Borosilicate Glass Waste Form Research Articles

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.

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.

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.

Effects of iron oxalate on rhenium incorporation into low-activity waste glass

Volatilization of technetium-99 (99Tc) is of concern during vitrification of Hanford low-activity waste (LAW) to produce a borosilicate glass waste form at the Waste Treatment and Immobilization Plant. Addition of reducing agents to LAW glass feed has been known to increase the retention of 99Tc in the final glass product. We investigated rhenium incorporation into glass by performing crucible melting tests on simulated LAW glass feeds, prepared without and with iron oxalate as a reducing agent, to understand the mechanism how the reducing agent influences Re retention in glass. Deionized water leaches were conducted on the dried feed and heat-treated samples to analyze the compositions of soluble and insoluble phases separately and to profile the partitioning of components into different phases during melting. The results in this study do not support the previous hypothesis that the low volatility of reduced Re4+ and Tc4+ is responsible for higher Re and Tc retention in glass when a reducing agent is added. We propose that early reaction of nitrate/nitrite and faster incorporation of sulfate into glass caused by iron oxalate are the two contributing factors that can explain the experimental findings in this study regarding the positive effect of iron oxalate on Re incorporation and retention.

Structural Characterization of Ternary Salt Melts for Low Activity Waste Applications

ABSTRACTReactions of alkali salts (nitrates, sulfates, carbonates, halides, borates) play a key role in the low temperature feed conversion occurring at the cold cap during processing of Hanford Low Activity Waste (LAW) glass melters. An alkali salt phase can sometimes form, and preferentially incorporate radionuclides of Cs, Cl, I, and Tc. During melting of the slurry feed, some of the feed components sequentially break down with increasing temperature to form gases (i.e., nitrates ➔ NOx, carbonates ➔ CO2, and boric acid ➔ H2O) or partially volatilize (halides). Sulfate, however, tends not to volatilize but has limited solubility in the final borosilicate glass waste form. To improve understanding of these low temperature processes and their composition dependencies, a scoping study was undertaken to synthesize salt systems that remain amorphous at room temperature, thus facilitating structural study. Melts of equimolar ratios of K2SO4-ZnSO4 (a known ionic glass-forming system) with added nitrates, halides, or carbonates, were melted and quenched. Some of the materials formed single phase glasses and some underwent crystallization upon quenching. Characterization of these quenched materials by thermal analysis, infrared absorption, and diffraction was performed. Addition of other anions to the sulfate base glass resulted in a distortion of the sulfate tetrahedron, as evidenced by infrared absorption. Carbonates strongly promoted crystallization, mostly of carbonate phases. Nitrates promoted crystallization of ZnO, and the nitrate volatilized with some incorporating into the glass. Halides tended to incorporate into the glass, but the small (F) and large (I) halogens promoted crystallization of sulfate-containing crystals, while moderate sized (Cl) halogens produced single-phase ionic glasses.

Vitrification processes of simulated cesium sorbing zeolite waste

A large amount of high-activity-level water (HALW) contaminated with mainly radioactive cesium has been generated at Fukushima Dai-ichi Nuclear Power Plant (NPP), and a decontamination system for the HALW has been operated. For decontamination of the HALW, modified chabazite-type zeolite has been used as a sorbent, and the amount of the cesium bearing zeolite waste is still increasing. One of conditioning options for the zeolite waste is vitrification into borosilicate glass waste form, which is considered to be a promising option for the back-end of nuclear fuel cycle. For the purpose of the optimum treatment and disposal, the effects of the vitrification conditions on the total performance are required to be evaluated quantitatively. In this study, vitrification tests of simulated zeolite waste, which is the same type of zeolite used in the decontamination system at Fukushima Dai-ichi NPP and is sorbing non-radioactive cesium, were examined to evaluate the homogeneity of glass products and the volume reduction ratio (glass specific volume to zeolite specific volume) with respect to the temperature of melting process. Furthermore, the volatilization loss of cesium was evaluated as a function of the additive loading, processing temperature and processing time. The results showed that the glass products are in a completely amorphous state and their volume reduction ratios were around 0.4 at any vitrification condition above 1100 °C. In addition, a certain amount of the volatilization loss occurred under the conditions of high additive loading and high temperature.

