Materials for Sequestration and Immobilization of 99Tc.

99Technetium (99Tc) is a hazardous radionuclide that poses a serious environmental threat. The wide variation and complex chemistries of liquid nuclear waste streams containing 99Tc often create unique, site specific challenges when sequestering and immobilizing the waste in a matrix suitable for long-term storage and disposal. Therefore, an effective management plan for 99Tc containing liquid radioactive wastes (such as storage (tanks) and decommissioned wastes) will likely require a variety of suitable materials/matrixes capable of adapting to and addressing these challenges. In this review, we discuss and highlight the key developments for effective removal and immobilization of 99Tc liquid waste in inorganic waste forms. Specifically, we review the synthesis, characterization, and application of materials for the targeted removal of 99Tc from (simulated) waste solutions under various experimental conditions. These materials include (i) layered double hydroxides (LDHs), (ii) metal-organic frameworks (MOFs), (iii) ion-exchange resins (IERs) as well as cationic organic polymers (COPs), (iv) surface modified natural clay materials (SMCMs), and (v) graphene-based materials (GBMs). Second, we discuss some of the major and recent developments toward 99Tc immobilization in (i) glass, (ii) cement, and (iii) iron mineral waste forms. Finally, we present future challenges that need to be addressed for the design, synthesis, and selection of suitable matrixes for the efficient sequestration and immobilization of 99Tc from targeted wastes. The purpose of this review is to inspire research on the design and application of various suitable materials/matrixes for selective removal of 99Tc present globally in different radioactive wastes and its immobilization in stable/durable waste forms.

Influence of Radioactive Sludge Content on Vitrification of High-Level Liquid Waste

The radioactive sludges formed at the bottom of high-level liquid waste (HLW) storage tanks pose challenges when the HLWs are vitrified. This study aims to determine the influence of the sludge content (enriched in Na2O, Al2O3, NiO, Fe2O3, and BaSO4) on the structure and properties of waste glasses in order to find the optimal ratio of sludges to HLW during vitrification. In the experiments, the simulated sludge and simulated HLW were mixed at different ratios from 0:8 to 4:4, with an overall waste content of 16 wt %, in a borosilicate glass wasteform. It is found that the glass density, molar volume, sulfur retention, and glass transition temperature changed little when increasing the sludge content of the glasses, while the viscosity, chemical durability, and crystallization features of the glasses varied notably. The crystals formed in the glasses during the thermal treatment were exclusively Fe-substituted diopside (Ca, Mg, Fe)2Si2O6. An increase in the Al2O3 and NiO content of the glasses may have been responsible for the increased crystallinity at high temperatures. The leaching rate of Si, B, Na, and Cs from the glasses declined with the increasing addition of sludge to the glasses. Although all the glasses fulfilled the requirements for vitrification processing and glass-product performance, it is recommended that the sludge content of the whole waste should not exceed 25 wt %. This study guides further research on the immobilization of high-level sludges.

Selective extraction of Sm from LiCl-KCl molten salt into wasteforms via liquid bismuth

The electrorefining of spent nuclear fuels in a molten LiCl-KCl salt electrolyte recovers the valuable actinides but also releases radioactive fission products in the salt. To minimize the nuclear waste volume, dissolved fission products such as rare earths (REs) must be removed from the electrorefining salt and immobilized into proper wasteforms. Here we propose a two-step approach for treating the electrorefining waste salt by (i) the electrolytic recovery of REs from the salt using a liquid metal cathode; and subsequent (ii) oxidative extraction of REs from the liquid metal into durable wasteforms. Experimental demonstration using rare earth element samarium (Sm) and liquid metal (Bi) was successful. The Sm ions dissolved in the salt can be selectively recovered as dendrite Sm-Bi intermetallic products on the top of liquid bismuth cathode through electrolysis at appropriate applied currents or potentials. Then the Sm in the Sm-Bi alloy was efficiently extracted into both zinc-in-titania (ZIT) glass ceramic and lanthanide aluminoborosilicate (LABS) glass wasteforms through selective oxidation at Sm2O3 loadings of 18–25 wt% and 50 wt%, respectively. Recycling of the electrorefining salt with minimum nuclear waste generation is foreseeable with the use of this simple, efficient, and highly compatible process.

