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Sample records for waste glass durability

  1. Factors influencing chemical durability of nuclear waste glasses

    International Nuclear Information System (INIS)

    Feng, Xiangdong; Bates, J.K.

    1993-01-01

    A short summary is given of our studies on the major factors that affect the chemical durability of nuclear waste glasses. These factors include glass composition, solution composition, SA/V (ratio of glass surface area to the volume of solution), radiation, and colloidal formation. These investigations have enabled us to gain a better understanding of the chemical durability of nuclear waste glasses and to accumulate.a data base for modeling the long-term durability of waste glass, which will be used in the risk assessment of nuclear waste disposal. This knowledge gained also enhances our ability to formulate optimal waste glass compositions

  2. Thermodynamic model of natural, medieval and nuclear waste glass durability

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Plodinec, M.J.

    1983-01-01

    A thermodynamic model of glass durability based on hydration of structural units has been applied to natural glass, medieval window glasses, and glasses containing nuclear waste. The relative durability predicted from the calculated thermodynamics correlates directly with the experimentally observed release of structural silicon in the leaching solution in short-term laboratory tests. By choosing natural glasses and ancient glasses whose long-term performance is known, and which bracket the durability of waste glasses, the long-term stability of nuclear waste glasses can be interpolated among these materials. The current Savannah River defense waste glass formulation is as durable as natural basalt from the Hanford Reservation (10 6 years old). The thermodynamic hydration energy is shown to be related to the bond energetics of the glass. 69 references, 2 figures, 1 table

  3. Chemical durability of glasses containing radioactive fission product waste

    International Nuclear Information System (INIS)

    Mendel, J.E.; Ross, W.A.

    1974-04-01

    Measurements made to determine the chemical durability of glasses for disposal of radioactive waste are discussed. The term glass covers materials varying from true glass with only minute quantities of crystallites, such as insoluble RuO 2 , to quasi glass-ceramics which are mostly crystalline. Chemical durability requirements and Soxhlet extractor leach tests are discussed

  4. Chemical durability of soda-lime-aluminosilicate glass for radioactive waste vitrification

    International Nuclear Information System (INIS)

    Eppler, F.H.; Yim, M.S.

    1998-01-01

    Vitrification has been identified as one of the most viable waste treatment alternatives for nuclear waste disposal. Currently, the most popular glass compositions being selected for vitrification are the borosilicate family of glasses. Another popular type that has been around in glass industry is the soda-lime-silicate variety, which has often been characterized as the least durable and a poor candidate for radioactive waste vitrification. By replacing the boron constituent with a cheaper substitute, such as silica, the cost of vitrification processing can be reduced. At the same time, addition of network intermediates such as Al 2 O 3 to the glass composition increases the environmental durability of the glass. The objective of this study is to examine the ability of the soda-lime-aluminosilicate glass as an alternative vitrification tool for the disposal of radioactive waste and to investigate the sensitivity of product chemical durability to variations in composition

  5. Evaluation of models of waste glass durability

    International Nuclear Information System (INIS)

    Ellison, A.

    1995-01-01

    The main variable under the control of the waste glass producer is the composition of the glass; thus a need exists to establish functional relationships between the composition of a waste glass and measures of processability, product consistency, and durability. Many years of research show that the structure and properties of a glass depend on its composition, so it seems reasonable to assume that there also is relationship between the composition of a waste glass and its resistance to attack by an aqueous solution. Several models have been developed to describe this dependence, and an evaluation their predictive capabilities is the subject of this paper. The objective is to determine whether any of these models describe the ''correct'' functional relationship between composition and corrosion rate. A more thorough treatment of the relationships between glass composition and durability has been presented elsewhere, and the reader is encouraged to consult it for a more detailed discussion. The models examined in this study are the free energy of hydration model, developed at the Savannah River Laboratory, the structural bond strength model, developed at the Vitreous State Laboratory at the Catholic University of America, and the Composition Variation Study, developed at Pacific Northwest Laboratory

  6. Effect of glass composition on waste form durability: A critical review

    International Nuclear Information System (INIS)

    Ellison, A.J.G.; Mazer, J.J.; Ebert, W.L.

    1994-11-01

    This report reviews literature concerning the relationship between the composition and durability of silicate glasses, particularly glasses proposed for immobilization of radioactive waste. Standard procedures used to perform durability tests are reviewed. It is shown that tests in which a low-surface area sample is brought into contact with a very large volume of solution provide the most accurate measure of the intrinsic durability of a glass composition, whereas high-surface area/low-solution volume tests are a better measure of the response of a glass to changes in solution chemistry induced by a buildup of glass corrosion products. The structural chemistry of silicate and borosilicate glasses is reviewed to identify those components with the strongest cation-anion bonds. A number of examples are discussed in which two or more cations engage in mutual bonding interactions that result in minima or maxima in the rheologic and thermodynamic properties of the glasses at or near particular optimal compositions. It is shown that in simple glass-forming systems such interactions generally enhance the durability of glasses. Moreover, it is shown that experimental results obtained for simple systems can be used to account for durability rankings of much more complex waste glass compositions. Models that purport to predict the rate of corrosion of glasses in short-term durability tests are evaluated using a database of short-term durability test results for a large set of glass compositions. The predictions of these models correlate with the measured durabilities of the glasses when considered in large groupings, but no model evaluated in this review provides accurate estimates of durability for individual glass compositions. Use of these models in long-term durability models is discussed. 230 refs

  7. Production and remediation of low sludge simulated Purex waste glasses, 2: Effects of sludge oxide additions on glass durability

    International Nuclear Information System (INIS)

    Ramsey, W.G.

    1993-01-01

    Glass produced during the Purex 4 campaigns of the Integrated DWPF Melter System (IDMS) and the 774 Research Melter contained a lower fraction of sludge components than targeted by the Product Composition Control System (PCCS). Purex 4 glass was more durable than the benchmark (EA) glass, but was less durable than most other simulated SRS high-level waste glasses. Further, the measured durability of Purex 4 glass was not as well correlated with the durability predicted from the DWPF process control algorithm, probably because the algorithm was developed to predict the durability of SRS high-level waste glasses with higher sludge content than Purex 4. A melter run, designated Purex 4 Remediation, was performed using the 774 Research Melter to determine if the initial PCCS target composition determined for Purex 4 would produce acceptable glass whose durability could be accurately modeled by the DWPF glass durability algorithm. Reagent grade oxides and carbonates were added to Purex 4 melter feed stock to simulate a higher sludge loading. Each canister of glass produced was sampled and the glass durability was determined by the Product Consistency Test method. This document details the durability data and subsequent analysis

  8. Neural network analysis of nuclear waste glass composition vs durability

    International Nuclear Information System (INIS)

    Seibel, C.K.

    1994-01-01

    The relationship between the chemical composition of oxide glasses and their physical properties is poorly understood, but it is becoming more important as vitrification (transformation into glass) of high-level nuclear waste becomes the favored method for long-term storage. The vitrified waste will be stored deep in geologic repositories where it must remain intact for at least 10,000 years. A strong resistance to groundwater exposure; i.c. a slow rate of glass dissolution, is of great importance. This project deals specifically with glass samples developed and tested for the nuclear fuel reprocessing facility near West Valley, New York. This facility needs to dispose of approximately 2.2 million liters of high-level radioactive liquid waste currently stored in stainless steel tanks. A self-organizing, artificial neural network was used to analyze the trends in the glass dissolution data for the effects of composition and the resulting durability of borosilicate glasses in an aqueous environment. This durability data can be used to systematically optimize the properties of the complex nuclear glasses and slow the dissolution rate of radionuclides into the environment

  9. Influence of Some Nuclear Waste on The Durability and Mechanical Properties of Borosilicate glass

    International Nuclear Information System (INIS)

    El-Alaily, N.A.

    2003-01-01

    Various glass systems have been shown to be suitable for producing waste glass forms that are thermally and mechanically stable and exhibit good chemical durability. In this study borosilicate glass containing sodium oxide and aluminum oxide was prepared as a host for high level nuclear waste. The glass durability when the samples were immersed either in distilled water or ground water at 70 degree was studied. The density, porosity and mechanical properties were also investigated. The effects of exposing the samples immersed in groundwater to gamma rays in the glass durability and all other mentioned properties were also studied. The results showed that immersing the glass in ground water causing a decrease in the glass durability. The exposure of the glass immersed in ground water to the gamma rays increases the durability of the glass. The mechanical properties of the prepared glass were good. Although these properties decrease for the corroded glass but they were still good

  10. Liquidus temperature and chemical durability of selected glasses to immobilize rare earth oxides waste

    Energy Technology Data Exchange (ETDEWEB)

    Mohd Fadzil, Syazwani, E-mail: mfsyazwani86@postech.ac.kr [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, 790784 Pohang, Gyeongbuk (Korea, Republic of); School of Applied Physics, Faculty of Science and Technology, The National University of Malaysia, 43650 Bandar Baru Bangi, Selangor (Malaysia); Hrma, Pavel [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, 790784 Pohang, Gyeongbuk (Korea, Republic of); Pacific Northwest National Laboratory, P.O. Box 999, Richland, WA (United States); Schweiger, Michael J.; Riley, Brian J. [Pacific Northwest National Laboratory, P.O. Box 999, Richland, WA (United States)

    2015-10-15

    Pyroprocessing is are processing method for managing and reusing used nuclear fuel (UNF) by dissolving it in an electrorefiner with a molten alkali or alkaline earth chloride salt mixture while avoiding wet reprocessing. Pyroprocessing UNF with a LiCl–KCl eutectic salt releases the fission products from the fuel and generates a variety of metallic and salt-based species, including rare earth (RE) chlorides. If the RE-chlorides are converted to oxides, borosilicate glass is a prime candidate for their immobilization because of its durability and ability to dissolve almost any RE waste component into the glass matrix at high loadings. Crystallization that occurs in waste glasses as the waste loading increases may complicate glass processing and affect the product quality. This work compares three types of borosilicate glasses in terms of liquidus temperature (T{sub L}): the International Simple Glass designed by the International Working Group, sodium borosilicate glass developed by Korea Hydro and Nuclear Power, and the lanthanide aluminoborosilicate (LABS) glass established in the United States. The LABS glass allows the highest waste loadings (over 50 mass% RE{sub 2}O{sub 3}) while possessing an acceptable chemical durability. - Highlights: • We investigated crystallization in borosilicate glasses containing rare earth oxides. • New crystallinity and durability data are shown for glasses proposed in the literature. • Both liquidus temperature and chemical durability increased as the waste loading increased.

  11. Characterization and durability testing of a glass-bonded ceramic waste form

    International Nuclear Information System (INIS)

    Johnson, S. G.

    1998-01-01

    Argonne National Laboratory is developing a glass bonded ceramic waste form for encapsulating the fission products and transuranics from the conditioning of metallic reactor fuel. This waste form is currently being scaled to the multi-kilogram size for encapsulation of actual high level waste. This paper will present characterization and durability testing of the ceramic waste form. An emphasis on results from application of glass durability tests such as the Product Consistency Test and characterization methods such as X-ray diffraction and scanning electron microscopy. The information presented is based on a suite of tests utilized for assessing product quality during scale-up and parametric testing

  12. Compositional threshold for Nuclear Waste Glass Durability

    International Nuclear Information System (INIS)

    Kruger, Albert A.; Farooqi, Rahmatullah; Hrma, Pavel R.

    2013-01-01

    Within the composition space of glasses, a distinct threshold appears to exist that separates 'good' glasses, i.e., those which are sufficiently durable, from 'bad' glasses of a low durability. The objective of our research is to clarify the origin of this threshold by exploring the relationship between glass composition, glass structure and chemical durability around the threshold region

  13. Standard test methods for determining chemical durability of nuclear, hazardous, and mixed waste glasses and multiphase glass ceramics: The product consistency test (PCT)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 These product consistency test methods A and B evaluate the chemical durability of homogeneous glasses, phase separated glasses, devitrified glasses, glass ceramics, and/or multiphase glass ceramic waste forms hereafter collectively referred to as “glass waste forms” by measuring the concentrations of the chemical species released to a test solution. 1.1.1 Test Method A is a seven-day chemical durability test performed at 90 ± 2°C in a leachant of ASTM-Type I water. The test method is static and conducted in stainless steel vessels. Test Method A can specifically be used to evaluate whether the chemical durability and elemental release characteristics of nuclear, hazardous, and mixed glass waste forms have been consistently controlled during production. This test method is applicable to radioactive and simulated glass waste forms as defined above. 1.1.2 Test Method B is a durability test that allows testing at various test durations, test temperatures, mesh size, mass of sample, leachant volume, a...

  14. Composition models for the viscosity and chemical durability of West Valley related nuclear waste glasses

    International Nuclear Information System (INIS)

    Feng, X.; Saad, E.E.; Freeborn, W.P.; Macedo, P.B.; Pegg, I.L.; Sassoon, R.E.; Barkatt, A.; Finger, S.M.

    1988-01-01

    There are two important criteria that must be satisfied by a nuclear waste glass durability and processability. The chemical composition of the glass must be such that it does not dissolve or erode appreciably faster than the decay of the radioactive materials embedded in it. The second criterion, processability, means that the glass must melt with ease, must be easily pourable, and must not crystallize appreciably. This paper summarizes the development of simple models for predicting the durability and viscosity of nuclear waste glasses from their composition

  15. Hanford enhanced waste glass characterization. Influence of composition on chemical durability

    International Nuclear Information System (INIS)

    Fox, K. M.; Edwards, T. B.

    2016-01-01

    This report provides a review of the complete high-level waste (HLW) and low-activity waste (LAW) data sets for the glasses recently fabricated at Pacific Northwest National Laboratory and characterized at Savannah River National Laboratory (SRNL). The review is from the perspective of relating the chemical durability performance to the compositions of these study glasses, since the characterization work at SRNL focused on chemical analysis and ASTM Product Consistency Test (PCT) performance.

  16. Canonical correlation of waste glass compositions and durability, including pH

    International Nuclear Information System (INIS)

    Oeksoy, D.; Pye, L.D.; Bickford, D.F.; Ramsey, W.G.

    1993-01-01

    Control of waste glass durability is a major concern in the immobilization of radioactive and mixed wastes. Leaching rate in standardized laboratory tests is being used as a demonstration of consistency of the response of waste glasses in the final disposal environment. The leaching of silicate and borosilicate glasses containing alkali or alkaline earth elements is known to be autocatalytic, in that the initial ion exchange of alkali in the glass for hydrogen ions in water results in the formation of OH and increases the pH of the leachate. The increased pH then increases the rate of silicate network attack, accelerating the leaching effect. In well formulated glasses this effect reaches a thermodynamic equilibrium when leachate saturation of a critical species, such as silica, or a dynamic equilibrium is reached when the pH shift caused by incremental leaching has negligible effect on pH. This report analyzes results of a seven leach test on waste glasses

  17. Durability, mechanical, and thermal properties of experimental glass-ceramic forms for immobilizing ICPP high level waste

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1990-01-01

    The high-level liquid waste generated at the Idaho Chemical Processing Plant (ICPP) is routinely solidified into granular calcined high-level waste (HLW) and stored onsite. Research is being conducted at the ICPP on methods of immobilizing the HLW, including developing a durable glass-ceramic form which has the potential to significantly reduce the final waste volume by up to 60% compared to a glass form. Simulated, pilot plant, non-radioactive, calcines similar to the composition of the calcined HLW and glass forming additives are used to produce experimental glass-ceramic forms. The objective of the research reported in this paper is to study the impact of ground calcine particle size on durability and mechanical and thermal properties of experimental glass-ceramic forms

  18. Effects of crystallization on thermal properties and chemical durability of the glasses containing simulated high level radioactive wastes

    International Nuclear Information System (INIS)

    Kawamoto, Takamichi; Terai, Ryohei; Hara, Shigeo

    1978-01-01

    In order to improve the thermodynamic stability of the glasses containing high level radioactive wastes, the conversion to glass-ceramics by the heat-treatment was carried out with two kinds of glasses, and the change of thermal properties and chemical durability by crystallization was investigated. One of the glasses has a composition of SiO 2 -Al 2 O 3 -ZnO-TiO 2 system, and another one has a composition which could grow the nephelite crystals from Na 2 O in wastes and Al 2 O 3 and SiO 2 added as glass-forming materials. Transition and yield points shifted to higher temperatures by the conversion and the glass-ceramics were found to be more stable than the original glasses. The glass-ceramics of the composition of SiO 2 -Al 2 O 3 -ZnO-TiO 2 showed poor durability, whereas the chemical durability of the glass-ceramics containing nephelite crystals was considerably improved. In the latter case, improvement of the durability is attributable to that some parts of glass are converted to nephelite crystals and the crystals are more durable than glass under most conditions. (auth.)

  19. First-order model for durability of Hanford waste glasses as a function of composition

    International Nuclear Information System (INIS)

    Hrma, P.; Piepel, G.F.; Schweiger, M.J.; Smith, D.E.

    1992-04-01

    Two standard chemical durability tests, the static leach test MCC-1 and product consistency test PCT, were conducted on simulated borosilicate glasses that encompass the expected range of compositions to be produced in the Hanford Waste Vitrification Plant (HWVP). A first-order empirical model was fitted to the data from each test method. The results indicate that glass durability is increased by addition of Al 2 O 3 , moderately increased by addition of ZrO 2 and SiO 2 , and decreased by addition of Li 2 O, Na 2 O, B 2 O 3 , and MgO. Addition of Fe 2 O 3 and CaO produce an indifferent or reducing effect on durability according to the test method. This behavior and a statistically significant lack of fit are attributed to the effects of multiple chemical reactions occurring during glass-water interaction. Liquid-liquid immiscibility is suspected to be responsible for extremely low durability of some glasses

  20. Chemical durability of borosilicate glasses containing simulated high-level nuclear wastes, 1

    International Nuclear Information System (INIS)

    Hara, Shigeo; Terai, Ryohei; Yamanaka, Hiroshi

    1983-01-01

    The Soxhlet-type leaching test apparatus has been developed to evaluate the chemical durability of some borosilicate glasses containing simulated High-Level nuclear Wastes, HLW. After the leaching over the temperature range of 50 0 -95 0 C, the weight loss of specimens with time was determined on both the samples of blocks and grains, and various components dissolved into water were analyzed by atomic absorption and colorimetry technique. It was found that Soxhlet-type test method was more useful than JIS test method, because the specimens in Soxhlet type apparatus were forced always to react with pure water and the mechanism of leaching could be evaluate accurately. The chemical durability of commercial glasses decreases generally with increasing of alkali contents in glasses. In the case of these borosilicate glasses containing HLW, however, the leachability was apparently independent on the alkali contents because of the complexity of these glass compositions. The variation of leaching rate with temperature suggests that dissolution mechanism changes with temperature. (author)

  1. Effect of lead species on the durability of simulated nuclear waste glass

    International Nuclear Information System (INIS)

    Kuchinski, F.A.