Research on chemical durability of iron phosphate glass wasteforms vitrifying SrF2 and CeF3

A part of the radioactive waste produced in the Thorium Molten Salt Reactor (TMSR) will be present as fluorides. Because of some characters of fluoride, it is undesirable for immobilizing the waste as other usual oxide waste in borosilicate glass that often produces immiscibility. However, iron phosphate glasses have been made with up to 30 wt% SrF2 and 20 wt% CeF3, and about 20 wt% SrF2 and CeF3 combined. The iron phosphate glass matrix mixed with fluoride waste could be melted at temperatures between 1050 °C and 1200 °C in only 30–60 min, its melting temperature is lower than that of borosilicate glass. Powder X-ray diffraction (XRD) showed that the glassy wasteforms that contains more than 20% wt% SrF2 and CeF3 mixture, or more than 20 wt% CeF3 would start crystallizing. The product consistency test (PCT) was used to identify the chemical durability of the glass. Glass powder (−100 to +200 mesh) after washing was soaked in 10 ml of deionized water at 90 °C for 7 days. The normalized elemental mass release of elements Sr, Ce and F were calculated at the level of 10−3 g/m2, through the concentration of element measured after PCT, that was equivalent to or better than that of borosilicate glass wasteforms. Iron phosphate glasses are concluded to be a practical alternative for vitrifying the radioactive waste of TMSR.

Preparation and Characterization of Borosilicate Glass Waste Form for Immobilization of HLW from WWER Spent Nuclear Fuel Reprocessing

ABSTRACTBorosilicate glassy materials for immobilization of HLW from Russian WWER (PWR) spent nuclear fuel reprocessing were designed, synthesized in a resistive furnace, and characterized by XRD, SEM/EDS, and FTIR spectroscopy. Chemical durability was determined by PCT-A procedure and compared to EPA glass and reference data. The glasses with 20 and 25 wt.% waste loading were found to be X-ray amorphous, homogeneous and chemically durable. Glass network formally had a relatively low degree of connectedness that was increased due to embedding of different structural groups thus improving chemical durability. Boron is present primarily in trigonal oxygen coordination. The glasses with 40-45 wt.% waste loading contained minor britholite phase concentrating rare earth elements and as expected trivalent actinides. Glassy product with up to 30 wt.% waste loading was also produced by cold crucible inductive melting at the IPCE RAS lab-scale unit equipped with 56 mm inner diameter copper cold crucible and energized from a 10 kW/5.28 MHz generator. The product was composed of vitreous phase and minor britholite with average composition K0.39Sr1.99Fe0.16Nd5.50Si7.96O26.60.

Immobilization of gadolinium in iron borophosphate glasses and iron borophosphate based glass-ceramics: Implications for the immobilization of plutonium(Ⅲ)

Immobilization of gadolinium (Gd), a nonradioactive surrogate for Pu3+, in iron borophosphate glasses/glass-ceramics (IBP glasses/glass-ceramics) has been investigated. The IBP glass containing 4 mol% Gd2O3 is homogeneously amorphous. At higher Gd2O3 concentrations, additional Gd is retained in the glasses as crystalline inclusions of monazite GdPO4 crystalline phase detected with X-ray diffraction. Moreover, Gd2O3 addition increases the Tg of the IBP glasses in glass formation range, which is consistent with the structural modification of the glasses. The structure of the Gd2O3-loaded IBP glasses/glass-ceramics is mainly based on pyrophosphate units. The chemical durability of Gd2O3-loaded IBP glasses/glass-ceramics is comparable to widely used borosilicate glass waste forms and the existence of monazite GdPO4 crystalline phase does not degrade the aqueous chemical durability of the IBP glasses/glass-ceramics. The Gd-loading results imply that the solubility should not be a limiting factor in processing nuclide Pu3+ if the formed crystalline phase(s) have high chemical durability.

