Glass Waste Form Research Articles

Fabrication and physical properties of lanthanide oxide glass wasteform for the immobilization of lanthanide oxide wastes generated from pyrochemical process

The pyrochemical process, which uses a dry method to recycle used nuclear fuel generates waste LiCl–KCl salt containing radioactive lanthanide elements. To reuse LiCl–KCl salt, the lanthanide elements are separated through a precipitation method promoted by oxygen sparging and the separated fission product of lanthanide oxide should be fabricated into durable wasteforms sustainable for several 1,000 years to store in a final geological repository. Herein, we report the fabrication of a borosilicate glass based wasteform with a glass matrix of SiO2–Al2O3–B2O3 having a high waste loading of 50 wt% lanthanide oxide. Th physical properties of four kinds of wasteforms having a different lanthanide oxide waste composition were evaluated. To investigate the long-term physical stability of each sample having 50 wt% lanthanide oxide waste loading, time–temperature–transformation (TTT) test was conducted at 500 and 700 °C for 60 and 180 h, and the physical properties were evaluated after each TTT test.

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.

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.

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 (&gt;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.

Conceptual model for radionuclide release, Yucca Mountain, Nevada, USA

A new conceptual model for release rate of radionuclides from the proposed repository for high level nuclear waste located at Yucca Mountain, Nevada is developed. The model predicts that heat generated from radioactive decay combined with the unsaturated environment will lead to an inward flow system that, under many relevant conditions, will slow the release of and sometimes sequester radionuclides at locations of higher heat release and lower water percolation. The amount of protection will be greatest for more concentrated waste forms such as spent fuel and less for glass waste forms. The redistribution and concentration of the radionuclides is anticipated to significantly delay radionuclide release and create a tendency towards gradual release over time that is independent of localized penetrations of metallic barriers.

飽和に近い条件におけるガラス固化体の溶解メカニズムに関する研究の現状と今後の課題

Based on an extended literature survey covering recent studies on high-level radioactive glass dissolution under nearly saturated conditions, we have reached the conclusion that the slow dissolution is controlled by the diffusion of oxonium or boron ions in an altered layer formed on the glass surface. Experimental approaches, such as an elaborate and systematic diffusion experiment using isotopes, were proposed to validate the mechanism. It was also pointed out that the dissolution model applicable to glass waste form performance evaluation takes into account the surface area evolution, the stability of the altered layer, and the interactions with near-field materials.

Calculation of maximum release rates in alternative design changes in the thickness of the buffer for the engineered barrier system (EBS) in deep repository by using Amber code

The multibarrier concept forms the basis for geological disposal concepts in most countries and the guideline states that research and development should aim to demonstrate the feasibility of constructing an engineered barrier system (EBS) which is appropriate for the range of relevant geological conditions. The multibarrier system (EBS) and its functions consisting of the glass waste form, overpack and buffer material was located in a sufficiently stable geological environment. When an overpack comes into contact with groundwater it will start to corrode. The wall thickness will then gradually reduce, and the overpack will eventually fail mechanically when its structural strength can no longer support the stress imposed by the surrounding environment. The requirements that influence the thickness of buffer include nuclide migration retardation and heat conductivity, as well as stress buffering capability, self-sealing ability and workability. The migration retardation function is assumed to be the most important of all these requirements with respect to setting the appropriate thickness of buffer. A consideration of these effects and relationship between buffer thicknesses has determined that a reasonable thickness for the buffer is between 400 mm and 700 mm [AECL, H12]. Therefore, the design thickness of buffer material can range from 0.4 m to 0.7 m to account for manufacturing and stress buffering. In this alternative design case, the thickness of the buffer material is set to 0.4 m, 0.5 m, 0.6 m and 0.7 m. The nuclide migration properties of the buffer material are assumed to be the same (PNC, Development and Management of the Technical Knowledge Base for the Geological Disposal of HLW, Supporting Report 2: “Repository” Engineering Technology). The results of calculation are presented that some nuclides such as Se-79, Tc-99, Pd-107, Th-233, U-236, Pb-210, Ra-226 and Np-237 virtually unchanged in case the maximum release rate from EBS corresponding to change thickness of buffer material. Some nuclides such as Cs-135, Nb-94, Nb-93 m, Zr-93, Sn-126, Th-230, Ph-240, Pu-242, U-233, Ac-227, Pa-231 and Th-229 are very little greater for 40 cm, 50 cm and 60 cm in the maximum release rate compared with 70 cm. Maximum release of nuclides U-235, U-234 and U-238 increases in case of 50 cm and 60 cm thickness of buffer and in case 40 cm are the same as 70 cm thickness because the amount of their parents in case 40 cm will decrease before decay and in case 70 cm amount of these nuclides will decrease due to decayed to other nuclides before release from the buffer, then maximum decay happened in case 50 cm. The maximum release rates of short-lived nuclides such as Cm-245, Am-243, Cm-245, Am-241, Pu-241 and Pu-239 increase significantly due to less decay occurring during the reduced buffer transit time.