    1987-01-01

    It has been shown that the incorporation of lead metal into the corrosion environment reduces the leaching rate of nuclear waste glasses. The present study evaluated the effects of lead metal, oxides, alloys, glasses and soluble species on the corrosion rate of a waste glass. The inherent durability of nuclear waste glasses comes from the about due to the insoluble surface film developed during corrosion. This surface film, enriched with iron, aluminum and calcium acts as a diffusion barrier to further corrosion. Except for PbO 2 , all lead species inhibited glass corrosion due to the formation of a surface film enriched in lead. No corroded glass layer was observed below the lead surface layer. Also, no glass corrosion products were found on the lead surface, except for small amounts of silicon. The transport and deposition of lead on the glass surface appears to be the key factors in preventing glass corrosion. At high glass surface area to volume ratios, the glass corroded considerably at short times since the dissolved lead source could not coat the entire glass surface rapidly enough to prevent continued corrosion. Also, experimental solution values did not agree with thermodynamics model predictions. This suggests that kinetic factors, namely diffusion barriers, are controlling the glass corrosion rate

  2. Preliminary results of durability testing with borosilicate glass compositions

    International Nuclear Information System (INIS)

    Adel-Hadadi, M.; Adiga, R.; Barkatt, Aa.

    1987-01-01

    This is a report on the first year of research conducted at the Vitreous State Laboratory of the Catholic University of America in support of the West Valley Demonstration Project. One objective is the vitrification of liquid waste generated by previous nuclear fuel reprocessing. This work has been directed principally at the problem of glass composition optimization. This has necessitated the development of a coordinated program of glass production, durability measurements, and processability assessment. A small-scale continuous melter has been constructed for melting uranium and thorium containing glasses and for studying glass processing characteristics. Glass viscosities have been measured over a range of temperatures. A large number of glasses have also been produced in small crucible melts. Glass durability has been assessed using four types of leach tests: MCC-3, MCC-1, IAEA/ISO, and pulsed-flow tests. Extensive data from these tests are reported. The data have led to the design of very durable glasses (comparable to the Savannah River Laboratory Defense Waste Reference Glass) which have the requisite waste loading and processing characteristics. 14 refs., 4 figs., 77 tabs

  3. Development of an ASTM standard glass durability test, the Product Consistency Test (PCT), for high level radioactive waste glass

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Ramsey, W.G.

    1994-01-01

    The nation's first, and the world's largest, facility to immobilize high-level nuclear waste in durable borosilicate glass has started operation at the Savannah River Site (SRS) in Aiken, South Carolina. The product specifications on the glass wasteform produced in the Defense Waste Processing Facility (DWPF) required extensive characterization of the glass product before actual production began and for continued characterization during production. To aid in this characterization, a glass durability (leach) test was needed that was easily reproducible, could be performed remotely on highly radioactive samples, and could yield results rapidly. Several standard leach tests were examined with a variety of test configurations. Using existing tests as a starting point, the DWPF Product Consistency Test (PCT was developed in which crushed glass samples are exposed to 90 ± 2 degree C deionized water for seven days. Based on extensive testing, including a seven-laboratory round robin and confirmatory testing with radioactive samples, the PCT is very reproducible, yields reliable results rapidly, and can be performed in shielded cell facilities with radioactive samples

  4. Durable Glass For Thousands Of Years

    International Nuclear Information System (INIS)

    Jantzen, C.

    2009-01-01

    The durability of natural glasses on geological time scales and ancient glasses for thousands of years is well documented. The necessity to predict the durability of high level nuclear waste (HLW) glasses on extended time scales has led to various thermodynamic and kinetic approaches. Advances in the measurement of medium range order (MRO) in glasses has led to the understanding that the molecular structure of a glass, and thus the glass composition, controls the glass durability by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. During the early stages of glass dissolution, a 'gel' layer resembling a membrane forms through which ions exchange between the glass and the leachant. The hydrated gel layer exhibits acid/base properties which are manifested as the pH dependence of the thickness and nature of the gel layer. The gel layer ages into clay or zeolite minerals by Ostwald ripening. Zeolite mineral assemblages (higher pH and Al 3+ rich glasses) may cause the dissolution rate to increase which is undesirable for long-term performance of glass in the environment. Thermodynamic and structural approaches to the prediction of glass durability are compared versus Ostwald ripening.

  5. DURABLE GLASS FOR THOUSANDS OF YEARS

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C.

    2009-12-04

    The durability of natural glasses on geological time scales and ancient glasses for thousands of years is well documented. The necessity to predict the durability of high level nuclear waste (HLW) glasses on extended time scales has led to various thermodynamic and kinetic approaches. Advances in the measurement of medium range order (MRO) in glasses has led to the understanding that the molecular structure of a glass, and thus the glass composition, controls the glass durability by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. During the early stages of glass dissolution, a 'gel' layer resembling a membrane forms through which ions exchange between the glass and the leachant. The hydrated gel layer exhibits acid/base properties which are manifested as the pH dependence of the thickness and nature of the gel layer. The gel layer ages into clay or zeolite minerals by Ostwald ripening. Zeolite mineral assemblages (higher pH and Al{sup 3+} rich glasses) may cause the dissolution rate to increase which is undesirable for long-term performance of glass in the environment. Thermodynamic and structural approaches to the prediction of glass durability are compared versus Ostwald ripening.

  6. Systems approach to nuclear waste glass development

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1986-01-01

    Development of a host solid for the immobilization of nuclear waste has focused on various vitreous wasteforms. The systems approach requires that parameters affecting product performance and processing be considered simultaneously. Application of the systems approach indicates that borosilicate glasses are, overall, the most suitable glasses for the immobilization of nuclear waste. Phosphate glasses are highly durable; but the glass melts are highly corrosive and the glasses have poor thermal stability and low solubility for many waste components. High-silica glasses have good chemical durability, thermal stability, and mechanical stability, but the associated high melting temperatures increase volatilization of hazardous species in the waste. Borosilicate glasses are chemically durable and are stable both thermally and mechanically. The borosilicate melts are generally less corrosive than commercial glasses, and the melt temperature miimizes excessive volatility of hazardous species. Optimization of borosilicate waste glass formulations has led to their acceptance as the reference nuclear wasteform in the United States, United Kingdom, Belgium, Germany, France, Sweden, Switzerland, and Japan

  7. A structural bond strength model for glass durability

    International Nuclear Information System (INIS)

    Feng, Xiangdong; Metzger, T.B.

    1996-01-01

    A glass durability model, structural bond strength (SBS) model was developed to correlate glass durability with its composition. This model assumes that the strengths of the bonds between cations and oxygens and the structural roles of the individual elements in the glass arc the predominant factors controlling the composition dependence of the chemical durability of glasses. The structural roles of oxides in glass are classified as network formers, network breakers, and intermediates. The structural roles of the oxides depend upon glass composition and the redox state of oxides. Al 2 O 3 , ZrO 2 , Fe 2 O 3 , and B 2 O 3 are assigned as network formers only when there are sufficient alkalis to bind with these oxides. CaO can also improve durability by sharing non-bridging oxygen with alkalis, relieving SiO 2 from alkalis. The percolation phenomenon in glass is also taken into account. The SBS model is applied to correlate the 7-day product consistency test durability of 42 low-level waste glasses with their composition with an R 2 of 0.87, which is better than 0.81 obtained with an eight-coefficient empirical first-order mixture model on the same data set

  8. The chemical durability of alkali aluminosilicate glasses

    International Nuclear Information System (INIS)

    Tait, J.C.; Mandolesi, D.L.

    1983-09-01

    The aqueous durabilities of a series of glasses based on the sodium aluminosilicate system (Na 2 O-Al 2 O 3 -SiO 2 ) have been studied. The effects of molecular substitution of K 2 O or CaO for Na 2 O, and B 2 O 3 for Al 2 O 3 have been investigated. The temperature dependence of leaching in the Na 2 O-B 2 O 3 -Al 2 O 3 -SiO 2 system was studied with glasses containing 2 wt percent simulated UO 2 fuel recycle waste. The results confirm that aluminosilicate glasses are more durable than their borosilicate counterparts. The leaching results are explained in terms of glass structure and bonding, and a general leaching mechanism for aluminosilicate glasses is presented

  9. Durability of glasses from the Hg-doped Integrated DWPF Melter System (IDMS) campaign

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1992-01-01

    The Integrated DWPF Melter System (IDMS) for the vitrification of high-level radioactive wastes is designed and constructed to be a 1/9th scale prototype of the full scale Defense Waste Processing Facility (DWPF) melter. The IDMS facility is the first engineering scale melter system capable of processing mercury, and flowsheet levels of halides and noble metals. In order to determine the effects of mercury on the feed preparation process, the off-gas chemistry, glass melting behavior, and glass durability, a three-run mercury (Hg) campaign was conducted. The glasses produced during the Hg campaign were composed of Batch 1 sludge, simulated precipitate hydrolysis aqueous product (PHA) from the Precipitate Hydrolysis Experimental Facility (PHEF), and Frit 202. The glasses were produced using the DWPF process/product models for glass durability, viscosity, and liquidus. The durability model indicated that the glasses would all be more durable than the glass qualified in the DWPF Environmental Assessment (EA). The glass quality was verified by performing the Product Consistency Test (PCT) which was designed for glass durability testing in the DWPF

  10. Glasses and nuclear waste vitrification

    International Nuclear Information System (INIS)

    Ojovan, Michael I.

    2012-01-01

    Glass is an amorphous solid material which behaves like an isotropic crystal. Atomic structure of glass lacks long-range order but possesses short and most probably medium range order. Compared to crystalline materials of the same composition glasses are metastable materials however crystallisation processes are kinetically impeded within times which typically exceed the age of universe. The physical and chemical durability of glasses combined with their high tolerance to compositional changes makes glasses irreplaceable when hazardous waste needs immobilisation for safe long-term storage, transportation and consequent disposal. Immobilisation of radioactive waste in glassy materials using vitrification has been used successfully for several decades. Nuclear waste vitrification is attractive because of its flexibility, the large number of elements which can be incorporated in the glass, its high corrosion durability and the reduced volume of the resulting wasteform. Vitrification involves melting of waste materials with glass-forming additives so that the final vitreous product incorporates the waste contaminants in its macro- and micro-structure. Hazardous waste constituents are immobilised either by direct incorporation into the glass structure or by encapsulation when the final glassy material can be in form of a glass composite material. Both borosilicate and phosphate glasses are currently used to immobilise nuclear wastes. In addition to relatively homogeneous glasses novel glass composite materials are used to immobilise problematic waste streams. (author)

  11. The effects of crystallization and residual glass on the chemical durability of iron phosphate waste forms containing 40 wt% of a high MoO3 Collins-CLT waste

    Science.gov (United States)

    Hsu, Jen-Hsien; Bai, Jincheng; Kim, Cheol-Woon; Brow, Richard K.; Szabo, Joe; Zervos, Adam

    2018-03-01

    The effects of cooling rate on the chemical durability of iron phosphate waste forms containing up to 40 wt% of a high MoO3 Collins-CLT waste simulant were determined at 90 °C using the product consistency test (PCT). The waste form, designated 40wt%-5, meets appropriate Department of Energy (DOE) standards when rapidly quenched from the melt (as-cast) and after slow cooling following the CCC (canister centerline cooling)-protocol, although the quenched glass is more durable. The analysis of samples from the vapor hydration test (VHT) and the aqueous corrosion test (differential recession test) reveals that rare earth orthophosphate (monazite) and Zr-pyrophosphate crystals that form on cooling are more durable than the residual glass in the 40wt%-5 waste form. The residual glass in the CCC-treated samples has a greater average phosphate chain length and a lower Fe/P ratio, and those contribute to its faster corrosion kinetics.

  12. Glass science tutorial: Lecture No. 7, Waste glass technology for Hanford

    International Nuclear Information System (INIS)

    Kruger, A.A.

    1995-07-01

    This paper presents the details of the waste glass tutorial session that was held to promote knowledge of waste glass technology and how this can be used at the Hanford Reservation. Topics discussed include: glass properties; statistical approach to glass development; processing properties of nuclear waste glass; glass composition and the effects of composition on durability; model comparisons of free energy of hydration; LLW glass structure; glass crystallization; amorphous phase separation; corrosion of refractories and electrodes in waste glass melters; and glass formulation for maximum waste loading

  13. Talc-silicon glass-ceramic waste forms for immobilization of high- level calcined waste

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1993-06-01

    Talc-silicon glass-ceramic waste forms are being evaluated as candidates for immobilization of the high level calcined waste stored onsite at the Idaho Chemical Processing Plant. These glass-ceramic waste forms were prepared by hot isostatically pressing a mixture of simulated nonradioactive high level calcined waste, talc, silicon and aluminum metal additives. The waste forms were characterized for density, chemical durability, and glass and crystalline phase compositions. The results indicate improved density and chemical durability as the silicon content is increased

  14. Crystallization in high-level waste glass: A review of glass theory and noteworthy literature

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-01

    There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-level waste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO₄) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe,Cr)₂O₄), while not detrimental to glass durability, can cause an array of processing problems inside of HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic, thermodynamic, and viscosity arguments, and several highlights of glass-crystallization research, as it pertains to high-level waste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies.

  15. Nanoporous Glasses for Nuclear Waste Containment

    Directory of Open Access Journals (Sweden)

    Thierry Woignier

    2016-01-01

    Full Text Available Research is in progress to incorporate nuclear waste in new matrices with high structural stability, resistance to thermal shock, and high chemical durability. Interactions with water are important for materials used as a containment matrix for the radio nuclides. It is indispensable to improve their chemical durability to limit the possible release of radioactive chemical species, if the glass structure is attacked by corrosion. By associating high structural stability and high chemical durability, silica glass optimizes the properties of a suitable host matrix. According to an easy sintering stage, nanoporous glasses such as xerogels, aerogels, and composite gels are alternative ways to synthesize silica glass at relatively low temperatures (≈1,000–1,200°C. Nuclear wastes exist as aqueous salt solutions and we propose using the open pore structure of the nanoporous glass to enable migration of the solution throughout the solid volume. The loaded material is then sintered, thereby trapping the radioactive chemical species. The structure of the sintered materials (glass ceramics is that of nanocomposites: actinide phases (~100 nm embedded in a vitreous silica matrix. Our results showed a large improvement in the chemical durability of glass ceramic over conventional nuclear glass.

  16. Durability of simulated waste glass: effects of pressure and formation of surface layers

    International Nuclear Information System (INIS)

    Wicks, G.G.; Mosley, W.C.; Whitkop, P.G.; Saturday, K.A.

    1981-01-01

    The leaching behavior of simulated Savannah River Plant (SRP) waste glass was studied at elevated pressures and anticipated storage temperatures. An integrated approach, which combined leachate solution analyses with both bulk and surface studies, was used to study the corrosion process. Compositions of leachates were evaluated by colorimetry and atomic absorption. Used in the bulk and surface analyses were optical microscopy, scanning electron microscopy, x-ray energy spectroscopy, wide-angle x-ray, diffraction, electron microprobe analysis, infrared reflectance spectroscopy, electron spectroscopy for chemical analysis, and Auger electron spectroscopy. Results from this study show that there is no significant adverse effect of pressure, up to 1500 psi and 90 0 C, on the chemical durability of simulated SPR waste glass leached for one month in deionized water. In addition, the leached glass surface layer was characterized by an adsorbed film rich in minor constituents from the glass. This film remained on the glass surface even after leaching in relatively alkaline solutions at elevated pressures at 90 0 C for one month. The sample surface area to volume of leachant ratios (SA/V) was 10:1 cm -1 and 1:10 cm -1 . The corrosion mechanisms and surface and subsurface layers produced will be discussed along with the potential importance of these results to repository storage

  17. Glass Durability Modeling, Activated Complex Theory (ACT)

    International Nuclear Information System (INIS)

    CAROL, JANTZEN

    2005-01-01

    The most important requirement for high-level waste glass acceptance for disposal in a geological repository is the chemical durability, expressed as a glass dissolution rate. During the early stages of glass dissolution in near static conditions that represent a repository disposal environment, a gel layer resembling a membrane forms on the glass surface through which ions exchange between the glass and the leachant. The hydrated gel layer exhibits acid/base properties which are manifested as the pH dependence of the thickness and nature of the gel layer. The gel layer has been found to age into either clay mineral assemblages or zeolite mineral assemblages. The formation of one phase preferentially over the other has been experimentally related to changes in the pH of the leachant and related to the relative amounts of Al +3 and Fe +3 in a glass. The formation of clay mineral assemblages on the leached glass surface layers ,lower pH and Fe +3 rich glasses, causes the dissolution rate to slow to a long-term steady state rate. The formation of zeolite mineral assemblages ,higher pH and Al +3 rich glasses, on leached glass surface layers causes the dissolution rate to increase and return to the initial high forward rate. The return to the forward dissolution rate is undesirable for long-term performance of glass in a disposal environment. An investigation into the role of glass stoichiometry, in terms of the quasi-crystalline mineral species in a glass, has shown that the chemistry and structure in the parent glass appear to control the activated surface complexes that form in the leached layers, and these mineral complexes ,some Fe +3 rich and some Al +3 rich, play a role in whether or not clays or zeolites are the dominant species formed on the leached glass surface. The chemistry and structure, in terms of Q distributions of the parent glass, are well represented by the atomic ratios of the glass forming components. Thus, glass dissolution modeling using simple

  18. First principles process-product models for vitrification of nuclear waste: Relationship of glass composition to glass viscosity, resistivity, liquidus temperature, and durability

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1991-01-01

    Borosilicate glasses will be used in the USA and in Europe to immobilize radioactive high level liquid wastes (HLLW) for ultimate geologic disposal. Process and product quality models based on glass composition simplify the fabrication of the borosilicate glass while ensuring glass processability and quality. The process model for glass viscosity is based on a relationship between the glass composition and its structural polymerization. The relationship between glass viscosity and electrical resistivity is also shown to relate to glass polymerization. The process model for glass liquidus temperature calculates the solubility of the liquidus phases based on the free energies of formation of the precipitating species. The durability product quality model is based on the calculation of the thermodynamic hydration free energy from the glass composition

  19. The chemical durability of glasses suitable for the storage of high level radioactive wastes, (1)

    International Nuclear Information System (INIS)

    Terai, Ryohei; Hara, Shigeo; Kawamoto, Takamichi; Nanbu, Tadahiko; Nakamura, Takao.