Melt processed multiphase ceramic waste forms for nuclear waste immobilization

Ceramic waste forms are promising hosts for nuclear waste immobilization as they have the potential for increased durability and waste loading compared with conventional borosilicate glass waste forms. Ceramics are generally processed using hot pressing, spark plasma sintering, and conventional solid-state reaction, however such methods can be prohibitively expensive or impractical at production scales. Recently, melt processing has been investigated as an alternative to solid-state sintering methods. Given that melter technology is currently in use for High Level Waste (HLW) vitrification in several countries, the technology readiness of melt processing appears to be advantageous over sintering methods. This work reports the development of candidate multi-phase ceramic compositions processed from a melt. Cr additions, developed to promote the formation and stability of a Cs containing hollandite phase were successfully incorporated into melt processed multi-phase ceramics. Control of the reduction–oxidation (Redox) conditions suppressed undesirable Cs–Mo containing phases, and additions of Al and Fe reduced the melting temperature.

Release of uranium from candidate wasteforms

AbstractLarge volumes of depleted natural and low-enriched uranium exist in the UK waste inventory. This work reports on initial investigations of the leaching performance of candidate glass and cement encapsulation matrices containing UO3 powder as well as that of uranium oxide powders. The surface areas of UO3 powder and the monolith samples of UO3 conditioned in the glass and cement matrices were very different making leaching comparisons difficult. The results showed that for both types of monolith conditioned samples a steady increase of uranium concentration in solution with time was generally not observed. The wt.% of uranium leached from UO3 conditioned in the lead borosilicate glass wasteform was approximately five orders of magnitude less than that leached from UO3 powder. Similarly, the quantities of uranium leached from UO3 conditioned in composite cement made with ordinary Portland cement, and from magnesium phosphate cement, were approximately four and three orders of magnitude, respectively, less than that leached from UO3 powder. The performance of a mixed oxide borosilicate glass wasteform was only slightly better than that of UO3 powder. This work shows that wasteforms based on encapsulation in lead borosilicate glass and cement matrices have the greatest potential for further development.

Tellurite glass as a waste form for mixed alkali–chloride waste streams: Candidate materials selection and initial testing

Tellurite glasses have historically been shown to host large concentrations of halides. They are here considered for the first time as a waste form for immobilizing chloride wastes, such as may be generated in the proposed molten alkali salt electrochemical separations step in nuclear fuel reprocessing. Key properties of several tellurite glasses are determined to assess acceptability as a chloride waste form. TeO2 glasses with other oxides (PbO, Al2O3+B2O3, WO3, P2O5, or ZnO) were fabricated with and without 10mass% of a simulated (non-radioactive) mixed alkali, alkaline-earth, and rare earth chloride waste. Measured chemical durability is compared for the glasses, as determined by the product consistency test (PCT), a common standardized chemical durability test often used to validate borosilicate glass waste forms. The glass with the most promise as a waste form is the TeO2–PbO system, as it offers good halide retention, a low sodium release (by PCT) comparable with high-level waste silicate glass waste forms, and a high storage density.

Development of an Improved Titanate-Based Sorbent for Strontium and Actinide Separations under Strongly Alkaline Conditions

High-level nuclear waste produced from fuel reprocessing operations at the Savannah River Site (SRS) requires pretreatment to remove 134,137Cs, 90Sr, and alpha-emitting radionuclides (i.e., actinides) prior to disposal onsite as low level waste. The separation processes at SRS include the sorption of 90Sr and alpha-emitting radionuclides onto monosodium titanate (MST) and caustic side solvent extraction of 137Cs. The MST and separated 137Cs is encapsulated along with the sludge fraction of high-level waste (HLW) into a borosilicate glass waste form for eventual entombment at a federal repository. The predominant alpha-emitting radionuclides in the highly alkaline waste solutions include plutonium isotopes 238Pu, 239Pu, and 240Pu; 237Np; and uranium isotopes, 235U and 238U. This article describes recent results evaluating the performance of an improved sodium titanate material that exhibits increased removal kinetics and capacity for 90Sr and alpha-emitting radionuclides compared to the current baseline material, MST.