The effect of metals on zeolite crystallization kinetics with relevance to nuclear waste glass corrosion

Geologic disposal of vitrified radioactive material is planned in several countries, but there are remaining uncertainties related to the long-term stability of glass exposed to groundwater. Specifically, the crystallization of aluminosilicate zeolite minerals can accelerate the rate at which glass corrodes and radioactive material is released into the biosphere. In this study, we identify elemental species that may accelerate or suppress zeolite formation using a protocol to examine their effects on zeolite synthesis over a three-day duration. Our results are consistent with previous works demonstrating glass corrosion acceleration in the presence of calcium. Furthermore, we identify two elements—tin and lithium—as inhibitors of zeolite P2 (gismondine, or GIS type) nucleation and, thus, promising components for promoting the long-term durability of glass waste forms.

Technetium (Tc)/Rhenium (Re) solubility and leaching behavior from waste forms: An overview

Technetium-99 (99Tc), a radionuclide generated from nuclear industry is a great environmental concern because of its long half-life (2.13 × 105 years) and high mobility in environment. Therefore, apposite management of 99Tc is imperative to control its hazardous radiological impact on humans and other livings. So far, the major strategy implementation has been the solidification and immobilization of 99Tc radioactive waste in various matrices as waste forms and disposal in deep geological repository. However, by passing the time, 99Tc may leach out/solubilize from the waste forms under different geochemical/environmental conditions. In this minireview, we discuss some key contributions towards the solubility of 99Tc and rhenium (Re; a well-known surrogate of 99Tc) from different waste forms. Specifically, we review the solubility of 99Tc from glass, cement, ceramic, and geopolymer waste forms. The final section (conclusion) presents a short summary and future challenges need be addressed to impede the solubility of 99Tc from the designed waste forms. We believe this minireview will be beneficial to provide a significant insight on the solubility of 99Tc from aforementioned waste forms and in the design of robust matrices to minimize/prevent 99Tc migration in various environments.

High Retention Immobilization of Iodine in B-Bi-Zn Oxide Glass Using Bi2O3 as a Stabilizer under a N2 Atmosphere.

The immobilization of iodine waste suffers from serious iodine loss during heat treatment. Herein, we reported on the high iodine retention immobilization of simulated radioiodine-contaminated Bi0-SiO2 sorbent in B-Bi-Zn oxide glass using Bi2O3 as a stabilizer under a N2 atmosphere. The effects of the Bi2O3 content and sintering atmosphere on the iodine immobilization behaviors (iodine retention ratio, phase composition, microstructure, and chemical stability) were investigated. It was found that the decomposition of BiI3 was prevented by adding Bi2O3 and sintering in a N2 atmosphere. The iodine retention ratio in the obtained glass waste form was significantly enhanced with increasing Bi2O3 content and sintering in the N2 atmosphere due to the synergistic effect. The achieved record-high iodine retention (92.22 ± 2.6%) was much higher than that of conventional heat treatment route (18.01 ± 3.5%). The results demonstrated that iodine was effectively immobilized through the formation of stable BixOyI (Bi5O7I and BiOI). Furthermore, the obtained iodine waste form exhibited excellent compactness and chemical stability. Owing to its high iodine retention ratio, this route can be employed to effectively immobilize radioactive iodine.

Impacts of substitution of Fe2O3 for SiO2 on structure and properties of borosilicate glasses containing MoO3

Borosilicate glass is an effective matrix for vitrification of high-level nuclear waste. However, the solubility of molybdenum is lower. In this paper, the impacts of Fe2O3 replacing SiO2 on the phase, glass structure, and chemical durability of a simplified calcium borosilicate glass containing molybdenum were investigated. According to XRD, Raman, and FTIR spectra, replacing SiO2 with Fe2O3 makes the glass structure polymerize first and then depolymerize. Furthermore, when Fe2O3 replaces SiO2 up to 5 mol%, the samples show a fully amorphous character, and the glasses show a more disordered structure. The solubility of Mo in the simplified calcium borosilicate glass can be enhanced by extending the depolymerization regions and introducing Fe3+ to compensate for the charge of [MoO4]2−. Moreover, the substitution of Fe2O3 can improve the chemical durability of glass waste form. As a result, the borosilicate glass containing 6 mol% Fe2O3 is a potential matrix to immobilize Mo-containing HLW.