    1975-01-01

    To develop the glassy materials suitable for the long-term storage of high level radioactive wastes, the chemical durability of the glasses of borax-alumina-silica system has been investigated. The test was carried out by the following three ways, (1) glass-disk immersion method, (2) continuous leach method and (3) method prescribed in JIS-R3502. In the continuous leach method, glass grains were exposed to circulating water at a constant temperature for a week to obtain the leach factor or leach rate. It was found from the experimental results that, as the silica content increased, the melting temperature of the glasses progressively increased and the chemical durability was considerably improved, and that B 2 O 3 and Na 2 O constituents were preferentially dissolved in water leaving relatively insoluble components such as SiO 2 and Al 2 O 3 . The rate at which B 2 O 3 and Na 2 O in glass are leached out is governed by three processes, that is, (1) the boundary reaction on the glass surface, (2) the diffusion process through the hydrated layer, and (3) the disintegration of hydrated layer. The first process probably corresponds to the hydration of boric oxides on the glass surface or to the ion exchange between protons in solution and Na + ions in glass, and the second process seems to correspond to the diffusion of protons through the hydrated layer on the glass surface. Although the ratio of [Na-BO 4 ]/[BO 3 ] in the borax-silica glasses was determined to be 0.5 by means of NMR measurement, Na 2 O/B 2 O 3 ratio in leached solution was less than 0.5, indicating that [BO 3 ] groups in glass were more soluble than [Na-BO 4 ] groups. From the viewpoint of appreciation of safety, the chemical durability of the glasses of borax-aluminasilica system was rather unsatisfactory, but that of the glasses containing silica in quantities was comparable to the soda-lime silicate sheet glasses. (auth.)

  20. Crystallization in high-level waste glass: A review of glass theory and noteworthy literature

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-18

    There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-level waste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO4) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe, Cr)2O4), while not detrimental to glass durability, can cause an array of processing problems inside HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic, thermodynamic, and viscosity arguments, and several highlights of glass-crystallization research, as it pertains to high-level waste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies. Higher waste loadings and more efficient processing strategies will reduce the overall HLW Hanford Tank Waste Treatment and Immobilization Plant (WTP) vitrification facilities mission life.

  1. Durability of Silicate Glasses: An Historical Approach

    International Nuclear Information System (INIS)

    Farges, Francois; Etcheverry, Marie-Pierre; Haddi, Amine; Trocellier, Patrick; Curti, Enzo; Brown, Gordon E. Jr.

    2007-01-01

    We present a short review of current theories of glass weathering, including glass dissolution, and hydrolysis of nuclear waste glasses, and leaching of historical glasses from an XAFS perspective. The results of various laboratory leaching experiments at different timescales (30 days to 12 years) are compared with results for historical glasses that were weathered by atmospheric gases and soil waters over 500 to 3000 years. Good agreement is found between laboratory experiments and slowly leached historical glasses, with a strong enrichment of metals at the water/gel interface. Depending on the nature of the transition elements originally dissolved in the melt, increasing elemental distributions are expected to increase with time for a given glass durability context

  2. Durability of Silicate Glasses: An Historical Approach

    Energy Technology Data Exchange (ETDEWEB)

    Farges, Francois; /Museum Natl. Hist. Natur. /Stanford U., Geo. Environ. Sci.; Etcheverry, Marie-Pierre; /Marne la Vallee U.; Haddi, Amine; /Marne la Valle U.; Trocellier,; /Saclay; Curti, Enzo; /PSI, Villigen; Brown, Gordon E., Jr.; /SLAC, SSRL

    2007-01-02

    We present a short review of current theories of glass weathering, including glass dissolution, and hydrolysis of nuclear waste glasses, and leaching of historical glasses from an XAFS perspective. The results of various laboratory leaching experiments at different timescales (30 days to 12 years) are compared with results for historical glasses that were weathered by atmospheric gases and soil waters over 500 to 3000 years. Good agreement is found between laboratory experiments and slowly leached historical glasses, with a strong enrichment of metals at the water/gel interface. Depending on the nature of the transition elements originally dissolved in the melt, increasing elemental distributions are expected to increase with time for a given glass durability context.

  3. Durability testing with West Valley borosilicate glass composition- Phase II

    International Nuclear Information System (INIS)

    Macedo, P.B.; Finger, S.M.; Barkatt, A.A.; Pegg, I.L.; Feng, X.; Freeborn, W.P.

    1988-06-01

    This report presents the research performed by the Catholic University of America Vitreous State Laboratory (VSL) during FY 1987 in support of the West Valley Demonstration Project (WVDP) nuclear waste vitrification process. A principal objective of this work is the optimization of the glass composition be used for the vitrification of the liquid high-level waste generated at West Valley during nuclear fuel reprocessing. This report discusses (1) the experimental investigations to optimize the reference glass composition (the current leading candidates are WVCM-50 and ATM-10) for the WVDP vitrification process; (2) the systematic experimental investigation performed to determine the effects of compositional variations in WVCM-50 and WV-205 reference glasses on their viscosity and durability (including initial results of long-term leach tests of WVCM-50 under repository conditions); (3) the development of short-time and predictive leach tests; (4) the development of a process model for the West Valley vitrification process which predicts the range of glass compositions which may be encountered during normal operations and the effects of deviations in process control parameters; and (5) the development of product models for predicting the durability and viscosity of nuclear waste glasses

  4. Composition effects on chemical durability and viscosity of nuclear waste glasses - systematic studies and structural thermodynamic models

    International Nuclear Information System (INIS)

    Feng, X.

    1988-01-01

    Two of the primary criteria for the acceptability of nuclear waste glasses are their durability, i.e. chemical resistance to aqueous attack for 10 4 to 10 5 years, and processability, which requires their viscosity at the desired melt temperature to be sufficiently low. Chapter 3 presents the results of systematic composition variation studies around the preliminary reference glass composition WV205 and an atomistic interpretation of the effects of individual oxides. Chapter 4 is concerned with modifications of the Jantzen-Plodinec hydration model which takes into account formation of complex aluminosilicate compounds in the glass. Chapter 5 is devoted to the development and validation of the structural-thermodynamic model for both durability and viscosity. This model assumes the strength of bonds between atoms to be the controlling factor in the composition dependence of these glass properties. The binding strengths are derived from the known heats of formation and the structural roles of constituent oxides. Since the coordination state of various oxides in the glass is temperature dependent and cation size has opposite effects on the two properties, the correlation between melt viscosity and rate of corrosion at low temperature is not simply linear. Chapter 6 surveys the effects of aqueous phase composition on the leach behavior of glasses. These studies provide a comprehensive view of the effects of both glass composition and leachant composition on leaching. The models developed correlate both durability and viscosity with glass composition. A major implication is that these findings can be used in the systematic optimization of the properties of complex oxide glasses

  5. DURABILITY TESTING OF FLUIDIZED BED STEAM REFORMER (FBSR) WASTE FORMS

    International Nuclear Information System (INIS)

    Jantzen, C

    2006-01-01

    Fluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium aqueous radioactive wastes. The addition of clay and a catalyst as co-reactants converts high sodium aqueous low activity wastes (LAW) such as those existing at the Hanford and Idaho DOE sites to a granular ''mineralized'' waste form that may be made into a monolith form if necessary. Simulant Hanford and Idaho high sodium wastes were processed in a pilot scale FBSR at Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low-activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium-bearing waste (SBW). The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The durability of the FBSR waste form products was tested in order to compare the measured durability to previous FBSR waste form testing on Hanford Envelope C waste forms that were made by THOR Treatment Technologies (TTT) and to compare the FBSR durability to vitreous LAW waste forms, specifically the Hanford low activity waste (LAW) glass known as the Low-activity Reference Material (LRM). The durability of the FBSR waste form is comparable to that of the LRM glass for the test responses studied

  6. High-level waste glass compendium; what it tells us concerning the durability of borosilicate waste glass

    International Nuclear Information System (INIS)

    Cunnane, J.C.; Allison, J.

    1993-01-01

    Facilities for vitrification of high-level nuclear waste in the United States are scheduled for startup in the next few years. It is, therefore, appropriate to examine the current scientific basis for understanding the corrosion of high-level waste borosilicate glass for the range of service conditions to which the glass products from these facilities may be exposed. To this end, a document has been prepared which compiles worldwide information on borosilicate waste glass corrosion. Based on the content of this document, the acceptability of canistered waste glass for geological disposal is addressed. Waste glass corrosion in a geologic repository may be due to groundwater and/or water vapor contact. The important processes that determine the glass corrosion kinetics under these conditions are discussed based on experimental evidence from laboratory testing. Testing data together with understanding of the long-term corrosion kinetics are used to estimate radionuclide release rates. These rates are discussed in terms of regulatory performance standards

  7. Encapsulation of TRISO particle fuel in durable soda-lime-silicate glasses

    International Nuclear Information System (INIS)

    Heath, Paul G.; Corkhill, Claire L.; Stennett, Martin C.; Hand, Russell J.; Meyer, Willem C.H.M.; Hyatt, Neil C.

    2013-01-01

    Tri-Structural Isotropic (TRISO) coated particle-fuel is a key component in designs for future high temperature nuclear reactors. This study investigated the suitability of three soda lime silicate glass compositions, for the encapsulation of simulant TRISO particle fuel. A cold press and sinter (CPS) methodology was employed to produce TRISO particle–glass composites. Composites produced were determined to have an aqueous durability, fracture toughness and Vickers’ hardness comparable to glasses currently employed for the disposal of high level nuclear wastes. Sintering at 700 °C for 30 min was found to remove all interconnected porosity from the composite bodies and oxidation of the outer pyrolytic carbon layer during sintering was prevented by processing under a 5% H 2 /N 2 atmosphere. However, the outer pyrolytic carbon layer was not effectively wetted by the encapsulating glass matrix. The aqueous durability of the TRISO particle–glass composites was investigated using PCT and MCC-1 tests combined with geochemical modelling. It was found that durability was dependent on silicate and calcium solution saturation. This study provides significant advancements in the preparation of TRISO particle encapsulant waste forms. The potential for the use of non-borosilicate sintered glass composites for TRISO particle encapsulation has been confirmed, although further refinements are required

  8. Encapsulation of TRISO particle fuel in durable soda-lime-silicate glasses

    Energy Technology Data Exchange (ETDEWEB)

    Heath, Paul G.; Corkhill, Claire L.; Stennett, Martin C.; Hand, Russell J. [Immobilisation Science Laboratory, Department of Materials Science and Engineering, Robert Hadfield Building, University of Sheffield, Sheffield S1 3JD (United Kingdom); Meyer, Willem C.H.M. [Necsa, South African Nuclear Energy Corporation, PO Box 582, Pretoria, Gauteng (South Africa); Hyatt, Neil C., E-mail: n.c.hyatt@sheffield.ac.uk [Immobilisation Science Laboratory, Department of Materials Science and Engineering, Robert Hadfield Building, University of Sheffield, Sheffield S1 3JD (United Kingdom)

    2013-05-15

    Tri-Structural Isotropic (TRISO) coated particle-fuel is a key component in designs for future high temperature nuclear reactors. This study investigated the suitability of three soda lime silicate glass compositions, for the encapsulation of simulant TRISO particle fuel. A cold press and sinter (CPS) methodology was employed to produce TRISO particle–glass composites. Composites produced were determined to have an aqueous durability, fracture toughness and Vickers’ hardness comparable to glasses currently employed for the disposal of high level nuclear wastes. Sintering at 700 °C for 30 min was found to remove all interconnected porosity from the composite bodies and oxidation of the outer pyrolytic carbon layer during sintering was prevented by processing under a 5% H{sub 2}/N{sub 2} atmosphere. However, the outer pyrolytic carbon layer was not effectively wetted by the encapsulating glass matrix. The aqueous durability of the TRISO particle–glass composites was investigated using PCT and MCC-1 tests combined with geochemical modelling. It was found that durability was dependent on silicate and calcium solution saturation. This study provides significant advancements in the preparation of TRISO particle encapsulant waste forms. The potential for the use of non-borosilicate sintered glass composites for TRISO particle encapsulation has been confirmed, although further refinements are required.

  9. Glass binder development for a glass-bonded sodalite ceramic waste form

    International Nuclear Information System (INIS)

    Riley, Brian J.; Vienna, John D.; Frank, Steven M.; Kroll, Jared O.; Peterson, Jacob A.

    2017-01-01

    This paper discusses work to develop Na_2O-B_2O_3-SiO_2 glass binders for immobilizing LiCl-KCl eutectic salt waste in a glass-bonded sodalite waste form following electrochemical reprocessing of used metallic nuclear fuel. In this paper, five new glasses with ~20 mass% Na_2O were designed to generate waste forms with high sodalite. The glasses were then used to produce ceramic waste forms with a surrogate salt waste. The waste forms made using these new glasses were formulated to generate more sodalite than those made with previous baseline glasses for this type of waste. The coefficients of thermal expansion for the glass phase in the glass-bonded sodalite waste forms made with the new binder glasses were closer to the sodalite phase in the critical temperature region near and below the glass transition temperature than previous binder glasses used. Finally, these improvements should result in lower probability of cracking in the full-scale monolithic ceramic waste form, leading to better long-term chemical durability.

  10. Remediation and production of low-sludge high-level waste glasses

    International Nuclear Information System (INIS)

    Ramsey, W.G.; Brown, K.G.; Beam, D.C.

    1994-01-01

    High-level radioactive sludge will constitute 24-28 oxide weight percent of the high-level waste glass produced at the Savannah River Site. A recent melter campaign using non-radioactive, simulated feed was performed with a sludge content considerably lower than 24 percent. The resulting glass was processed and shown to have acceptable durability. However, the durability was lower than predicted by the durability algorithm. Additional melter runs were performed to demonstrate that low sludge feed could be remediated by simply adding sludge oxides. The Product Composition Control System, a computer code developed to predict the proper feed composition for production of high-level waste glass, was utilized to determine the necessary chemical additions. The methodology used to calculate the needed feed additives, the effects of sludge oxides on glass production, and the resulting glass durability are discussed

  11. Effects of composition on waste glass properties

    International Nuclear Information System (INIS)

    Mellinger, G.B.; Chick, L.A.

    1979-01-01

    The electrical conductivity, viscosity, chemical durability, devitrification, and crystallinity of a defense waste glass were measured. Each oxide component in the glass was varied to determine its effect on these properties. A generic study is being developed which will determine the effects of 26 oxides on the above and additional properties of a wide field of possible waste glasses. 5 figures, 2 tables

  12. Reuse of ground waste glass as aggregate for mortars.

    Science.gov (United States)

    Corinaldesi, V; Gnappi, G; Moriconi, G; Montenero, A

    2005-01-01

    This work was aimed at studying the possibility of reusing waste glass from crushed containers and building demolition as aggregate for preparing mortars and concrete. At present, this kind of reuse is still not common due to the risk of alkali-silica reaction between the alkalis of cement and silica of the waste glass. This expansive reaction can cause great problems of cracking and, consequently, it can be extremely deleterious for the durability of mortar and concrete. However, data reported in the literature show that if the waste glass is finely ground, under 75mum, this effect does not occur and mortar durability is guaranteed. Therefore, in this work the possible reactivity of waste glass with the cement paste in mortars was verified, by varying the particle size of the finely ground waste glass. No reaction has been detected with particle size up to 100mum thus indicating the feasibility of the waste glass reuse as fine aggregate in mortars and concrete. In addition, waste glass seems to positively contribute to the mortar micro-structural properties resulting in an evident improvement of its mechanical performance.

  13. The effect of compositional parameters on the TCLP and PCT durability of environmental glasses

    International Nuclear Information System (INIS)

    Resce, J.L.; Overcamp, T.J.

    1995-01-01

    The relationship between glass composition and the chemical durability of environmental waste glass is very important for both the development of glass formulations and the prediction of glass durability for process control. The development of such a model is extremely difficult for several reasons. Firstly, chemical durability is dependent upon the type of leach test employed; the leach tests themselves being only crude approximations of actual environmental conditions or long term behavior. Secondly, devitrification or crystallinity can also play a major role in durability, but is much more difficult to quantify. Lastly, the development of any one model for all glass types is impractical because of the wide variety of wastestreams, the heterogeneity of the wastestreams, and the large variety of components within each wastestream. Several ongoing efforts have been directed toward this goal, but as yet, no model has been proven acceptable

  14. Characterization of Savannah River Plant waste glass

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1985-01-01

    The objective of the glass characterization programs at the Savannah River Laboratory (SRL) is to ensure that glass containing Savannah River Plant high-level waste can be permanently stored in a federal repository, in an environmentally acceptable manner. To accomplish this objective, SRL is carrying out several experimental programs, including: fundamental studies of the reactions between waste glass and water, particularly repository groundwater; experiments in which candidate repository environments are simulated as accurately as possible; burial tests of simulated waste glass in candidate repository geologies; large-scale tests of glass durability; and determination of the effects of process conditions on glass quality. In this paper, the strategy and current status of each of these programs is discussed. The results indicate that waste packages containing SRP waste glass will satisfy emerging regulatory criteria

  15. Operating Range for High Temperature Borosilicate Waste Glasses: (Simulated Hanford Enveloped)

    International Nuclear Information System (INIS)

    Mohammad, J.; Ramsey, W. G.; Toghiani, R. K.