Liquid–Liquid Phase Separation Process in Borosilicate Liquids Enriched in Molybdenum and Phosphorus Oxides

In the context of French nuclear waste vitrification, specific borosilicate glass waste forms are being developed to immobilize the large quantities (>6 mol%) of molybdenum and phosphorus (2 mol%) oxides produced by reprocessing uranium–molybdenum spent fuel. The presence of these elements at high concentrations induces liquid–liquid phase separation during melting. In order to control the microstructure of the phases formed and to define the vitrification operating conditions, it is crucial to determine the phase separation temperature. This study is based on the melt rheological behavior during the phase separation step, and focuses on determining the phase separation temperature by measuring the transition of the liquid to non‐Newtonian behavior or a deviation from the Vogel–Fulcher–Tammann (VFT) law. The transformation mechanisms in glass containing 6 mol% molybdenum oxide are first described as a function of temperature using various microstructural characterization techniques at microscopic (scanning electron microscopy, Raman spectroscopy) and submicrometer scale (transmission electron microscopy). These results are correlated with rheological measurements to determine the phase separation temperatures of molten borosilicates containing 5–6 mol% molybdenum oxide. The immiscibility temperature of liquid–liquid phase separation can be determined by the change in the liquid rheological behavior. The transition of the liquid to non‐Newtonian behavior or a deviation from the VFT law are indicators of liquid–liquid phase separation observable at submicrometer scale.

Encapsulated gadolinium zirconate pyrochlore particles in soda borosilicate glass as novel radioactive waste form

Gadolinium zirconate pyrochlore crystals, with a potential for hosting radioactive elements, were encapsulated in a soda borosilicate glass matrix forming a composite material. The fabrication process involved mixing as a powder followed by hot pressing at a relatively low temperature of 620 °C. Composites containing pyrochlore particles of mean particle size <60 μm in concentration of up to 40 vol% exhibited a homogeneous distribution of the particles in the glass matrix. The absence of debonding of Gd zirconate particles or deflection of cracks during fracture, indicates that a strong interface between the glass matrix and the gadolinium zirconate particles was achieved, which suggests a high mechanical strength and hardness of the composites. Mechanical tests were carried out and it was confirmed that the presence of the gadolinium zirconate particles increased the Young's modulus and fracture toughness of the composite by 35% and >100%, respectively, in comparison with plain borosilicate glass. The results indicate that the new waste forms are superior in terms of structural integrity and fracture behaviour than plain borosilicate glass waste forms.

Properties and vibrational spectra of magnesium phosphate glasses for nuclear waste immobilization

The leaching behavior and structure of magnesium phosphate glasses containing 45–55 mol% MgO incorporated with simulated high level nuclear wastes (HLW) were studied. The leach rate of the waste glasses decrease with increasing of the simulated HLW content. The gross leach rate of the glass waste form containing 50 mol% MgO and 45 mass% simulated HLW is of the order of 10 −6 g/cm 2 day at 90 °C, which is small enough as compared with the corresponding release from a currently used borosilicate glass waste form. The isolated ions such as dimeric (P 2O 7) 4− and monomeric (PO 4) 3− ions increase upon as increasing the incorporating amount of the simulated HLW. The changes in properties can be attributed to the structure changes owing to the incorporation of the simulated HLW.

Effects of electron irradiation in nuclear waste glasses

This article summarizes recent studies of electron irradiation damage in sodium borosilicate, iron phosphate and aluminophosphate glass waste forms using a modern analytical electron microscope. Three different borosilicate (wt%) (17.72% B2O3–16.67% Na2O–64.61% SiO2, 17.78% B2O3–15.83% Na2O–61.39% SiO2–4.99% Fe2O3 and 17.86% B2O3–15.90% Na2O–61.63% SiO2–4.61% FeO) and iron phosphate (mol%) (45% Fe2O3–55% P2O5, 20% Fe2O3–80% P2O5, and 20% Fe2O3–20% Na2O–60% P2O5) glasses, and an aluminophosphate (mol%) (44.5% P2O5–31.5% Al2O3–20.2% Na2O–3.8% K2O) glass were studied. Results indicate that all these glasses decomposed under the 200 kV electron irradiation (at doses higher than 1.0 × 1026 e m−2). Migration of alkali elements from the irradiated centres to the peripheries under irradiation occurred in the alkali element-containing glasses, which results in the formation of alkali element-depleted and -enriched phases. Formation of bubbles was only observed in the alkali element-containing iron phosphate and aluminophosphate glasses, not in sodium borosilicate glasses when irradiated over a broad of dose rates (1.6 × 1022 e m−2 s to 8 × 1026 e m−2 s). Separation of boron-rich phase from silicon-rich phase, iron-rich/aluminium-rich phase from phosphorous-rich domains were observed in the three types of glasses, respectively. Further irradiation resulted in formation of small particles. In Fe-containing borosilicate glasses, the Fe is associated with the boron-rich phases after phase separation.