Enhanced Crevice Corrosion of Stainless Steel 316 By Degradation of Cr-Containing Hollandite Crevice Former

The disposal of high-level nuclear waste (HLW) is an important societal problem. It is also an extremely challenging corrosion research topic due to the highly complicated waste chemistry, long lasting radiation effect, and inevitable exposure to aqueous environments for hundreds of thousands of years. To safely accommodate some critical radionuclides such as volatile 129I and strongly heat generating 137Cs, numerous crystalline ceramic waste forms have been created and studied. Among the promising nuclear waste forms, hollandite was developed to isolate Cs waste and the corresponding decay product Ba due to the open tunnel structure capable of accommodating mono- or divalent cations. In this study, the synergistic corrosion interactions between a Cr-containing hollandite (Ba1.15Cr2.3Ti5.7O16) and stainless steel (SS) 316 is explored. This is relevant to the permanent disposal of HLW involving the encasement of glass or crystalline ceramic waste forms containing immobilized radionuclides in a metallic canister made from corrosion resistant alloys such as SS. The experiments performed in this study simulate the potential corrosion interactions occurring at the interface of the SS and ceramic. After corroding the SS316 and Cr-containing hollandite in proximity in 0.6 M NaCl at 90 oC for 28 days, severe crevice corrosion was identified on the surface of the SS, as evidenced by the presence of large pits and crevice damage with diameters of hundreds of microns. Similarly, localized damage was also found at matching sites on the Cr-hollandite surface, indicating interactions between the two materials. The synergistic corrosion interaction was likely due to the continuous release of Cr3+ cations from both materials, including the passive dissolution of the SS and the heterogenous degradation of the Cr-hollandite. The extra Cr3+ cations originated from the hollandite reduces the incubation time required to reach the critical crevice solution condition, thereby accelerating the breakdown of the passive film of SS and the subsequent onset of active dissolution. An adapted crevice corrosion model is applied to quantitively explain the accelerated corrosion of SS observed in this study.

Radioactive waste immobilization of Hanford sludge in magnesium potassium phosphate ceramic forms

This paper evaluates immobilization of Hanford K-Basin tank sludge in magnesium potassium phosphate ceramic forms. The waste forms were produced using two simulated non-radioactive sludge streams, each with distinct characteristics and composition. Ceramicrete with wollastonite as filler was used as the matrix for this purpose. The resulting waste forms were tested for their mechanical properties, radiation stability, and leaching resistance. In another series of tests, the samples were vitrified in borosilicate glass and glass waste forms were produced. The Product Consistency Test, the American Nuclear Society's ANS 16.1 test, and the Toxicity Characteristic Leaching Procedure, which are used to develop waste acceptance criteria in the United States for permanent storage of treated waste, were used for evaluation of the leaching resistance of all waste forms.

Low-temperature microwave solidification of soil with radioactive tailings for sustainable applications

To rapidly treat soil contaminated with radioactive tailings and high aluminum content, such soil samples with different Na2CO3 (0–30 wt%) contents were vitrified using microwave sintering. To understand the mechanisms involved, the composition, phase evolution, and microstructure of the sintered matrix were validated. When the Na2CO3 doping amount was 25 wt%, the radioactive contaminated soil with high aluminum content was successfully vitrified at a low temperature (1200 °C) within 40 min. The glass waste form had a lower sintering temperature and a higher density than the solidified body formed by the raw soil at high temperature (1500 °C). Moreover, the added simulated radionuclide Ce was uniformly distributed throughout the glass matrix. These results indicate that doping with Na2CO3 can cause rapid vitrification of soil contaminated with radioactive tailings and high aluminum content at low temperatures.