    2003-01-01

    The following results are a part of an independent thesis study conducted at Diagnostic Instrumentation and Analysis Laboratory-Mississippi State University. A series of small-scale borosilicate glass melts from high-level waste simulant were produced with waste loadings ranging from 20% to 55% (by mass). Crushed glass was allowed to react in an aqueous environment under static conditions for 7 days. The data obtained from the chemical analysis of the leachate solutions were used to test the durability of the resulting glasses. Studies were performed to determine the qualitative effects of increasing the B2O3 content on the overall waste glass leaching behavior. Structural changes in a glass arising due to B2O3 were detected indirectly by its chemical durability, which is a strong function of composition and structure. Modeling was performed to predict glass durability quantitatively in an aqueous environment as a direct function of oxide composition

  16. Nuclear waste immobilization in iron phosphate glasses

    International Nuclear Information System (INIS)

    Garcia, D.A.; Rodriguez, Diego A.; Menghini, Jorge E.; Bevilacqua, Arturo

    2007-01-01

    Iron-phosphate glasses have become important in the nuclear waste immobilization area because they have some advantages over silicate-based glasses, such as a lower processing temperature and a higher nuclear waste load without losing chemical and mechanical properties. Structure and chemical properties of iron-phosphate glasses are determined in terms of the main components, in this case, phosphate oxide along with the other oxides that are added to improve some of the characteristics of the glasses. For example, Iron oxide improves chemical durability, lead oxide lowers fusion temperature and sodium oxide reduces viscosity at high temperature. In this work a study based on the composition-property relations was made. We used different techniques to characterize a series of iron-lead-phosphate glasses with uranium and aluminium oxide as simulated nuclear waste. We used the Arquimedes method to determine the bulk density, differential temperature analysis (DTA) to determine both glass transition temperature and crystallization temperature, dilatometric analysis to calculate the linear thermal expansion coefficient, chemical durability (MCC-1 test) and X-ray diffraction (XRD). We also applied some theoretic models to calculate activation energies associated with the glass transition temperature and crystallization processes. (author)

  17. Chemical durability of simulated nuclear glasses containing water

    International Nuclear Information System (INIS)

    Li, H.; Tomozawa, M.

    1995-04-01

    The chemical durability of simulated nuclear waste glasses having different water contents was studied. Results from the product consistency test (PCT) showed that glass dissolution increased with water content in the glass. This trend was not observed during MCC-1 testing. This difference was attributed to the differences in reactions between glass and water. In the PCT, the glass network dissolution controlled the elemental releases, and water in the glass accelerated the reaction rate. On the other hand, alkali ion exchange with hydronium played an important role in the MCC-1. For the latter, the amount of water introduced into a leached layer from ion-exchange was found to be much greater than that of initially incorporated water in the glass. Hence, the initial water content has no effect on glass dissolution as measured by the MCC-1 test

  18. Final Report for Crucible -Scale Radioactive Vitrification and Product Test of Waste Envelope B (AZ-102) Low-Activity Waste Glass

    International Nuclear Information System (INIS)

    CRAWFORD, CHARLES

    2004-01-01

    A proof-of-technology demonstration for the Hanford River Protection Project (RPP) Waste treatment and Immobilization Plant (WTP) was performed by the Savannah River Technology Center (SRTC). As part of this demonstration, treated AZ-102 Low-Activity Waste supernate was vitrified using a crucible-scale furnace. Initial glass samples were quench-cooled and characterized for metals and radionuclides. The glass was also durability tested using the American Society for Testing and Materials (ASTM) Product Consistency Test (PCT) protocol. These tests used the AZ-102 glass formulation Low Activity Waste (LAW) B88 that targeted AZ-102 waste loading at 5 wt% Na2O. After these initial results were obtained with the quench-cooled LAWB88 glass, a prototypical container centerline cooling (CCC) program was supplied to SRTC by WTP. A portion of the quench-cooled LAWB88 glass was remelted and centerline cooled. Samples from the CCC low-activity AZ-102 glass waste form were durability tested using the PCT and characterized for crystalline phase identification.This final report documents the characterization and durability of this AZ-102 glass

  19. Am/Cm target glass durability dependence on pH (U). Revision 1

    International Nuclear Information System (INIS)

    Daniel, W.E.; Best, D.R.

    1996-03-01

    At the Westinghouse Savannah River Company near Aiken, South Carolina, a process is being developed to safely vitrify all of the highly radioactive americium/curium (Am/Cm) material and a portion of the other fissile actinide materials stored on site. One goal of this campaign is to provide Oak Ridge National Laboratory with the excess Am/Cm so it can be recycled as opposed to simply disposing of it as waste. The vitrification will allow safe transportation of the Am/Cm to Oak Ridge as well as safe storage once it arrives. The Am/Cm Target glass being used in this project has been specifically designed to be extremely durable in aqueous environments while it can be selectively attacked by nitric acid to recover the valuable Am and Cm isotopes. Similar glass compositions could be used for storage and retrieval of other actinides on the WSRC site. Previous reports have presented the time, temperature, and compositional dependence of the Am/Cm glass durability. This paper will show results from a pH study on the Am/Cm Target glass durability. The data indicate that the Am/Cm Target Glass durability decreases as pH decreases from a neutral reading. These findings support the extraction of the valuable isotopes from the glass using nitric acid

  20. Time-temperature-transformation kinetics in SRL waste glass

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Bickford, D.F.; Karraker, D.G.

    1983-01-01

    Time-temperature-transformation (TTT) curves have been determined for SRL 165 waste glass. Extent and sequence of crystallization were determined by XRD and SEM. The incipient crystallization product, spinel, can be determined at one volume percent by magnetic susceptibility. The type and percentage of crystallization is correlated with waste glass durability. 20 references, 5 figures, 1 table

  1. Evaluation of lead-iron-phosphate glass as a high-level waste form

    International Nuclear Information System (INIS)

    Chick, L.A.; Bunnell, L.R.; Strachan, D.M.; Kissinger, H.E.; Hodges, F.N.

    1986-01-01

    The lead-iron-phosphate (Pb-Fe-P) nuclear waste glass developed at Oak Ridge National Laboratory (ORNL) was evaluated for its potential as an improvement over the current reference waste form, borosilicate (B-Si) glass. Vitreous Pb-Fe-P glass appears to have substantially better chemical durability than B-Si glass. However, severe crystallization leading to deteriorated chemical durability would result if this glass were poured into large canisters, as is presently done with B-Si glass. Cesium leach rates from this crystallized material are orders of magnitude greater than those from B-Si glass. Therefore, to realize the performance advantages of the Pb-Fe-P material in a nuclear waste form, it would be necessary to process it so that it is cooled rapidly, thus retaining its vitreous structure

  2. Evaluation of lead-iron-phosphate glass as a high-level waste form

    International Nuclear Information System (INIS)

    Chick, L.A.; Bunnell, L.R.; Strachan, D.M.; Kissinger, H.E.; Hodges, F.N.

    1986-01-01

    The lead-iron-phosphate nuclear waste glass developed at Oak Ridge National Laboratory (ORNL) was evaluated for its potential as an improvement over the current reference waste form, borosilicate glass. Vitreous lead-iron-phosphate glass appears to have substantially better chemical durability than borosilicate glass. However, severe crystallization leading to deteriorated chemical durability would result if this glass were poured into large canisters as is presently done with borosilicate glass. Cesium leach rates from this crystallized material are orders of magnitude greater than those from borosilicate glass. Therefore, in order to realize the performance advantages of the lead-iron-phosphate material in a nuclear waste form, it would be necessary to process it so that it is rapidly cooled, thus retaining its vitreous structure. 22 refs., 4 figs., 4 tabs

  3. Control of high-level radioactive waste-glass melters

    International Nuclear Information System (INIS)

    Bickford, D.F.; Coleman, C.J.

    1990-01-01

    The Defense Waste Processing Facility (DWPF) will immobilize Savannah River Site High Level Waste as a durable borosilicate glass for permanent disposal in a repository. The DWPF will be controlled based on glass composition. The following discussion is a preliminary analysis of the capability of the laboratory methods that can be used to control the glass composition, and the relationships between glass durability and glass properties important to glass melting. The glass durability and processing properties will be controlled by controlling the chemical composition of the glass. The glass composition will be controlled by control of the melter feed transferred from the Slurry Mix Evaporator (SME) to the Melter Feed Tank (MFT). During cold runs, tests will be conducted to demonstrate the chemical equivalence of glass sampled from the pour stream and glass removed from cooled canisters. In similar tests, the compositions of glass produced from slurries sampled from the SME and MFT will be compared to final product glass to determine the statistical relationships between melter feed and glass product. The total error is the combination of those associated with homogeneity in the SME or MFT, sampling, preparation of samples for analysis, instrument calibration, analysis, and the composition/property model. This study investigated the sensitivity of estimation of property data to the combination of variations from sampling through analysis. In this or a similar manner, the need for routine glass product sampling will be minimized, and glass product characteristics will be assured before the melter feed is committed to the melter

  4. Influence of the leaching mode on the durability of a glass for fission product containment

    International Nuclear Information System (INIS)

    Nogues, J.L.; Terki, A.

    1984-06-01

    The chemical durability of a glass containing wastes from light water reactor (LWR) has been studied with three different lixiviation modes: ''static'' leach test, ''soxhlet'' test and ''continuous flow'' test. After a description of these tests, the leaching mode influence on the glass durability is reported as obtained from weight loss measurements, analyses of the leaching solutions and surface analyses of the samples. Finally, the corrosion mechanisms of this type of glass are approached and a phenomenological explanation of attack from an aqueous solution is proposed [fr

  5. Predicting liquid immiscibility in multicomponent nuclear waste glasses

    International Nuclear Information System (INIS)

    Peeler, D.K.; Hrma, P.R.

    1994-04-01

    Taylor's model for predicting amorphous phase separation in complex, multicomponent systems has been applied to high-level (simulated) radioactive waste glasses at the US Department of Energy's Hanford site. Taylor's model is primarily based on additions of modifying cations to a Na 2 O-B 2 O 3 -SiO 2 (NBS) submixture of the multicomponent glass. The position of the submixture relative to the miscibility dome defines the development probability of amorphous phase separation. Although prediction of amorphous phase separation in Hanford glasses (via experimental SEM/TEM analysis) is the primary thrust of this work; reported durability data is also provides limited insight into the composition/durability relationship. Using a modified model similar to Taylor's, the results indicate that immiscibility may be predicted for multicomponent waste glasses by the addition of Li 2 O to the ''alkali'' corner of the NBS submixture

  6. Nuclear waste glass product consistency test (PCT), Version 5.0

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Ramsey, W.G.; Waters, B.J.

    1992-06-01

    Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced in the Defense Waste Processing Facility (DWPF), poured into stainless steel canisters, and eventually disposed of in a geologic repository. In order to comply with the Waste Acceptance Preliminary Specifications (WAPS), the durability of the glass needs to be measured during production to assure its long term stability and radionuclide release properties. A durability test, designated the Produce Consistency Test (PCT), was developed for DWPF glass in order to meet the WAPS requirements. The response of the PCT procedure was based on extensive testing with glasses of widely different compositions. The PCT was determined to be very reproducible, to yield reliable results rapidly, and to be easily performed in shielded cell facilities with radioactive samples. Version 5.0 of the PCT procedure is attached

  7. Surface chemistry and durability of borosilicate glass

    International Nuclear Information System (INIS)

    Carroll, S.A.; Bourcier, W.L.; Phillips, B.L.

    1994-01-01

    Important glass-water interactions are poorly understood for borosilicate glass radioactive waste forms. Preliminary results show that glass durability is dependent on reactions occurring at the glass-solution interface. CSG glass (18.2 wt. % Na 2 O, 5.97 wt. % CaO, 11.68 wt. % Al 2 O 3 , 8.43 wt. % B 2 O 3 , and 55.73 wt. % SiO 2 ) dissolution and net surface H + and OH - adsorption are minimal at near neutral pH. In the acid and alkaline pH regions, CSG glass dissolution rates are proportional to [H + ] adsorbed 2 and [OH - ] adsorbed 0.8 , respectively. In contrast, silica gel dissolution and net H + and OH - adsorption are minimal and independent of pH in acid to neutral solutions. In the alkaline pH region, silica gel dissolution is proportional to [OH - ] adsorbed 0.9 adsorbed . Although Na adsorption is significant for CSG glass and silica gel in the alkaline pH regions, it is not clear if it enhances dissolution, or is an artifact of depolymerization of the framework bonds

  8. SETTLING OF SPINEL IN A HIGH-LEVEL WASTE GLASS MELTER

    International Nuclear Information System (INIS)

    Pavel Hrma; Pert Schill; Lubomir Nemec

    2002-01-01

    High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors called melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150 decrees C to create a melt that becomes glass on cooling

  9. Defense Waste Processing Facility (DWPF) Durability-Composition Models and the Applicability of the Associated Reduction of Constraints (ROC) Criteria for High TiO2 Containing Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Trivelpiece, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-30

    Radioactive high-level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the DWPF since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it has been poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than relying on statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to determine, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository. The DWPF SPC system is known as the Product Composition Control System (PCCS). One of the process models within PCCS is known as the Thermodynamic Hydration Energy Reaction MOdel (THERMO™). The DWPF will soon be receiving increased concentrations of TiO2-, Na2O-, and Cs2O-enriched wastes from the Salt Waste Processing Facility (SWPF). The SWPF has been built to pretreat the high-curie fraction of the salt waste to be removed from the HLW tanks in the F- and H-Area Tank Farms at the SRS. In order to validate the existing TiO2 term in THERMO™ beyond 2.0 wt% in the DWPF, new durability data were developed over the target range of 2.00 to 6.00 wt% TiO2 and evaluated against the 1995 durability model. The durability was measured by the 7-day Product Consistency Test. This study documents the adequacy of the existing THERMO™ terms. It is recommended that the modified THERMO™ durability models and

  10. Investigation of lead-iron-phosphate glass for SRP waste

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1986-10-01

    The search for a host solid for the immobilization of nuclear waste has focused on various vitreous waste forms. Recently, lead-iron-phosphate (LIP) glasses have been proposed for solidification of all types of HLLW. Investigation of this glass for vitrification of SRP waste demonstrated that the phosphate glass is incompatible with the current borosilicate glass technology. The durability of LIP glasses in deionized water was comparable to current borosilicate waste glass formulations, and the LIP glass has a low melt temperature. However, many of the defense waste constituents have low solubility in the phosphate melt, producing an inhomogeneous product. Also, the LIP melt is highly corrosive which prevents the use of current melter materials, in particular Inconel 690, and thus requires more exotic materials of construction such as platinum

  11. Characterization of the Defense Waste Processing Facility (DWPF) Environmental Assessment (EA) glass Standard Reference Material. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Crawford, C.L.; Pickett, M.A.

    1993-06-01

    Liquid high-level nuclear waste at the Savannah River Site (SRS) will be immobilized by vitrification in borosilicate glass. The glass will be produced and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). Other waste form producers, such as West Valley Nuclear Services (WVNS) and the Hanford Waste Vitrification Project (HWVP), will also immobilize high-level radioactive waste in borosilicate glass. The canistered waste will be stored temporarily at each facility for eventual permanent disposal in a geologic repository. The Department of Energy has defined a set of requirements for the canistered waste forms, the Waste Acceptance Product Specifications (WAPS). The current Waste Acceptance Primary Specification (WAPS) 1.3, the product consistency specification, requires the waste form producers to demonstrate control of the consistency of the final waste form using a crushed glass durability test, the Product Consistency Test (PCI). In order to be acceptable, a waste glass must be more durable during PCT analysis than the waste glass identified in the DWPF Environmental Assessment (EA). In order to supply all the waste form producers with the same standard benchmark glass, 1000 pounds of the EA glass was fabricated. The chemical analyses and characterization of the benchmark EA glass are reported. This material is now available to act as a durability and/or redox Standard Reference Material (SRM) for all waste form producers.

  12. Effect of different glasses in glass bonded zeolite

    International Nuclear Information System (INIS)

    Lewis, M.A.; Ackerman, J.P.; Verma, S.

    1995-01-01

    A mineral waste form has been developed for chloride waste salt generated during the pyrochemical treatment of spent nuclear fuel. The waste form consists of salt-occluded zeolite powders bound within a glass matrix. The zeolite contains the salt and immobilizes the fission products. The zeolite powders are hot pressed to form a mechanically stable, durable glass bonded zeolite. Further development of glass bonded zeolite as a waste form requires an understanding of the interaction between the glass and the zeolite. Properties of the glass that enhance binding and durability of the glass bonded zeolite need to be identified. Three types of glass, boroaluminosilicate, soda-lime silicate, and high silica glasses, have a range of properties and are now being investigated. Each glass was hot pressed by itself and with an equal amount of zeolite. MCC-1 leach tests were run on both. Soda-lime silicate and high silica glasses did not give a durable glass bonded zeolite. Boroaluminosilicate glasses rich in alkaline earths did bind the zeolite and gave a durable glass bonded zeolite. Scanning electron micrographs suggest that the boroaluminosilicate glasses wetted the zeolite powders better than the other glasses. Development of the glass bonded zeolite as a waste form for chloride waste salt is continuing

  13. Predicting liquid immiscibility in multicomponent nuclear waste glasses

    International Nuclear Information System (INIS)

    Peeler, D.K.; Hrma, P.R.

    1994-01-01

    Taylor's model for predicting amorphous phase separation in complex, multicomponent systems has been applied to high-level (simulated) radioactive waste glasses at the U.S. Department of Energy's Hanford site. Taylor's model is primarily based on additions of modifying cations to a Na 2 O-B 2 O 3 -SiO 2 (NBS) submixture of the multicomponent glass. The position of the submixture relative to the immiscibility dome defines the development probability of amorphous phase separation. Although prediction of amorphous phase separation in Hanford glasses (via experimental SEM/TEM analysis) is the primary thrust of this work; reported durability data is also provides limited insight into the composition/durability relationship. Using a modified model similar to Taylor's, the results indicate that immiscibility may be predicted for multicomponent waste glasses by the addition of Li 2 O to the open-quotes alkaliclose quotes corner of the NBS submixture

  14. Chemically durable iron phosphate glasses for vitrifying sodium bearing waste (SBW) using conventional and cold crucible induction melting (CCIM) techniques

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C.W. E-mail: cheol@umr.edu; Ray, C.S.; Zhu, D.; Day, D.E.; Gombert, D.; Aloy, A.; Mogus-Milankovic, A.; Karabulut, M

    2003-11-01

    A simulated sodium bearing waste (SBW) was successfully vitrified in iron phosphate glasses (IPG) at a maximum waste loading of 40 wt% using conventional and cold crucible induction melting (CCIM) techniques. No sulfate segregation or crystalline phases were detectable in the IPG when examined by SEM and XRD. The IPG wasteforms containing 40 wt% SBW satisfy current DOE requirements for aqueous chemical durability as evaluated from their bulk dissolution rate (D{sub R}), product consistency test, and vapor hydration test. The fluid IPG wasteforms can be melted at a relatively low temperature (1000 deg. C) and for short times (<6 h). These properties combined with a significantly higher waste loading, and the feasibility of CCIM melting offer considerable savings in time, energy, and cost for vitrifying the SBW stored at the Idaho National Engineering and Environmental Laboratory in iron phosphate glasses.