A Trade Study for Waste Concepts to Minimize HLW Volume

AbstractAdvanced nuclear fuel reprocessing can partition wastes into groups of common chemistry. This enables new waste management strategies not possible with the plutonium, uranium extraction (PUREX) process alone. Combining all of the metallic fission products in an alloy and the balance as oxides in glass minimizes high level waste (HLW) volume. Implementing a waste management strategy using state-of-the-art combined waste forms and storage to allow radioactive decay and heat dissipation prior to placement in a repository makes it possible to place almost 10x the HLW equivalent of spent nuclear fuel (SNF) in the same repository space. However, using generic costs based on preliminary studies for waste stabilization facilities and separations modules, this analysis shows that combining the non-actinide wastes and using only one glass waste form is the most cost-effective.

The National Nuclear Laboratory and Collaborative University Research in the UK

AbstractThe UK has recently established its first National Nuclear Laboratory (NNL). This has been formed out of the Nexia Solutions organisation, formally the Research and Technology subsidiary of BNFL. Over the next year the NNL will develop so that it can fulfil its mission, which will include the development and maintenance of key skills and undertaking strategic R&amp;D programmes both in the UK and in international collaborations. A key role of the NNL will be to enhance its interactions with universities to facilitate skills transfer into the nuclear industry as well as support its R&amp;D programmes. Over the past decade the NNL and its predecessor has already established close relationships with leading universities in the UK including Sheffield, Manchester, Leeds and Imperial College London. One of the key objectives of the NNL has been to work in an integrated way with university researchers with programmes spanning fundamental research through to applied R&amp;D. This paper describes the future plans for working with universities as the NNL develops, building on the success to date. Joint R&amp;D research programmes already cover many important areas for the nuclear industry. This includes support for the UK Pu disposition programme where R&amp;D at the Immobilisation Science Laboratory, Sheffield has included investigating a range of ceramic and glass wasteforms which has allowed a number of ceramic wasteforms to be down selected for detailed evaluation. Cementation research has included understanding wasteform performance, for example long-term durability. This has involved work on determining free water and the implications for immobilisation of reactive wastes. Other work involving the NNL and Universities, in particular at Leeds and Manchester, has considered the characteristics and behaviour of intermediate and high level waste sludges. This has included determining the chemical speciation of actinides and fission products, and physical properties of active and simulated sludges using experimental and modelling techniques. A significant programme on environmental work has been undertaken by the NNL. In research applicable to low level waste disposal and contaminated land reactive transport modelling is utilised to apply experimental and field based research in environmental geochemistry and radiochemistry undertaken by university collaborators. This paper will describe a number of examples of R&amp;D carried out by NNL in association with its university partners.

Glass fabrication and product consistency testing of lanthanide borosilicate glass for plutonium disposition

The Department of Energy Office of Environmental Management (DOE/EM) plans to conduct the Plutonium Disposition Project at the Savannah River Site (SRS) in Aiken, SC, to disposition excess weapons-usable plutonium. A plutonium glass waste form is a leading candidate for immobilization of the plutonium for subsequent disposition in a geologic repository. The objectives of this present task were to fabricate plutonium-loaded lanthanide borosilicate (LaBS) Frit B glass and perform testing to provide near-term data that will increase confidence that LaBS glass product is suitable for disposal in the proposed Federal Repository. Specifically, testing was conducted in an effort to provide data to Yucca Mountain Project (YMP) personnel for use in performance assessment calculations. Plutonium-containing LaBS glass with the Frit B composition with a 9.5 wt% PuO 2 loading was prepared for testing. Glass was prepared to support glass durability testing via the ASTM product consistency testing (PCT) at Savannah River National Laboratory (SRNL). The glass was characterized with X-ray diffraction (XRD) and scanning electron microscopy coupled with energy dispersive spectroscopy (SEM/EDS) prior to performance testing. This characterization revealed some crystalline PuO 2 inclusions with disk-like morphology present in the as-fabricated, quench-cooled glass. A series of PCTs was conducted at SRNL with varying exposed surface area and test durations. Filtered leachates from these tests were analyzed to determine the dissolved concentrations of key elements. The leachate solutions were also ultrafiltered to quantify colloid formation. Leached solids from select PCTs were examined in an attempt to evaluate the Pu and neutron absorber release behavior from the glass and to investigate formation of alteration phases on the glass surface. A series of PCTs was conducted at 90 °C in ASTM Type 1 water to compare the Pu LaBS Frit B glass durability to current requirements for high level waste (HLW) glass in a geologic repository. The PCT (7-day static test with powdered glass) results on the Pu-containing LaBS Frit B glass at SA/V of ≈2000 m −1 showed that the glass was very durable with an average normalized elemental release value for boron of 0.013 g/m 2. This boron release value is ≈640× lower than normalized boron release from current environmental assessment (EA) glass used for repository acceptance. The PCT-B (7-, 14-, 28- and 56-day, static test with powdered glass) normalized elemental releases were similar to the normalized elemental release values from PCT-A testing, indicating that the LaBS Frit B glass is very durable as measured by the PCT. Normalized plutonium releases were essentially the same within the analytical uncertainty of the ICP-MS methods used to quantify plutonium in the 0.45 μm-filtered leachates and ultra-filtered leachates, indicating that colloidal plutonium species do not form under the PCT conditions used in this study.