The incorporation of Li2SO4 into barium borosilicate glass for nuclear waste immobilisation

This study has systematically investigated the effect of Li2SO4 addition (2.75 −16.5 wt%) in barium borosilicate glass, to provide a pathway to optimise the glass composition and maximise sulphate incorporation. The work also provides a mechanistic understanding as to how SO42- is incorporated within the glass structure. The highest sulphate incorporation of 2.78 wt% SO3 (from 11 wt% Li2SO4 addition) was achieved without crystallisation following melting at 1200 °C. Sulphate incorporation in glass was confirmed by XRF, ICP, EDS and Raman analysis. Addition of Li2SO4 along with sodium and barium oxides improved the sulphate incorporation by mixed alkali network depolymerisation and the larger Ba cations helped to create sufficient space within the boron-silicate network to incorporate sulphate ions into the glass. An immiscible sulphate layer rich in BaSO4 and Na2SO4 formed on top of the glass at lower temperature (800–1100 °C) and subsequent diffusion of Na, Ba oxides and sulphur from this layer increased with increasing time and temperature to form a sulphate incorporated amorphous glass. Addition of Na2O played an important role to improve sulphate incorporation in the glass, as well as formation of an immiscible layer on top of the glass however, the formation of Na2SO4 lowered the sulphur incorporation rate at high temperature compared to BaSO4. Increasing the Li2SO4 content in the glass decreased the glass transition temperature. Aqueous durability testing using the standard PCT tests indicated the glass had satisfactory aqueous durability compared to benchmark environmental assessment glass. This study provides opportunities to convert Li+ and SO42- rich nuclear wastes into appropriate glass wasteforms.

Melting and draining tests on glass waste form for the immobilization of Cs, Sr, and rare-earth nuclides using a cold-crucible induction melting system

Cold-crucible induction melting (CCIM) technology has been intensively studied as an advanced vitrification process for the immobilization of highly radioactive waste. This technology uses high-frequency induction to melt a glass matrix and waste, while the outer surface of the crucible is water-cooled, resulting in the formation of a frozen glass layer (skull). In this study, for the fabrication of borosilicate glass waste form, CCIM operation test with 60 kg of glass per batch was conducted using surrogate wastes composed of Cs, Sr, and Nd as a representative of highly radioactive nuclides generated during spent nuclear fuel management. A 60 kg-scale glass waste form was successfully fabricated through melting and draining processes using a CCIM system, and its physicochemical properties were analyzed. In particular, to enhance the controllability and reliability of the draining process, an air-cooling drain control method that can control draining through air-cooling near drain holes was developed, and its validity for draining control was verified. The method can offer controllability on various draining processes, such as molten salt or molten metal draining processes, and can be applied to a process requiring high throughput draining.

Evaluating thermal stability of rare-earth containing wasteforms at extraordinary nuclear disposal conditions

The thermal stability and crystallization behaviors of La2O3 containing B2O3-CaO–Al2O3 glass waste forms were investigated to evaluate the stability of waste form during emergencies in deep geological disposal. For glasses containing 15% La2O3, LaBO3 phases were observed as major crystals from 780 °C and exhibited needlelike structures. Al, Ca, and O were homogeneously distributed throughout the entire specimen, while some portions of B and La were concentrated in some parts. By differential thermal analysis at various heating rates, the activation energy for grain growth and the crystallization rate of LaBO3 were calculated to be 12.6 kJ/mol and 199.5 kJ/mol, respectively. These values are comparable to other waste forms being developed for the same purpose.

Investigating the Leaching Rate of TiTe&lt;sub&gt;3&lt;/sub&gt;O&lt;sub&gt;8&lt;/sub&gt; Towards a Potential Ceramic Solid Waste Form

An important property of glass and ceramic solid waste forms is processability. Tellurite materials with low melting temperatures and high halite solubilities have potential as solid waste forms. Crystalline TiTe3O8 was synthesized through a solid-state reaction between stoichiometric amounts of TiO2 and TeO2 powder. The resultant TiTe3O8 crystal had a three-dimensional (3D) structure consisting of TiO6 octahedra and asymmetric TeO4 seesaw moiety groups. The melting temperature of the TiTe3O8 powder was 820℃, and the constituent TeO2 began to evaporate selectively from TiTe3O8 above around 840℃. The leaching rate, as determined using the modified American Society of Testing and Materials static leach test method, of Ti in the TiTe3O8 crystal was less than the order of 10-4 g·m-2·d-1 at 90℃ for durations of 14 d over a pH range of 2-12. The chemical durability of the TiTe3O8 crystal, even under highly acidic and alkaline conditions, was comparable to that of other well-known Ti-based solid waste forms.