  15. MINERALIZATION OF RADIOACTIVE WASTES BY FLUIDIZED BED STEAM REFORMING (FBSR): COMPARISONS TO VITREOUS WASTE FORMS, AND PERTINENT DURABILITY TESTING

    International Nuclear Information System (INIS)

    Jantzen, C.

    2008-01-01

    The Savannah River National Laboratory (SRNL) was requested to generate a document for the Washington State Department of Ecology and the U.S. Environmental Protection Agency that would cover the following topics: (1) A description of the mineral structures produced by Fluidized Bed Steam Reforming (FBSR) of Hanford type Low Activity Waste (LAW including LAWR which is LAW melter recycle waste) waste, especially the cage structured minerals and how they are formed. (2) How the cage structured minerals contain some contaminants, while others become part of the mineral structure (Note that all contaminants become part of the mineral structure and this will be described in the subsequent sections of this report). (3) Possible contaminant release mechanisms from the mineral structures. (4) Appropriate analyses to evaluate these release mechanisms. (5) Why the appropriate analyses are comparable to the existing Hanford glass dataset. In order to discuss the mineral structures and how they bond contaminants a brief description of the structures of both mineral (ceramic) and vitreous waste forms will be given to show their similarities. By demonstrating the similarities of mineral and vitreous waste forms on atomic level, the contaminant release mechanisms of the crystalline (mineral) and amorphous (glass) waste forms can be compared. This will then logically lead to the discussion of why many of the analyses used to evaluate vitreous waste forms and glass-ceramics (also known as glass composite materials) are appropriate for determining the release mechanisms of LAW/LAWR mineral waste forms and how the durability data on LAW/LAWR mineral waste forms relate to the durability data for LAW/LAWR glasses. The text will discuss the LAW mineral waste form made by FBSR. The nanoscale mechanism by which the minerals form will be also be described in the text. The appropriate analyses to evaluate contaminant release mechanisms will be discussed, as will the FBSR test results to

  16. Impact of crystallization on the structure and chemical durability of borosilicate glass

    International Nuclear Information System (INIS)

    Nicoleau, Elodie

    2016-01-01

    This work describes a new approach to help understand the chemical durability of partially crystallized nuclear waste conditioning matrices. Among the studies carried out on nuclear waste deep geological disposal, long term behavior studies have so far been conducted on homogeneous glassy matrices. However, as the crystalline phases may generate modifications in the chemical composition and properties of such matrices, the description and a better understanding of their effects on the chemical durability of waste packages are of primary importance. A protocol to study the durability of heterogeneous model matrices of nuclear interest containing different types of crystalline phases was developed. It is based on a detailed description of the morphology, microstructure and structure of the glassy matrix and crystalline phases, and on the study of various alteration regimes. Three crystal phases that may form when higher concentrations of waste are immobilized in Uranium Oxide type conditioning glasses were studied: alkali and alkaline earth molybdates, rare earth silicates and ruthenium oxide. The results highlight the roles of the composition and the structure of the surrounding glassy matrix as the parameters piloting the alteration kinetics of the partially crystallized glassy matrices. This behavior is identical whatever the nature of the crystalline phases, as long as these phases do not lead to a composition gradient and do not percolate within the glassy matrix. Given these results, a methodology to study partially crystallized matrices with no composition gradient is then suggested. Its key development lies firstly in the evaluation of the behavior of partially crystallized matrices through the experimental study of the residual glassy matrix in various alteration regimes. This methodology may be adapted to the case of new glass formulations with more complex compositions (e.g. highly waste-loaded glass), which may contain crystals formed during cooling

  17. Evaluation of lead-iron-phosphate glass as a high-level waste form

    International Nuclear Information System (INIS)

    Chick, L.A.; Bunnell, L.R.; Strachan, D.M.; Kissinger, H.E.; Hodges, F.N.

    1986-09-01

    The lead-iron-phosphate (Pb-Fe-P) glass developed at Oak Ridge National Laboratory was evaluated for its potential as an improvement over the current reference nuclear waste form, borosilicate (B-Si) glass. The evaluation was conducted as part of the Second Generation HLW Technology Subtask of the Nuclear Waste Treatment Program at Pacific Northwest Laboratory. The purpose of this work was to investigate possible alternatives to B-Si glass as second-generation waste forms. While vitreous Pb-Fe-P glass appears to have substantially better chemical durability than B-Si glass, severe crystallization or devitrification leading to deteriorated chemical durability would result if this glass were poured into large canisters as is the procedure with B-Si glass. Cesium leach rates from this crystallized material are orders of magnitude greater than those from B-Si glass. Therefore, to realize the potential performance advantages of the Pb-Fe-P material in a nuclear waste form, the processing method would have to cool the material rapidly to retain its vitreous structure

  18. Test plan: Effects of phase separation on waste loading for high level waste glasses

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    2000-01-01

    As part of the Tanks Focus Area's (TFA) effort to increase waste loading for high-level waste (HLW) vitrification at various facilities in the Department of Energy (DOE) complex, the occurrence of phase separation in waste glasses spanning the Savannah River Site (SRS) and Idaho National Engineering and Environmental Laboratory (INEEL) composition ranges were studied during FY99. The type, extent, and impact of phase separation on glass durability for a series of HLW glasses, e.g., SRS-type and INEEL-type, were examined

  19. Temperature effects on waste glass performance

    International Nuclear Information System (INIS)

    Mazer, J.J.

    1991-02-01

    The temperature dependence of glass durability, particularly that of nuclear waste glasses, is assessed by reviewing past studies. The reaction mechanism for glass dissolution in water is complex and involves multiple simultaneous reaction proceeded, including molecular water diffusion, ion exchange, surface reaction, and precipitation. These processes can change in relative importance or dominance with time or changes in temperature. The temperature dependence of each reaction process has been shown to follow an Arrhenius relationship in studies where the reaction process has been isolated, but the overall temperature dependence for nuclear waste glass reaction mechanisms is less well understood, Nuclear waste glass studies have often neglected to identify and characterize the reaction mechanism because of difficulties in performing microanalyses; thus, it is unclear if such results can be extrapolated to other temperatures or reaction times. Recent developments in analytical capabilities suggest that investigations of nuclear waste glass reactions with water can lead to better understandings of their reaction mechanisms and their temperature dependences. Until a better understanding of glass reaction mechanisms is available, caution should be exercised in using temperature as an accelerating parameter. 76 refs., 1 tab

  20. A Revised Method For Estimating Oxide Basicity Per The Smith Scale With Example Application To Glass Durability

    International Nuclear Information System (INIS)

    Reynolds, J.G.

    2011-01-01

    Previous researchers have developed correlations between oxide electronegativity and oxide basicity. The present paper revises those correlations using a newer method of calculating electronegativity of the oxygen anion. Basicity is expressed using the Smith α parameter scale. A linear relation was found between the oxide electronegativity and the Smith α parameter, with an R 2 of 0.92. An example application of this new correlation to the durability of high-level nuclear waste glass is demonstrated. The durability of waste glass was found to be directly proportional to the quantity and basicity of the oxides of tetrahedrally coordinated network forming ions.

  1. Nanoporous Glasses for Nuclear Waste Containment

    OpenAIRE

    Woignier, Thierry; Primera, Juan; Reynes, Jerôme

    2016-01-01

    Research is in progress to incorporate nuclear waste in new matrices with high structural stability, resistance to thermal shock, and high chemical durability. Interactions with water are important for materials used as a containment matrix for the radio nuclides. It is indispensable to improve their chemical durability to limit the possible release of radioactive chemical species, if the glass structure is attacked by corrosion. By associating high structural stability and high chemical dura...

  2. Durability of Mortar Made with Fine Glass Powdered Particles

    Directory of Open Access Journals (Sweden)

    Rosemary Bom Conselho Sales

    2017-01-01

    Full Text Available Different studies investigate the use of waste glass in Portland cement compounds, either as aggregates or as supplementary cementitious materials. Nevertheless, it seems that there is no consensus about the influence of particle color and size on the behavior of the compounds. This study addresses the influence of cement replacement by 10 and 20% of the colorless and amber soda-lime glass particles sized around 9.5 μm on the performance of Portland cement mortars. Results revealed that the partial replacement of cement could contribute to the production of durable mortars in relation to the inhibition of the alkali-aggregate reaction. This effect was more marked with 20% replacement using amber glass. Samples containing glass microparticles were more resistant to corrosion, in particular those made of colorless glass. The use of colorless and amber glass microparticles promoted a reduction in wear resistance.

  3. Recent results on the effect of gamma radiation on the durability and microstructure of DWPF glass

    International Nuclear Information System (INIS)

    Bibler, N.E.; Tosten, M.H.; Beam, D.C.

    1989-01-01

    The effect of gamma radiation on the durability and microstructure of a simulated nuclear waste glass from the Savannah River Site has been carefully investigated. Three large pieces of glass were irradiated with a Co-60 source to three doses up to a maximum dose of 3.1 x 10 10 rad. Internal samples of the large pieces of irradiated and unirradiated glass were leached in deionized water to investigate durability changes and were examined by transmission electron microscopy (TEM) to investigate microstructure changes. Leach tests were performed in triplicate at 90 degree C with crushed glass samples in deionized water. A statistical analysis of the results indicated to the 95% confidence level that the radiation did not affect the glass durability. Careful examination by TEM indicated no effect of gamma radiation on the microstructure of the glass although severe damage could be induced by the electron beam from the microscope. 19 refs., 2 figs., 3 tabs

  4. Minor component study for simulated high-level nuclear waste glasses (Draft)

    International Nuclear Information System (INIS)

    Li, H.; Langowskim, M.H.; Hrma, P.R.; Schweiger, M.J.; Vienna, J.D.; Smith, D.E.

    1996-02-01

    Hanford Site single-shell tank (SSI) and double-shell tank (DSI) wastes are planned to be separated into low activity (or low-level waste, LLW) and high activity (or high-level waste, HLW) fractions, and to be vitrified for disposal. Formulation of HLW glass must comply with glass processibility and durability requirements, including constraints on melt viscosity, electrical conductivity, liquidus temperature, tendency for phase segregation on the molten glass surface, and chemical durability of the final waste form. A wide variety of HLW compositions are expected to be vitrified. In addition these wastes will likely vary in composition from current estimates. High concentrations of certain troublesome components, such as sulfate, phosphate, and chrome, raise concerns about their potential hinderance to the waste vitrification process. For example, phosphate segregation in the cold cap (the layer of feed on top of the glass melt) in a Joule-heated melter may inhibit the melting process (Bunnell, 1988). This has been reported during a pilot-scale ceramic melter run, PSCM-19, (Perez, 1985). Molten salt segregation of either sulfate or chromate is also hazardous to the waste vitrification process. Excessive (Cr, Fe, Mn, Ni) spinel crystal formation in molten glass can also be detrimental to melter operation

  5. Immobilization of hazardous and radioactive waste into glass structures

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1997-01-01

    As a result of more than three decades of international research, glass has emerged as the material of choice for immobilization of a wide range of potentially hazardous radioactive and non-radioactive materials. The ability of glass structures to incorporate and then immobilize many different elements into durable, high integrity, waste glass products is a direct function of the unique random network structure of the glassy state. Every major country involved with long-term management of high-level radioactive waste (HLW) has either selected or is considering glass as the matrix of choice for immobilizing and ultimately, disposing of the potentially hazardous, high-level radioactive material. There are many reasons why glass is preferred. Among the most important considerations are the ability of glass structures to accommodate and immobilize the many different types of radionuclides present in HLW, and to produce a product that not only has excellent technical properties, but also possesses good processing features. Good processability allows the glass to be fabricated with relative ease even under difficult remote-handling conditions necessary for vitrification of highly radioactive material. The single most important property of the waste glass produced is its ability to retain hazardous species within the glass structure and this is reflected by its excellent chemical durability and corrosion resistance to a wide range of environmental conditions. In addition to immobilization of HLW glass matrices are also being considered for isolation of many other types of hazardous materials, both radioactive as well as nonradioactive. This includes vitrification of various actinides resulting from clean-up operations and the legacy of the cold war, as well as possible immobilization of weapons grade plutonium resulting from disarmament activities. Other types of wastes being considered for immobilization into glasses include transuranic wastes, mixed wastes, contaminated

  6. Nuclear waste glass product consistency test (PCT): Version 7.0. Revision 3

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Ramsey, W.G.

    1994-06-01

    Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced in the Defense Waste Processing Facility (DWPF), poured into stainless steel canisters, and eventually disposed of in a geologic repository. In order to comply with the Waste Acceptance Product Specifications (WAPS), the durability of the glass needs to be measured during production to assure its long term stability and radionuclide release properties. A durability test, designated the Product Consistency Test (PCT), was developed for DWPF glass in order to meet the WAPS requirements. The response of the PCT procedure was based on extensive testing with glasses of widely different compositions. The PCT was determined to be very reproducible, to yield reliable results rapidly, and to be easily performed in shielded cell facilities with radioactive samples. Version 7.0 of the PCT procedure is attached. This draft version has been submitted to ASTM for full committee (C26, Nuclear Fuel Cycle) ballot after being balloted successfully through subcommittee C26.13 on Repository Waste Package Materials Testing

  7. Production and remediation of low-sludge, simulated Purex waste glasses, 1: Effects of sludge oxide additions on melter operation

    International Nuclear Information System (INIS)

    Ramsey, W.G.

    1993-01-01

    Glass produced during the Purex 4 campaigns of the Integrated Defense Waste Processing Facility (DWPF) Melter System (IDMS) and the 774 Research Melter contained a lower fraction of sludge components than targeted by the Product Composition Control System (PCCS). Purex 4 glass was more durable than the benchmark (EA) glass, but less durable than most simulated SRS high-level waste glasses. Also, Purex 4 glass was considerably less durable than predicted by the algorithm which will be used to control production of DWPF glass. A melter run was performed using the 774 Research Melter to determine if the initial PCCS target composition determined for Purex 4 would produce acceptable glass whose durability could be accurately modeled by Hydration Thermodynamics. Reagent grade oxides and carbonates were added to Purex 4 melter feed stock to simulate a higher sludge loading. Each canister of glass produced was sampled and the composition, crystallinity, and durability was determined. This document details the melter operation and composition and crystallinity analyses

  8. Technical Note: Updated durability/composition relationships for Hanford high-level waste glasses

    International Nuclear Information System (INIS)

    Piepel, G.F.; Hartley, S.A.; Redgate, P.E.

    1996-03-01

    This technical note presents empirical models developed in FYI 995 to predict durability as functions of glass composition. Models are presented for normalized releases of B, Li, Na, and Si from the 7-day Product Consistency Test (PCT) applied to quenched and canister centerline cooled (CCC) glasses as well as from the 28-day Materials Characterization Center-1 (MCC-1) test applied to quenched glasses. Models are presented for Composition Variation Study (CVS) data from low temperature melter (LTM) studies (Hrma, Piepel, et al. 1994) and high temperature melter (HTM) studies (Vienna et al. 1995). The data used for modeling in this technical note are listed in Appendix A

  9. Low leach rate glasses for immobilization of nuclear wastes

    International Nuclear Information System (INIS)

    Chick, L.A.; Buckwalter, C.Q.

    1980-10-01

    Improved defense and commercial waste glass have about one order of magnitude lower leach rates at 90 0 C in static deionized water than reference glasses. This durability difference diminishes as the leaching temperature is raised, but at repository temperature less than 150 0 C, the improved compositions would have considerable advantages over reference glases. At the melting temperatures necessary for most of the high-durability glasses, volatility was found to be higher than that experienced in processing current reference glases. Higher volatilities might be compensated for by specific design of the off-gas system for improved off-gas treatment and volatile materials recovery. 6 figures, 2 tables

  10. Glassy slag: A complementary waste form to homogeneous glass for the implementation of MAWS in treating DOE low level/mixed wastes

    International Nuclear Information System (INIS)

    Feng, X.; Ordaz, G.; Krumrine, P.

    1994-01-01

    Glassy slag waste forms are being developed to complement glass waste forms in implementing the Minimum Additive Waste Stabilization (MAWS) Program for supporting DOE's environmental restoration efforts. These glassy slags are composed of various metal oxide crystalline phases embedded in an alumino-silicate glass phase. The slags are appropriate final waste forms for waste streams that contain large amounts of scrap metals and elements with low solubilities in glass, and that have low-flux contents. Homogeneous glass waste forms are appropriate for wastes with sufficient fluxes and low metal contents. Therefore, utilization of both glass and glassy slag waste forms will make vitrification technology applicable to the treatment of a much larger range of radioactive and mixed wastes. The MAWS approach was a plied to glassy slags by blending multiple waste streams to produce the final waste form, minimizing overall waste form volume and reducing costs. The crystalline oxide phases formed in the glassy slags can be specially formulated so that they are very durable and contain hazardous and radioactive elements in their lattice structures. The Structural Bond Strength (SBS) Model was used to predict the chemical durability of the product from the slag composition so that optimized slag compositions could be obtain with a limited number of crucible melts and testing

  11. Chemical durability and characterization of nuclear waste forms in a hydrothermal environment

    International Nuclear Information System (INIS)

    Braithwaite, J.W.; Johnstone, J.K.

    1979-01-01

    The chemical durability of a simulated copper borosilicate waste glass and titanate waste ceramic has been studied in hydrothermal environments which could possibly be encountered in a bedded salt or sub-sealed waste isolation repository. The major parameters investigated which affect matrix corrosion and cesium solubilization include solution saturation and equilibrium phenomena, solution composition (especially the Mg +2 ion concentration), pH, particle size, temperature, and time

  12. Comparison of the corrosion behaviors of the glass-bonded sodalite ceramic waste form and reference HLW glasses

    International Nuclear Information System (INIS)

    Ebert, W. L.; Lewis, M. A.