Concepts in Waste Management: Decontamination of Plutonium-Bearing Mixed Wastes with Efficient Water and Acid Recycle

Two integrated flowsheet concepts are presented for treating high-activity plutonium-contaminated mixed wastes containing materials that generate hydrogen on storage. A specialized electrocell can be used to generate cerium(IV) to dissolve high-fired plutonium oxides and simultaneously to oxidize organic species in the wastes that are susceptible to hydrogen formation. Nitric acid and water decontamination and recycle can be achieved using resin adsorption without evaporation. Transuranic contamination may be concentrated in a precipitate suitable for incorporation into borosilicate glass or other final waste forms using Fenton's chemistry or, preferably, immobilized in a silicate matrix directly from an adsorption resin and converted to a final waste form. Decontaminated TRU wastes are either converted to low-level wastes or packed to meet transuranic waste packing and shipping requirements. Preliminary results from Oak Ridge National Laboratory suggest that HEPA filter wastes can be processed to achieve decontamination factors of 104 for highly fired Pu oxide. Studies at the Argonne National Laboratory and Florida State University suggest that nitric acid and water can be recycled with decontamination factors in excess of 104 without evaporation. Alternative methods for using this technology are discussed.

High Level Liquid Waste Solidification Using a “Cold” Crucible Induction Melter

ABSTRACTAt the present time the primary problem in a closed nuclear fuel cycle is the management of high level liquid waste (HLLW) generated by the recovery of uranium and plutonium from spent nuclear fuel. Long-term storage of the HLLW, even in special storage facilities, poses a real threat of ecological accidents. This problem can be solved by incorporating the radioactive waste into solid fixed forms that minimize the potential for biosphere pollution by long-lived radionuclides and ensure ecologically acceptable safe storage, transportation, and disposal. In the present report, the advantages of a two-stage HLLW solidification process using a “cold” crucible induction melter (CCIM) are considered in comparison with a one-stage vitrification process in a ceramic melter.This paper describes the features of a process and equipment for a two-stage HLLW solidification technology using a “cold” crucible induction melter (CCIM) and identifies the advantages compared to a one-stage ceramic melter. A two-stage pilot facility and the technical characteristics of the equipment are described using a once-through evaporator and cold-crucible induction melter currently operational at the IA.Mayak. facility in Ozersk, Russia. The results of pilot-plant tests with simulated HLLW to produce a phosphate glass are described. Features of the new mineral-like waste form matrices synthesized by the CCIM method are also described. Subject to further development, the CCIM technology is planned to be used to solidify all accumulated HLLW at Mayak – first to produce borosilicate glass waste forms and then mineral-like waste forms.

99/02759 Iron-phosphate-based chemically bonded phosphate ceramics for mixed waste stabilization

The leaching behavior and structure of magnesium phosphate glasses containing 45–55 mol% MgO incorporated with simulated high level nuclear wastes (HLW) were studied. The leach rate of the waste glasses decrease with increasing of the simulated HLW content. The gross leach rate of the glass waste form containing 50 mol% MgO and 45 mass% simulated HLW is of the order of 10−6 g/cm2 day at 90 °C, which is small enough as compared with the corresponding release from a currently used borosilicate glass waste form. The isolated ions such as dimeric (P2O7)4− and monomeric (PO4)3− ions increase upon as increasing the incorporating amount of the simulated HLW. The changes in properties can be attributed to the structure changes owing to the incorporation of the simulated HLW.