Nuclear Waste Glasses - How Durable?

High-level nuclear wastes (HLW) are the liquid effluents that result from the reprocessing of spent nuclear fuel. These wastes are typically solidified in a glass for final disposal in deep geologic formations. At present, there is no geologic repository receiving these vitrified wastes. A primary issue in nuclear waste management is whether there can be societal, regulatory, and political confidence that the radiotoxic constituents of HLW can be safely disposed of for hundreds of thousands of years. If a glass waste form, placed at a depth of hundreds of meters, is stable and essentially insoluble in groundwater, it would be almost impossible for radioactivity to reach the environment. This paper summarizes the state of knowledge of glass performance in a geologic repository and examines the question of whether the long-term stability of the glass and radionuclide retention can be assured.

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.

Immobilisation of radioactive waste in glasses, glass composite materials and ceramics

The basic principles of incorporating high level radioactive waste into glasses, ceramics (Synroc type) and glass composites including glass ceramics are described. Current UK technology uses glass wasteforms for the products of reprocessing, although many countries are temporarily storing the ceramic spent fuel for eventual disposal. Some waste streams may be incorporated into ceramics, but difficult or legacy wastes will require the development of other wasteforms comprising composite systems of crystals and glass. The importance of processing–property–structure (especially durability) relations in such systems over size scales from the atomic to the geological and on timescales to hundreds of thousands of years is highlighted.

Applications of high energy ion beam techniques in environmental science: Investigation associated with glass and ceramic waste forms

High energy ion beam capabilities including Rutherford backscattering spectrometry (RBS) and nuclear reaction analysis (NRA) have been very effectively used in environmental science to investigate the ion-exchange mechanisms in glass waste forms and the effects of irradiation in glass and ceramic waste forms in the past. In this study, RBS and NRA along with SIMNRA simulations were used to monitor the Na depletion and D and 18O uptake in alumina silicate glasses, respectively, after the glass coupons were exposed to aqueous solution. These results show that the formation of a reaction layer and an establishment of a region where diffusion limited ion exchange occur in these glasses during exposure to silica-saturated solutions. Different regions including reaction and diffusion regions were identified on the basis of the depth distributions of these elements. In the case of ceramics, damage accumulation was studied as a function of ion dose at different irradiation temperatures. A sigmoidal dependence of relative disorder on the ion dose was observed. The defect-dechanneling factors were calculated for two irradiated regions in SrTiO 3 using the critical angles determined from the angular yield curves. The dependence of defect-dechanneling parameter on the incident energy was investigated and it was observed that the generated defects are mostly interstitial atoms and amorphous clusters. Thermal recovery experiments were performed to study the damage recovery processes up to a maximum temperature of 870 K.

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.

The effect of Cs 2O additions on HLW wasteform glasses

Physical parameters and structural characteristics are reported for the mixed alkali borosilicate glass system, xCs 2O(100 − x)(MW) (0 ⩽ x ⩽ 9.66) where MW is the BNFL wasteform glass (10.29 mol% Li 2O, 10.53 mol% Na 2O, 18.57 mol% B 2O 3, 60.61 mol% SiO 2). Glass density was found to increase with x whilst the glass transition temperature ( T g) decreased. The fraction of four-coordinated boron (N 4) determined from 11B magic angle spinning nuclear magnetic resonance (MAS NMR) increased with x in a manner consistent with previous reports on the sodium–lithium–borosilicate system and contrary to the Dell model. 29Si MAS NMR was used to identify the fraction of Q 4 (B) and Q 3 units as a function of x. Raman spectra indicated the presence of reedmergnerite and danburite superstructural units. Mass loss measurements showed a non-linear variation with x, having a maximum between 7 and 8 mol% Cs 2O and a correlation was found between the amount of Q 3 units in the solid and the volatilization loss from the melt.

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.

Immobilization of Hanford LAW in iron phosphate glasses

Iron phosphate glasses containing 30 w% of a high sodium and sulfur Hanford low-activity waste (LAW) simulant was successfully melted in electric furnaces at 1000–1050 °C for 2–3 h. No sulfate salt segregation or crystalline phases were detectable in the glassy wasteform when examined by scanning electron microscope and X-ray diffractometer. This suggests that the waste loading in the iron phosphate glasses will not be limited by the SO 3 content of the LAW as is currently the case in borosilicate glasses. At 30 wt% LAW, the iron phosphate glass wasteform satisfies DOE’s product consistency test and vapor hydration test requirements for aqueous chemical durability, but the chemical durability depends upon the overall composition. The high fluidity and electrical conductivity of the iron phosphate melts suggest that iron phosphate wasteforms can be melted in a hot or cold crucible induction melter as an alternative to melting in a joule-heated melter. These properties combined with a significantly higher waste loading offer a considerable savings in time, energy and cost for vitrifying the Hanford LAW in iron phosphate glasses.