    1999-01-01

    A glass-bonded sodalite ceramic waste form is being developed for the long-term immobilization of salt wastes that are generated during spent nuclear fuel conditioning activities. A durable waste form is prepared by hot isostatic pressing (HIP) a mixture of salt-loaded zeolite powders and glass frit. A mechanistic description of the corrosion processes is being developed to support qualification of the CWF for disposal. The initial set of characterization tests included two standard tests that have been used extensively to study the corrosion behavior of high level waste (HLW) glasses: the Material Characterization Center-1 (MCC-1) Test and the Product Consistency Test (PCT). Direct comparison of the results of tests with the reference CWF and HLW glasses indicate that the corrosion behaviors of the CWF and HLW glasses are very similar

  13. Optimization of glass composition for the vitrification of nuclear waste at the Savannah River Plant

    International Nuclear Information System (INIS)

    Soper, P.D.; Roberts, G.J.; Lightner, L.F.; Walker, D.D.; Plodinec, M.J.

    1982-01-01

    Waste glasses of different compositions were compared in terms of leachability, viscosity, liquidus temperature, and coefficient of expansion. The compositions of the glasses were determined by statistical optimization. Waste glass of the optimized composition is more durable than the current reference composition but can still be processed at low temperature

  14. Development and testing of matrices for the encapsulation of glass and ceramic nuclear waste forms

    International Nuclear Information System (INIS)

    Wald, J.W.; Brite, D.W.; Gurwell, W.E.; Buckwalter, C.Q.; Bunnell, L.R.; Gray, W.J.; Blair, H.T.; Rusin, J.M.

    1982-02-01

    This report details the results of research on the matrix encapsulation of high level wastes at PML over the past few years. The demonstrations and tests described were designed to illustrate how the waste materials are effected when encapsulated in an inert matrix. Candidate materials evaluated for potential use as matrices for encapslation of pelletized ceramics or glass marbles were categorized into four groups: metals, glasses, ceramics, and graphite. Two processing techniques, casting and hot pressing, were investigated as the most promising methods of formation or densification of the matrices. The major results reported deal with the development aspects. However, chemical durability tests (leach tests) of the matrix materials themselves and matrix-waste form composites are also reported. Matrix waste forms can provide a low porosity, waste-free barrier resulting in increased leach protection, higher impact strength and improved thermal conductivity compared to unencapsulated glass or ceramic waste materials. Glass marbles encapsulated in a lead matrix offer the most significant improvement in waste form stability of all combinations evaluated. This form represents a readily demonstrable process that provides high thermal conductivity, mechanical shock resistance, radiation shielding and increased chemical durability through both a chemical passivation mechanism and as a physical barrier. Other durable matrix waste forms evaluated, applicable primarily to ceramic pellets, involved hot-pressed titanium or TiO 2 materials. In the processing of these forms, near 100% dense matrices were obtained. The matrix materials had excellent compatibility with the waste materials and superior potential chemical durability. Cracking of the hot-pressed ceramic matrix forms, in general, prevented the realization of their optimum properties

  15. Utilization of borosilicate glass for transuranic waste immobilization

    International Nuclear Information System (INIS)

    Ledford, J.A.; Williams, P.M.

    1979-01-01

    Incinerated transuranic waste and other low-level residues have been successfully vitrified by mixing with boric acid and sodium carbonate and heating to 1050 0 C in a bench-scale continuous melter. The resulting borosilicate glass demonstrates excellent mechanical durability and chemical stability

  16. High-level radioactive waste glass and storage canister design

    International Nuclear Information System (INIS)

    Slate, S.C.; Ross, W.A.

    1979-01-01

    Management of high-level radioactive wastes is a primary concern in nuclear operations today. The main objective in managing these wastes is to convert them into a solid, durable form which is then isolated from man. A description is given of the design and evaluation of this waste form. The waste form has two main components: the solidified waste and the storage canister. The solid waste form discussed in this study is glass. Waste glasses have been designed to be inert to water attack, physically rugged, low in volatility, and stable over time. Two glass-making processes are under development at PNL. The storage canister is being designed to provide high-integrity containment for solidified wastes from processing to terminal storage. An outline is given of the steps in canister design: material selection, stress and thermal analyses, quality verification, and postfill processing. Examples are given of results obtained from actual nonradioactive demonstration tests. 14 refs

  17. Properties and solubility of chrome in iron alumina phosphate glasses containing high level nuclear waste

    International Nuclear Information System (INIS)

    Huang, W.; Day, D.E.; Ray, C.S.; Kim, C.W.; Reis, S.T.D.

    2004-01-01

    Chemical durability, glass formation tendency, and other properties of iron alumina phosphate glasses containing 70 wt% of a simulated high level nuclear waste (HLW), doped with different amounts of Cr 2 O 3 , have been investigated. All of the iron alumina phosphate glasses had an outstanding chemical durability as measured by their small dissolution rate (1 . 10 -9 g/(cm 2 . min)) in deionized water at 90 C for 128 d, their low normalized mass release as determined by the product consistency test (PCT) and a barely measurable corrosion rate of 2 . d) after 7 d at 200 C by the vapor hydration test (VHT). The solubility limit for Cr 2 O 3 in the iron phosphate melts was estimated at 4.1 wt%, but all of the as-annealed melts contained a few percent of crystalline Cr 2 O 3 that had no apparent effect on the chemical durability. The chemical durability was unchanged after deliberate crystallization, 48 h at 650 C. These iron phosphate waste forms, with a waste loading of at least 70 wt%, can be readily melted in commercial refractory crucibles at 1250 C for 2 to 4 h, are resistant to crystallization, meet all current US Department of Energy requirements for chemical durability, and have a solubility limit for Cr 2 O 3 which is at least three times larger than that for borosilicate glasses. (orig.)

  18. West Valley high-level nuclear waste glass development: a statistically designed mixture study

    Energy Technology Data Exchange (ETDEWEB)

    Chick, L.A.; Bowen, W.M.; Lokken, R.O.; Wald, J.W.; Bunnell, L.R.; Strachan, D.M.

    1984-10-01

    The first full-scale conversion of high-level commercial nuclear wastes to glass in the United States will be conducted at West Valley, New York, by West Valley Nuclear Services Company, Inc. (WVNS), for the US Department of Energy. Pacific Northwest Laboratory (PNL) is supporting WVNS in the design of the glass-making process and the chemical formulation of the glass. This report describes the statistically designed study performed by PNL to develop the glass composition recommended for use at West Valley. The recommended glass contains 28 wt% waste, as limited by process requirements. The waste loading and the silica content (45 wt%) are similar to those in previously developed waste glasses; however, the new formulation contains more calcium and less boron. A series of tests verified that the increased calcium results in improved chemical durability and does not adversely affect the other modeled properties. The optimization study assessed the effects of seven oxide components on glass properties. Over 100 melts combining the seven components into a wide variety of statistically chosen compositions were tested. Viscosity, electrical conductivity, thermal expansion, crystallinity, and chemical durability were measured and empirically modeled as a function of the glass composition. The mathematical models were then used to predict the optimum formulation. This glass was tested and adjusted to arrive at the final composition recommended for use at West Valley. 56 references, 49 figures, 18 tables.

  19. Exploration and Modeling of Structural changes in Waste Glass Under Corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Pantano, Carlos; Ryan, Joseph; Strachan, Denis

    2013-11-10

    Vitrification is currently the world-wide treatment of choice for the disposition of high-level nuclear wastes. In glasses, radionuclides are atomistically bonded into the solid, resulting in a highly durable product, with borosilicate glasses exhibiting particularly excellent durability in water. Considering that waste glass is designed to retain the radionuclides within the waste form for long periods, it is important to understand the long-term stability of these materials when they react in the environment, especially in the presence of water. Based on a number of previous studies, there is general consensus regarding the mechanisms controlling the initial rate of nuclear waste glass dissolution. Agreement regarding the cause of the observed decrease in dissolution rate at extended times, however, has been elusive. Two general models have been proposed to explain this behavior, and it has been concluded that both concepts are valid and must be taken into account when considering the decrease in dissolution rate. Furthermore, other processes such as water diffusion, ion exchange, and precipitation of mineral phases onto the glass surface may occur in parallel with dissolution of the glass and can influence long-term performance. Our proposed research will address these issues through a combination of aqueous-phase dissolution/reaction experiments and probing of the resulting surface layers with state-of-the-art analytical methods. These methods include solid-state nuclear magnetic resonance (SSNMR) and time-of-flight secondary ion mass spectrometry (TOF-SIMS). The resulting datasets will then be coupled with computational chemistry and reaction-rate modeling to address the most persistent uncertainties in the understanding of glass corrosion, which indeed have limited the performance of the best corrosion models to date. With an improved understanding of corrosion mechanisms, models can be developed and improved that, while still conservative, take advantage of

  20. Glass-bonded iodosodalite waste form for immobilization of 129I

    Science.gov (United States)

    Chong, Saehwa; Peterson, Jacob A.; Riley, Brian J.; Tabada, Diana; Wall, Donald; Corkhill, Claire L.; McCloy, John S.

    2018-06-01

    Immobilization of radioiodine is an important requirement for current and future nuclear fuel cycles. Iodosodalite [Na8(AlSiO4)6I2] was synthesized hydrothermally from metakaolin, NaI, and NaOH. Dried unwashed sodalite powders were used to synthesize glass-bonded iodosodalite waste forms (glass composite materials) by heating pressed pellets at 650, 750, or 850 °C with two types of sodium borosilicate glass binders. These heat-treated specimens were characterized with X-ray diffraction, Fourier-transform infrared spectroscopy, scanning electron microscopy, energy dispersive spectroscopy, thermal analysis, porosity and density measurements, neutron activation analysis, and inductively-coupled plasma mass spectrometry. For the best waste form produced (pellets mixed with 10 mass% of glass binder and heat-treated at 750 °C), the maximum possible elemental iodine loading was 19.8 mass%, but only ∼8-9 mass% waste loading of iodine was retained in the waste form after thermal processing. Other pellets with higher iodine retention either contained higher porosity or were incompletely sintered. ASTM C1308 and C1285 (product consistency test, PCT) experiments were performed to understand chemical durability under diffusive and static conditions. The C1308 test resulted in significantly higher normalized loss compared to the C1285 test, most likely because of the strong effect of neutral pH solution renewal and prevention of ion saturation in solution. Both experiments indicated that release rates of Na and Si were higher than for Al and I, probably due to a poorly durable Na-Si-O phase from the glass bonding matrix or from initial sodalite synthesis; however the C1308 test result indicated that congruent dissolution of iodosodalite occurred. The average release rates of iodine obtained from C1308 were 0.17 and 1.29 g m-2 d-1 for 80 or 8 m-1, respectively, and the C1285 analysis gave a value of 2 × 10-5 g m-2 d-1, which is comparable to or better than the durability of

  1. Effect of the addition of Na2O on the thermal properties and chemical durability of glasses of iron and uranium phosphates

    International Nuclear Information System (INIS)

    Arboleda Zuluaga, P.A; Rodriguez, D.S; Gonzalez Oliver, C; Rincon Lopez, J.M; Soldera, F

    2012-01-01

    A series of glass compositions including (54,6-73,5P 2 O 5 .14-22Fe 2 O 3.x Na 2 O.2,8-4,25 UO 2 ) %mol. x=0-28,4 were studied in function of sodium oxide content for the thermal properties and chemical durability. By means differential dilatometer measurements was possible establish the variation of Tg, and α Tsoft and analysis of the kinetics of sintering by means of High Temperature Microscopy (MAT) and dilatometric data of pressed pellets. The presence of modifier oxides in the structure of iron phosphate glasses causes slightly onset sintering anticipation in almost 25 o C The chemical durability was estimated performing the named Product Consistency Test (PCT-B) focused on determining the resistance of glasses for nuclear wastes. These glasses exhibit good chemical durability but it is significant impaired by the addition of x≥6wt%Na 2 O. It is aimed to achieve more stable compositions and get glass matrixes able to contain more uranium oxides allowing evaluating the potential application of these iron phosphate glasses for special, industrial and nuclear waste immobilization

  2. Zirconolite glass-ceramics for plutonium immobilization: The effects of processing redox conditions on charge compensation and durability

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yingjie, E-mail: yzx@ansto.gov.au; Gregg, Daniel J.; Kong, Linggen; Jovanovich, Miodrag; Triani, Gerry

    2017-07-15

    Zirconolite glass-ceramic samples doped with plutonium have been prepared via hot isostatic pressing. The effects of processing redox and plutonium loadings on plutonium valences, the presence of cation vacancies, zirconolite phase compositions, microstructures and durability have been investigated. Either tetravalent or trivalent plutonium ions may be incorporated on the Ca-site of CaZrTi{sub 2}O{sub 7} zirconolite with the Ca-site cation vacancies and the incorporation of Al{sup 3+} ions on the Ti-site for charge compensation. Plutonium and gadolinium (as a neutron absorber) are predominantly partitioned in zirconolite phases leading to the formation of chemically durable glass-ceramics suitable for the immobilization of impure plutonium wastes arising from the nuclear fuel cycle. - Highlights: •Plutonium validations of zirconolite glass-ceramics. •Effects of processing redox and plutonium loading. •Zirconolite phase compositions and plutonium valences. •Cation vacancies and chemical durability.

  3. Modeling a novel glass immobilization waste treatment process using flow

    International Nuclear Information System (INIS)

    Ferrada, J.J.; Nehls, J.W. Jr.; Welch, T.D.; Giardina, J.L.

    1996-01-01

    One option for control and disposal of surplus fissile materials is the Glass Material Oxidation and Dissolution System (GMODS), a process developed at ORNL for directly converting Pu-bearing material into a durable high-quality glass waste form. This paper presents a preliminary assessment of the GMODS process flowsheet using FLOW, a chemical process simulator. The simulation showed that the glass chemistry postulated ion the models has acceptable levels of risks

  4. Analysis for silicon in solution in high level waste glass durability studies

    International Nuclear Information System (INIS)

    Lewis, R.A.; Smart, R.St.C.; Dale, L.S.; Levins, D.M.

    1982-01-01

    In comparative studies of the durability of HLW glasses, the measurement of the dissolution of the silicate network, in terms of both rate and extent, is of prime importance. To achieve this, analytical techniques such as colorimetry, flame atomic absorption spectrometry and inductively-coupled plasma emission spectrometry are used. The reliability of these analytical techniques for determination of silicon concentration in dissolution of HLW glasses, is examined. At high concentrations both FAA and ICP are accurate but colorimetry, even with HF pretreatment or NaOH digestion, does not give agreement with ICP. At concentrations below 40 mg l -1 all three methods are reliable. (Auth.)

  5. Porous glass matrix method for encapsulating high-level nuclear wastes

    International Nuclear Information System (INIS)

    Macedo, P.B.; Tran, D.C.; Simmons, J.H.; Saleh, M.; Barkatt, A.; Simmons, C.J.; Lagakos, N.; DeWitt, E.

    1979-01-01

    A novel process which uses solidified porous high-silica glass powder to fixate radioactive high-level wastes is described. The process yields cylinders consisting of a core of high-silica glass containing the waste elements in its structure and a protective layer also of high-silica glass completely free of waste elements. The process can be applied to waste streams containing 0 to 100% solids. The core region exhibits a higher coefficient of thermal expansion and a lower glass transition temperature than the outer protective layer. This leads to mechanical strengthening of the glass and good resistance to stress corrosion by the development of a high residual compressive stress on the surface of the sample. Both the core and the protective layer exhibit extremely high chemical durability and offer an effective fixation of the radioactive waste elements, including 239 Pu and 99 Tc which have long half-lives, for calculated periods of more than 1 million years, when temperatures are not allowed to rise above 100 0 C

  6. Magnetic Glass Ceramics by Sintering of Borosilicate Glass and Inorganic Waste

    Directory of Open Access Journals (Sweden)

    Inès M. M. M. Ponsot

    2014-07-01

    Full Text Available Ceramics and glass ceramics based on industrial waste have been widely recognized as competitive products for building applications; however, there is a great potential for such materials with novel functionalities. In this paper, we discuss the development of magnetic sintered glass ceramics based on two iron-rich slags, coming from non-ferrous metallurgy and recycled borosilicate glass. The substantial viscous flow of the glass led to dense products for rapid treatments at relatively low temperatures (900–1000 °C, whereas glass/slag interactions resulted in the formation of magnetite crystals, providing ferrimagnetism. Such behavior could be exploited for applying the obtained glass ceramics as induction heating plates, according to preliminary tests (showing the rapid heating of selected samples, even above 200 °C. The chemical durability and safety of the obtained glass ceramics were assessed by both leaching tests and cytotoxicity tests.

  7. Glass Ceramic Waste Forms for Combined CS+LN+TM Fission Products Waste Streams

    International Nuclear Information System (INIS)

    Crum, Jarrod V.; Turo, Laura A.; Riley, Brian J.; Tang, Ming; Kossoy, Anna; Sickafus, Kurt E.

    2010-01-01

    In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.(1) The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development of a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both combined waste

  8. Glass Ceramic Waste Forms for Combined CS+LN+TM Fission Products Waste Streams

    Energy Technology Data Exchange (ETDEWEB)

    Crum, Jarrod V.; Turo, Laura A.; Riley, Brian J.; Tang, Ming; Kossoy, Anna; Sickafus, Kurt E.

    2010-09-23

    In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.[1] The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development of a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both combined waste

  9. Investigation of metastable immiscibility in nuclear-waste-glasses. I-III

    International Nuclear Information System (INIS)

    Egnell, J.; Larsen, J.G.; Moeller, L.; Roed, G.

    1981-12-01

    Metastable liquid-liquid separation in glasses can often cause significant changes in physical and chemical properties of the original homogeneous glass. In some technical borosilicate glasses this phenomenon is used to change the chemical durability of the glass. For potential nuclear-waste-glasses the slow cooling through the temperature range 550 0 C - 700 0 C may lead to such a liquid-liquid phase separation. In order to investigate the susceptibility of phase separation of nuclear-waste-glasses, two KBS model glasses, ABS-39 and ABS-41, were investigated. Two of the subsequent reports are concerned with this problem. The third report also takes into consideration the effects of MoO 3 on the immiscibility gap. The maximum amount of MoO 3 that can be dissolved in ABS-39 and ABS 41 is also determined. (Auth.)

  10. Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

    2017-11-01

    This paper provides an overview of research evaluating the use of tellurite glass as a waste form for salt wastes from electrochemical processing. The capacities to immobilize different salts were evaluated including: a LiCl-Li2O oxide reduction salt (for oxide fuel) containing fission products, a LiCl-KCl eutectic salt (for metallic fuel) containing fission products, and SrCl2. Physical and chemical properties of the glasses were characterized by using X-ray diffraction, bulk density measurements, chemical durability tests, scanning electron microscopy, and energy dispersive X-ray emission spectroscopy. These glasses were found to accommodate high concentrations of halide salts and have high densities. However, improvements are needed to meet chemical durability requirements.

  11. Effects of radionuclide decay on waste glass behavior: A critical review

    International Nuclear Information System (INIS)

    Wronkiewicz, D.J.

    1993-12-01

    This paper is an extension of a chapter in an earlier report [1] that provides an updated review on the status of radiation damage problems in nuclear waste glasses. This report will focus on radiation effects on vitrified borosilicate nuclear waste glasses under conditions expected in the proposed Yucca mountain repository. Radiation effects on high-level waste glasses and their surrounding repository environment are important considerations for radionuclide immobilization because of the potential to alter the glass stability and thereby influence the radionuclide retentive properties of this waste form. The influence of radionuclide decay on vitrified nuclear waste may be manifested by several changes, including volume, stored energy, structure, microstructure, mechanical properties, and phase separation. Radiation may also affect the composition of aqueous fluids and atmospheric gases in relatively close proximity to the waste form. What is important to the radionuclide retentive properties of the repository is how these radiation effects collectively or individually influence the durability and radionuclide release from the glass in the event of liquid water contact

  12. Settling of Spinel in A High-Level Waste Glass Melter

    International Nuclear Information System (INIS)

    Pavel Hrma; Pert Schill; Lubomir Nemec

    2002-01-01

    High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors call melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150 degree C (or even higher in advanced melters) to create a melt that becomes glass on cooling. This process is slow and expensive. Moreover, the melters that are currently in use or are going to be used in the U.S. are sensitive to clogging and thus cannot process melt in which solid particles are suspended. These particles settle and gradually accumulate on the melter bottom. Such particles, most often small crystals of spinel ( a mineral containing iron, nickel, chromium, and other minor oxides), inevitably occurred in the melt when the content of the waste in the glass (called waste loading) increases above a certain limit. To avoid the presence of solid particles in the melter, the waste loading is kept rather low, in average 15% lower than in glass formulated for more robust melters

  13. Examining the role of canister cooling conditions on the formation of nepheline from nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-01

    Nepheline (NaAlSiO₄) crystals can form during slow cooling of high-level waste (HLW) glass after it has been poured into a waste canister. Formation of these crystals can adversely affect the chemical durability of the glass. The tendency for nepheline crystallization to form in a HLW glass increases with increasing concentrations of Al₂O₃ and Na₂O.

  14. Immobilization of simulated high-level radioactive waste in borosilicate glass: Pilot scale demonstrations

    International Nuclear Information System (INIS)

    Ritter, J.A.; Hutson, N.D.; Zamecnik, J.R.; Carter, J.T.

    1991-01-01

    The Integrated DWPF Melter System (IDMS), operated by the Savannah River Laboratory, is a pilot scale facility used in support of the start-up and operation of the Department of Energy's Defense Waste Processing Facility. The IDMS has successfully demonstrated, on an engineering scale (one-fifth), that simulated high level radioactive waste (HLW) sludge can be chemically treated with formic acid to adjust both its chemical and physical properties, and then blended with simulated precipitate hydrolysis aqueous (PHA) product and borosilicate glass frit to produce a melter feed which can be processed into a durable glass product. The simulated sludge, PHA and frit were blended, based on a product composition program, to optimize the loading of the waste glass as well as to minimize those components which can cause melter processing and/or glass durability problems. During all the IDMS demonstrations completed thus far, the melter feed and the resulting glass that has been produced met all the required specifications, which is very encouraging to future DWPF operations. The IDMS operations also demonstrated that the volatile components of the melter feed (e.g., mercury, nitrogen and carbon, and, to a lesser extent, chlorine, fluorine and sulfur) did not adversely affect the melter performance or the glass product

  15. High-silica glass matrix process for high-level waste solidification

    International Nuclear Information System (INIS)

    Simmons, J.H.; Macedo, P.B.

    1981-01-01

    In the search for an optimum glass matrix composition, we have determined that chemical durability and thermal stability are maximized, and that stress development is minimized for glass compositions containing large concentrations of glass-forming oxides, of which silica is the major component (80 mol%). These properties and characteristics were recently demonstrated to belong to very old geological glasses known as tektites (ages of 750,000 to 34 million years.) The barrier to simulating tektite compositions for the waste glasses was the high melting temperature (1600 to 1800 0 C) needed for these glasses. Such temperatures greatly complicate furnace design and maintenance and lead to an intolerable vaporization of many of the radioisotopes into the off-gas system. Research conducted at our laboratory led to the development of a porous high-silica waste glass material with approximately 80% SiO 2 by mole and 30% waste loading by weight. The process can handle a wide variety of compositions, and yields long, elliptical, monolithic samples, which consist of a loaded high-silica core completely enveloped in a high-silica glass tube, which has collapsed upon the core and sealed it from the outside. The outer glass layer is totally free of waste isotopes and provides an integral multibarrier protection system

  16. Glass: a candidate engineered material for management of high level nuclear waste

    International Nuclear Information System (INIS)

    Mishra, R.K.; Kaushik, C.P.

    2011-01-01

    While the commercial importance of glass is generally recognized, a few people are aware of extremely wide range of glass formulations that can be made and of the versatility of this engineered material. Some of the recent developments in the field of glass leading to various technological applications include glass fiber reinforcement of cement to give new building materials, substrates for microelectronics circuitry in form of semiconducting glasses, nuclear waste immobilization and specific medical applications. The present paper covers fundamental understanding of glass structure and its application for immobilization of high level radioactive liquid waste. High level radioactive liquid waste (HLW) arising during reprocessing of spent fuel are immobilized in sodium borosilicate glass matrix developed indigenously. Glass compositions are modified according to the composition of HLW to meet the criteria of desirable properties in terms. These glass matrices have been characterized for different properties like homogeneity, chemical durability, thermal stability and radiation stability. (author)

  17. Predictive modeling of crystal accumulation in high-level waste glass melters processing radioactive waste

    Science.gov (United States)

    Matyáš, Josef; Gervasio, Vivianaluxa; Sannoh, Sulaiman E.; Kruger, Albert A.

    2017-11-01

    The effectiveness of high-level waste vitrification at Hanford's Waste Treatment and Immobilization Plant may be limited by precipitation/accumulation of spinel crystals [(Fe, Ni, Mn, Zn)(Fe, Cr)2O4] in the glass discharge riser of Joule-heated ceramic melters during idling. These crystals do not affect glass durability; however, if accumulated in thick layers, they can clog the melter and prevent discharge of molten glass into canisters. To address this problem, an empirical model was developed that can predict thicknesses of accumulated layers as a function of glass composition. This model predicts well the accumulation of single crystals and/or small-scale agglomerates, but excessive agglomeration observed in high-Ni-Fe glass resulted in an underprediction of accumulated layers, which gradually worsened over time as an increased number of agglomerates formed. The accumulation rate of ∼53.8 ± 3.7 μm/h determined for this glass will result in a ∼26 mm-thick layer after 20 days of melter idling.

  18. Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

    2017-11-01

    This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability. (C) 2017 Published by Elsevier B.V.

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

    International Nuclear Information System (INIS)

    Zhang, Hua; Corkhill, Claire L.; Heath, Paul G.; Hand, Russell J.; Stennett, Martin C.; Hyatt, Neil C.

    2015-01-01

    Highlights: • Spinel crystallization incorporates ZnO from base glass, displacing Mg and Ni. • Raman spectroscopy demonstrates significant impact on glass structure by addition of ZnO to base glass. • Addition of ZnO reduces glass dissolution rate at early time periods (up to 28 days). - Abstract: 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 Na 2 O/Li 2 O and the ZnO/CaO modified base glass compositions; the volume fraction of the spinel crystallites increased with increasing waste loading from 15 to 20 wt%. 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 (Zn 0.60 Ni 0.20 Mg 0.20 )(Cr 1.37 Fe 0.63 )O 4 . 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 Q 3 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 Na 2 O/Li 2 O base glass up to 28 days, due to a combination of the enhanced network

  20. Apatite and sodalite based glass-bonded waste forms for immobilization of 129I and mixed halide radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Goel, Ashutosh [Rutgers Univ., New Brunswick, NJ (United States); McCloy, John S. [Washington State Univ., Pullman, WA (United States); Riley, Brian J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-12-30

    The goal of the project was to utilize the knowledge accumulated by the team, in working with minerals for chloride wastes and biological apatites, toward the development of advanced waste forms for immobilizing 129I and mixed-halide wastes. Based on our knowledge, experience, and thorough literature review, we had selected two minerals with different crystal structures and potential for high chemical durability, sodalite and CaP/PbV-apatite, to form the basis of this project. The focus of the proposed effort was towards: (i) low temperature synthesis of proposed minerals (iodine containing sodalite and apatite) leading to the development of monolithic waste forms, (ii) development of a fundamental understanding of the atomic-scale to meso-scale mechanisms of radionuclide incorporation in them, and (iii) understanding of the mechanism of their chemical corrosion, alteration mechanism, and rates. The proposed work was divided into four broad sections. deliverables. 1. Synthesis of materials 2. Materials structural and thermal characterization 3. Design of glass compositions and synthesis glass-bonded minerals, and 4. Chemical durability testing of materials.

  1. Comprehensive data base of high-level nuclear waste glasses: September 1987 status report: Volume 1, Discussion and glass durability data

    International Nuclear Information System (INIS)

    Kindle, C.H.; Kreiter, M.R.

    1987-12-01

    The Materials Characterization Center (MCC) at Pacific Northwest Laboratory is assembling a comprehensive data base (CDB) of experimental data collected for high-level nuclear waste package components. Data collected throughout the world are included in the data base; current emphasis is on waste glasses and their properties. The goal is to provide a data base of properties and compositions and an analysis of dominant property trends as a function of composition. This data base is a resource that nuclear waste producers, disposers, and regulators can use to compare properties of a particular high-level nuclear waste glass product with the properties of other glasses of similar compositions. Researchers may use the data base to guide experimental tests to fill gaps in the available knowledge or to refine empirical models. The data are incorporated into a computerized data base that will allow the data to be extracted based on, for example, glass composition or test duration. 3 figs

  2. Glass consistency and glass performance

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Ramsey, W.G.

    1994-01-01

    Glass produced by the Defense Waste Processing Facility (DWPF) will have to consistently be more durable than a benchmark glass (evaluated using a short-term leach test), with high confidence. The DWPF has developed a Glass Product Control Program to comply with this specification. However, it is not clear what relevance product consistency has on long-term glass performance. In this report, the authors show that DWPF glass, produced in compliance with this specification, can be expected to effectively limit the release of soluble radionuclides to natural environments. However, the release of insoluble radionuclides to the environment will be limited by their solubility, and not glass durability

  3. A generalized definition for waste form durability

    International Nuclear Information System (INIS)

    Fanning, T. H.; Bauer, T. H.; Morris, E. E.; Wigeland, R. A.

    2002-01-01

    When evaluating waste form performance, the term ''durability'' often appears in casual discourse, but in the technical literature, the focus is often on waste form ''degradation'' in terms of mass lost per unit area per unit time. Waste form degradation plays a key role in developing models of the long-term performance in a repository environment, but other factors also influence waste form performance. These include waste form geometry; density, porosity, and cracking; the presence of cladding; in-package chemistry feedback; etc. The paper proposes a formal definition of waste form ''durability'' which accounts for these effects. Examples from simple systems as well as from complex models used in the Total System Performance Assessment of Yucca Mountain are provided. The application of ''durability'' in the selection of bounding models is also discussed

  4. The influence of penetrating gamma radiation on the reaction of simulated nuclear waste glass in tuff groundwater

    International Nuclear Information System (INIS)

    Ebert, W.L.; Bates, J.K.; Abrajano, T.A. Jr.; Gerding, T.J.

    1989-01-01

    Static leaching experiments have been performed to determine the influence of penetrating gamma radiation on the reaction of simulated nuclear waste glass in tuff groundwater at 90/degree/C. Both the leachates and the reacted glass monoliths were analyzed to characterize the reaction. Radiation was seen to acidify the leachates, but the high bicarbonate content of the groundwater prevented the pHs from dropping below 6.4. The glass reaction tended to raise the pH. Glass based on SRL 165 black frit and PNL 76-68 glass compositions were leached. The SRL 165 type glasses were quite durable and unaffected by radiation [NL(B) /approximately/4 g/m 2 after 278 days at all exposure rates]. The PNL 76-68 glasses were much less durable, with the durability decreasing as the exposure rate was increased [NL(B) was about 20 g/m 2 after 278 days at 1 /times/ 10 3 R/h for both ATM-1c and ATM-8 glasses]. 8 refs., 5 figs

  5. XRF and leaching characterization of waste glasses derived from wastewater treatment sludges

    International Nuclear Information System (INIS)

    Ragsdale, R.G., Jr.

    1994-12-01

    Purpose of this study was to investigate use of XRF (x-ray fluorescence spectrometry) as a near real-time method to determine melter glass compositions. A range of glasses derived from wastewater treatment sludges associated with DOE sites was prepared. They were analyzed by XRF and wet chemistry digestion with atomic absorption/inductively coupled emission spectrometry. Results indicated good correlation between these two methods. A rapid sample preparation and analysis technique was developed and demonstrated by acquiring a sample from a pilot-scale simulated waste glass melter and analyzing it by XRF within one hour. From the results, XRF shows excellent potential as a process control tool for waste glass vitrification. Glasses prepared for this study were further analyzed for durability by toxicity characteristic leaching procedure and product consistency test and results are presented

  6. Insertion of marble waste in the production chain of glass wool

    International Nuclear Information System (INIS)

    Rodrigues, G.F.; Alves, J.O.; Espinosa, D.C.R.; Tenorio, J.A.S.

    2010-01-01

    The work aimed the study of the recycle of the waste from marble cutting, aiming the reuse as partial raw material in the production of glass wool. Glass wool are materials with chemical and mechanical resistance, durability and lightness, and also important thermo-acoustic properties. A mixture of the waste with chemical additives was melted in a laboratory electric furnace using temperature of 1450 deg C. The melted material was directly poured in a water-filled recipient aiming the rapidly cooling. Samples of the produced material were characterized by XRD, SEM and DTA. The results showed that the residue from marble cutting can be inserted into the productive chain of glass wool, providing a decrease in the extraction of mineral resources, a profitable destination for this waste, and a economy for the companies producer of thermo-acoustic insulators. (author)

  7. Chemical durability of lead borosilicate glass matrix under simulated geological conditions

    International Nuclear Information System (INIS)

    Yalmali, Vrunda S.; Deshingkar, D.S.; Wattal, P.K.

    2002-03-01

    The lead borosilicate glass has been developed for vitrification of High Level Waste (HLW) stored at Trombay. This waste is contains especially high contents of sodium, uranium sulphate and iron. The glasses containing HLW are to be ultimately disposed into deep geological repositories. Long term leach rates under simulated geological conditions need to be evaluated for glass matrix. Studies were taken up to estimate the lead borosilicate glass WTR-62 matrix for chemical durability in presence of synthetic ground water. The leachant selected was based on composition of ground water sample near proposed repository site. In the first phase of these tests, the experiments were conducted for short duration of one and half month. The leaching experiments were conducted in presence of a) distilled water b) synthetic ground water c) synthetic ground water containing granite, bentonite and ferric oxide and d) synthetic ground water containing humic acid at 100 0 C. The leachate samples were analysed by pHmetry , ion chromatography and UV -VIS spectrophotometry. The normalised leach rates for lead borosilicate WTR- 62 glass matrix based on silica, boron and sulphate analyses of leachates were of the order of 10 -3 to 10 -5 gms/cm 2 /day for 45 days test period in presence of synthetic ground water as well as in presence of other materials likely to be present along with synthetic ground water. These rates are comparable to those of sodium borsilicate glass matrices reported in literature. It is known that the leach rates of glass matrix decrease with longer test durations due to formation of leached layer on its surface. The observed leach rates of lead borosilicate WTR- 62 glass matrix for 45 day tests under simulated geological conditions were found to be sufficiently encouraging to take up long term tests for evaluating its performances under repository conditions. (author)

  8. Performance of a buried radioactive high level waste (HLW) glass after 24 years

    International Nuclear Information System (INIS)

    Jantzen, Carol M.; Kaplan, Daniel I.; Bibler, Ned E.; Peeler, David K.; John Plodinec, M.

    2008-01-01

    A radioactive high level waste glass was made in 1980 with Savannah River Site (SRS) Tank 15 waste. This glass was buried in a lysimeter in the SRS burial ground for 24 years. Lysimeter leachate data was available for the first 8 years. The glass was exhumed in 2004. The glass was predicted to be very durable and laboratory tests confirmed this. Scanning electron microscopy of the glass burial surface showed no significant glass alteration consistent with results of other laboratory and field tests. Radionuclide profiling for alpha, beta, and 137 Cs indicated that Pu was not enriched in the soil while 137 Cs and 9 deg. C Sr were enriched in the first few centimeters surrounding the glass. Lysimeter leachate data indicated that 9 deg. C Sr and 137 Cs leaching from the glass was diffusion controlled

  9. Chemistry and kinetics of waste glass corrosion

    International Nuclear Information System (INIS)

    Bates, J.K.

    1996-01-01

    Under repository disposal conditions, the reaction of glass with water comprises the source term for release of radionuclides to the near-field environment. An understanding of glass reaction and the manner by which radionuclides are released is needed to design the waste package and to evaluate the total performance of the repository. The ASTM Standard C-1174-91 provides a general methodology for obtaining information related to the behavior of glass. This paper reviews the application of this standard to glass reaction. In the first step in the ASTM approach, the researcher identifies the materials and the conditions under which the long-term behavior is to be determined. Glass compositions have undergone a genesis over the past 15 years in response to concerns about feed streams, processing, and durability. A range of borosilicate compositions has been identified, but as new applications for vitrification occur, for example, immobilization of weapons plutonium and residue from plutonium processing, different compositions must be evaluated. The repository environment depends on the spatial emplacement of waste containers (glass and spent fuel), and both open-quotes hotclose quotes and open-quotes coldclose quotes scenarios have been proposed for the Yucca Mountain site. Regardless of the exact configuration, the near-field hydrology is expected to be unsaturated: that is, the waste packages are contacted initially by water vapor, and ultimately by small amounts of dripping or standing water. The behavior of glass can be studied as a function of composition within the constraints the environmental conditions place on the physical parameters that affect glass reaction (temperature, radiation field, groundwater composition, etc.). In the second step, the researcher reviews the literature and proposes a reaction pathway by which glass reacts in an unsaturated environment

  10. A Review of Iron Phosphate Glasses and Recommendations for Vitrifying Hanford Waste

    Energy Technology Data Exchange (ETDEWEB)

    Delbert E. Ray; Chandra S. Ray

    2013-11-01

    This report contains a comprehensive review of the research conducted, world-wide, on iron phosphate glass over the past ~30 years. Special attention is devoted to those iron phosphate glass compositions which have been formulated for the purpose of vitrifying numerous types of nuclear waste, with special emphasis on the wastes stored in the underground tanks at Hanford WA. Data for the structural, chemical, and physical properties of iron phosphate waste forms are reviewed for the purpose of understanding their (a) outstanding chemical durability which meets all current DOE requirements, (b) high waste loadings which can exceed 40 wt% (up to 75 wt%) for several Hanford wastes, (c) low melting temperatures, can be as low as 900°C for certain wastes, and (d) high tolerance for “problem” waste components such as sulfates, halides, and heavy metals (chromium, actinides, noble metals, etc.). Several recommendations are given for actions that are necessary to smoothly integrate iron phosphate glass technology into the present waste treatment plans and vitrification facilities at Hanford.

  11. Chemical durability of silicoborate glasses

    International Nuclear Information System (INIS)

    Nieto, M.I.; Rodriguez, M.A.; Rubio, J.; Fernandez, A.; Oteo, J.L.

    1987-01-01

    A general view of the durability in silicoborate glasses is presented with more emphasis on the etching factors (chemical composition, lattice structure, pH...) the techniques used for this study and the experimental results. Likewise, the research presently developed in this area at the Instituto de Ceramica y Vidrio, CSIC, is related to the applications. Future research in this field is also mentioned. (author) 15 figs

  12. Corrosion behaviors of a glass-bonded sodalite ceramic waste form and its constituents

    International Nuclear Information System (INIS)

    Lewis, M. A.; Ebert, W. L.; Morss, L.

    1999-01-01

    A ceramic waste form (CWF) of glass bonded sodalite is being developed as a waste form for the long-term immobilization of fission products and transuranic elements from the U.S. Department of Energy's activities on spent nuclear fuel conditioning. A durable waste form was prepared by hot isostatic pressing (HIP) a mixture of salt-loaded zeolite powders and glass frit. During HIP the zeolite is converted to sodalite, and the resultant CWF is been completed for durations of up to 182 days. Four dissolution modes were identified: dissolution of free salt, dissolution of the aluminosilicate matrix of sodalite and the accompanying dissolution of occluded salt, dissolution of the boroaluminosilicate matrix of the glass, and ion exchange. Synergies inherent to the CWF were identified by comparing the results of the tests with pure glass and sodalite with those of the composite CWF

  13. Simulation of Self-Irradiation of High-Sodium Content Nuclear Waste Glasses

    International Nuclear Information System (INIS)

    Pankov, Alexey S.; Ojovan, Michael I.; Batyukhnova, Olga G.; Lee, William E.

    2007-01-01

    Alkali-borosilicate glasses are widely used in nuclear industry as a matrix for immobilisation of hazardous radioactive wastes. Durability or corrosion resistance of these glasses is one of key parameters in waste storage and disposal safety. It is influenced by many factors such as composition of glass and surrounding media, temperature, time and so on. As these glasses contain radioactive elements most of their properties including corrosion resistance are also impacted by self-irradiation. The effect of external gamma-irradiation on the short-term (up to 27 days) dissolution of waste borosilicate glasses at moderate temperatures (30 deg. to 60 deg. C) was studied. The glasses studied were Magnox Waste glass used for immobilisation of HLW in UK, and K-26 glass used in Russia for ILW immobilisation. Glass samples were irradiated under γ-source (Co-60) up to doses 1 and 11 MGy. Normalised rates of elemental release and activation energy of release were measured for Na, Li, Ca, Mg, B, Si and Mo before and after irradiation. Irradiation up to 1 MGy results in increase of leaching rate of almost all elements from both MW and K-26 with the exception of Na release from MW glass. Further irradiation up to a dose of 11 MGy leads to the decrease of elemental release rates to nearly initial value. Another effect of irradiation is increase of activation energies of elemental release. (authors)

  14. SUMMARY OF 2010 DOE EM INTERNATIONAL PROGRAM STUDIES OF WASTE GLASS STRUCTURE AND PROPERTIES

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K.; Choi, A.; Marra, J.; Billings, A.

    2011-02-07

    Collaborative work between the Savannah River National Laboratory (SRNL) and SIA Radon in Russia was divided among three tasks for calendar year 2010. The first task focused on the study of simplified high level waste glass compositions with the objective of identifying the compositional drivers that lead to crystallization and poor chemical durability. The second task focused on detailed characterization of more complex waste glass compositions with unexpectedly poor chemical durabilities. The third task focused on determining the structure of select high level waste glasses made with varying frit compositions in order to improve models under development for predicting the melt rate of the Defense Waste Processing Facility (DWPF) glasses. The majority of these tasks were carried out at SIA Radon. Selection and fabrication of the glass compositions, along with chemical composition measurements and evaluations of durability were carried out at SRNL and are described in this report. SIA Radon provided three summary reports based on the outcome of the three tasks. These reports are included as appendices to this document. Briefly, the result of characterization of the Task 1 glasses may indicate that glass compositions where iron is predominantly tetrahedrally coordinated have more of a tendency to crystallize nepheline or nepheline-like phases. For the Task 2 glasses, the results suggested that the relatively low fraction of tetrahedrally coordinated boron and the relatively low concentrations of Al{sub 2}O{sub 3} available to form [BO{sub 4/2}]{sup -}Me{sup +} and [AlO{sub 4/2}]{sup -}Me{sup +} tetrahedral units are not sufficient to consume all of the alkali ions, and thus these alkali ions are easily leached from the glasses. All of the twelve Task 3 glass compositions were determined to be mainly amorphous, with some minor spinel phases. Several key structural units such as metasilicate chains and rings were identified, which confirms the current modeling

  15. Barium borosilicate glass - a potential matrix for immobilization of sulfate bearing high-level radioactive liquid waste

    International Nuclear Information System (INIS)

    Kaushik, C.P.; Mishra, R.K.; Sengupta, P.; Kumar, Amar; Das, D.; Kale, G.B.; Raj, Kanwar

    2006-01-01

    Borosilicate glass formulations adopted worldwide for immobilization of high-level radioactive liquid waste (HLW) is not suitable for sulphate bearing HLW, because of its low solubility in such glass. A suitable glass matrix based on barium borosilicate has been developed for immobilization of sulphate bearing HLW. Various compositions based on different glass formulations were made to examine compatibility with waste oxide with around 10 wt% sulfate content. The vitrified waste product obtained from barium borosilicate glass matrix was extensively evaluated for its characteristic properties like homogeneity, chemical durability, glass transition temperature, thermal conductivity, impact strength, etc. using appropriate techniques. Process parameters like melt viscosity and pour temperature were also determined. It is found that SB-44 glass composition (SiO 2 : 30.5 wt%, B 2 O 3 : 20.0 wt%, Na 2 O: 9.5 wt% and BaO: 19.0 wt%) can be safely loaded with 21 wt% waste oxide without any phase separation. The other product qualities of SB-44 waste glass are also found to be on a par with internationally adopted waste glass matrices. This formulation has been successfully implemented in plant scale

  16. Product consistency leach tests of Savannah River Site radioactive waste glasses

    International Nuclear Information System (INIS)

    Bibler, N.E.; Bates, J.K.

    1989-01-01

    The Product Consistency Test (PCT) is a glass leach test that was developed at the Savannah River Site (SRS) to routinely confirm the durability of nuclear waste glasses that will be produced in the Defense Waste Processing Facility. The PCT is a 7 day, crushed glass leach test in deionized water at 90 degree C. Final leachates are filtered and acidified prior to analysis. To demonstrate the reproducibility of the PCT when performed remotely, SRS and Argonne National Laboratory have performed the PCT on samples of two radioactive glasses. The tests were also performed to compare the releases of the radionuclides with the major nonradioactive glass components and to determine if radiation from the glass was affecting the results of the PCT. The test was performed in triplicate at each laboratory. For the major soluble elements, B, Li, Na, and Si, in the glass, each investigator obtained relative precisions in the range 2--5% in the triplicate tests. This range indicates good precision for the PCT when performed remotely with master slave manipulators in a shielded cell environment

  17. Comparative study of seven glasses for solidification of nuclear wastes

    International Nuclear Information System (INIS)

    Nogues, J.L.; Hench, L.L.; Zarzycki, J.

    1982-06-01

    The relative leaching behavior of seven alkali borosilicate glasses considered for immobilization of high level radioactive wastes was compared using a static 90 0 C leach test. Leaching times studied were 1, 3, 7, 14 and 28 days with ratios of glass surface area (SA) to solution volume (V) being SA/V = 1.0 cm -1 and 0.1 cm -1 . With the range of glass compositions studied, it was not possible to determine the effect of each element on leaching behavior, however some conclusions regarding the general influence of the glass network formers can be made: the addition of Al 2 O 3 , results in a large increase in the chemical durability of the glass. The presence of Fe 2 O 3 , is necessary to develop with Al 2 O 3 a second protective layer on top of the silica-rich film that results from rapid dealkalization. The difference between the results obtained at SA/V = 1.0 cm -1 and 0.1 cm -1 shows the importance of understanding both the effects of glass composition and solution concentrations on the behavior of nuclear waste glasses

  18. Effects of composition on properties in an 11-component nuclear waste glass system

    International Nuclear Information System (INIS)

    Chick, L.A.; Piepel, G.F.; Mellinger, G.B.; May, R.P.; Gray, W.J.; Buckwalter, C.Q.

    1981-09-01

    Ninety simplified nuclear waste glass compositions within an 11-component oxide composition matrix were tested for crystallinity, viscosity, volatility, and chemical durability. Empirical models of property response as a function of glass composition were developed using statistical experimental design and modeling techniques. A new statistical technique was developed to calculate the effects of oxide components on each property. Independent melts were used to check the prediction accuracy of the models

  19. Low sintering temperature glass waste forms for sequestering radioactive iodine

    Science.gov (United States)

    Nenoff, Tina M.; Krumhansl, James L.; Garino, Terry J.; Ockwig, Nathan W.

    2012-09-11

    Materials and methods of making low-sintering-temperature glass waste forms that sequester radioactive iodine in a strong and durable structure. First, the iodine is captured by an adsorbant, which forms an iodine-loaded material, e.g., AgI, AgI-zeolite, AgI-mordenite, Ag-silica aerogel, ZnI.sub.2, CuI, or Bi.sub.5O.sub.7I. Next, particles of the iodine-loaded material are mixed with powdered frits of low-sintering-temperature glasses (comprising various oxides of Si, B, Bi, Pb, and Zn), and then sintered at a relatively low temperature, ranging from 425.degree. C. to 550.degree. C. The sintering converts the mixed powders into a solid block of a glassy waste form, having low iodine leaching rates. The vitrified glassy waste form can contain as much as 60 wt % AgI. A preferred glass, having a sintering temperature of 500.degree. C. (below the silver iodide sublimation temperature of 500.degree. C.) was identified that contains oxides of boron, bismuth, and zinc, while containing essentially no lead or silicon.

  20. Treatment of copper industry waste and production of sintered glass-ceramic.

    Science.gov (United States)

    Coruh, Semra; Ergun, Osman Nuri; Cheng, Ta-Wui

    2006-06-01

    Copper waste is iron-rich hazardous waste containing heavy metals such as Cu, Zn, Co, Pb. The results of leaching tests show that the concentration of these elements exceeds the Turkish and EPA regulatory limits. Consequently, this waste cannot be disposed of in its present form and therefore requires treatment to stabilize it or make it inert prior to disposal. Vitrification was selected as the technology for the treatment of the toxic waste under investigation. During the vitrification process significant amounts of the toxic organic and inorganic chemical compounds could be destroyed, and at the same time, the metal species are immobilized as they become an integral part of the glass matrix. The copper flotation waste samples used in this research were obtained from the Black Sea Copper Works of Samsun, Turkey. The samples were vitrified after being mixed with other inorganic waste and materials. The copper flotation waste and their glass-ceramic products were characterized by X-ray analysis (XRD), scanning electron microscopy and by the toxicity characteristic leaching procedure test. The products showed very good chemical durability. The glass-ceramics fabricated at 850 degrees C/2 h have a large application potential especially as construction and building materials.

  1. PLUTONIUM SOLUBILITY IN HIGH-LEVEL WASTE ALKALI BOROSILICATE GLASS

    Energy Technology Data Exchange (ETDEWEB)

    Marra, J.; Crawford, C.; Fox, K.; Bibler, N.

    2011-01-04

    The solubility of plutonium in a Sludge Batch 6 (SB6) reference glass and the effect of incorporation of Pu in the glass on specific glass properties were evaluated. A Pu loading of 1 wt % in glass was studied. Prior to actual plutonium glass testing, surrogate testing (using Hf as a surrogate for Pu) was conducted to evaluate the homogeneity of significant quantities of Hf (Pu) in the glass, determine the most appropriate methods to evaluate homogeneity for Pu glass testing, and to evaluate the impact of Hf loading in the glass on select glass properties. Surrogate testing was conducted using Hf to represent between 0 and 1 wt % Pu in glass on an equivalent molar basis. A Pu loading of 1 wt % in glass translated to {approx}18 kg Pu per Defense Waste Processing Facility (DWPF) canister, or about 10X the current allowed limit per the Waste Acceptance Product Specifications (2500 g/m{sup 3} of glass or about 1700 g/canister) and about 30X the current allowable concentration based on the fissile material concentration limit referenced in the Yucca Mountain Project License Application (897 g/m{sup 3}3 of glass or about 600 g Pu/canister). Based on historical process throughput data, this level was considered to represent a reasonable upper bound for Pu loading based on the ability to provide Pu containing feed to the DWPF. The task elements included evaluating the distribution of Pu in the glass (e.g. homogeneity), evaluating crystallization within the glass, evaluating select glass properties (with surrogates), and evaluating durability using the Product Consistency Test -- Method A (PCT-A). The behavior of Pu in the melter was evaluated using paper studies and corresponding analyses of DWPF melter pour samples.The results of the testing indicated that at 1 wt % Pu in the glass, the Pu was homogeneously distributed and did not result in any formation of plutonium-containing crystalline phases as long as the glass was prepared under 'well-mixed' conditions

  2. Plutonium Solubility In High-Level Waste Alkali Borosilicate Glass

    International Nuclear Information System (INIS)

    Marra, J.; Crawford, C.; Fox, K.; Bibler, N.

    2011-01-01

    The solubility of plutonium in a Sludge Batch 6 (SB6) reference glass and the effect of incorporation of Pu in the glass on specific glass properties were evaluated. A Pu loading of 1 wt % in glass was studied. Prior to actual plutonium glass testing, surrogate testing (using Hf as a surrogate for Pu) was conducted to evaluate the homogeneity of significant quantities of Hf (Pu) in the glass, determine the most appropriate methods to evaluate homogeneity for Pu glass testing, and to evaluate the impact of Hf loading in the glass on select glass properties. Surrogate testing was conducted using Hf to represent between 0 and 1 wt % Pu in glass on an equivalent molar basis. A Pu loading of 1 wt % in glass translated to ∼18 kg Pu per Defense Waste Processing Facility (DWPF) canister, or about 10X the current allowed limit per the Waste Acceptance Product Specifications (2500 g/m 3 of glass or about 1700 g/canister) and about 30X the current allowable concentration based on the fissile material concentration limit referenced in the Yucca Mountain Project License Application (897 g/m 3 3 of glass or about 600 g Pu/canister). Based on historical process throughput data, this level was considered to represent a reasonable upper bound for Pu loading based on the ability to provide Pu containing feed to the DWPF. The task elements included evaluating the distribution of Pu in the glass (e.g. homogeneity), evaluating crystallization within the glass, evaluating select glass properties (with surrogates), and evaluating durability using the Product Consistency Test -- Method A (PCT-A). The behavior of Pu in the melter was evaluated using paper studies and corresponding analyses of DWPF melter pour samples.The results of the testing indicated that at 1 wt % Pu in the glass, the Pu was homogeneously distributed and did not result in any formation of plutonium-containing crystalline phases as long as the glass was prepared under 'well-mixed' conditions. The incorporation of 1 wt

  3. Standard test method for measuring waste glass or glass ceramic durability by vapor hydration test

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    1.1 The vapor hydration test method can be used to study the corrosion of a waste forms such as glasses and glass ceramics upon exposure to water vapor at elevated temperatures. In addition, the alteration phases that form can be used as indicators of those phases that may form under repository conditions. These tests; which allow altering of glass at high surface area to solution volume ratio; provide useful information regarding the alteration phases that are formed, the disposition of radioactive and hazardous components, and the alteration kinetics under the specific test conditions. This information may be used in performance assessment (McGrail et al, 2002 (1) for example). 1.2 This test method must be performed in accordance with all quality assurance requirements for acceptance of the data. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practice...

  4. Strontium chloroapatite based glass-ceramics composites for nuclear waste immobilisation

    International Nuclear Information System (INIS)

    Jena, Hrudananda; Maji, Binoy Kumar; Asuvathraman, R.; Govindan Kutty, K.V.

    2013-01-01

    Apatites are naturally occurring minerals with a general formula of M 10 (PO 4 ) 6 X 2 , (M= Ca, Sr, Ba, X= OH, Cl, F) with a hexagonal crystal structure (S.G :P6 3 /m) and can accommodate alkaline earth and various other aliovalent cations and anions into its crystal structure. Apatites are also known to have high resistance to leaching of the constituent elements under geological conditions. It may not often be possible to immobilize the whole spectrum of the radioactive waste in a single phase M 10 (PO 4 ) 6 Cl 2 , then a combination of M-chloroapatite encapsulated in borosilicate glass (BSG) can immobilize most of the radwaste elements in the composite glass-ceramic matrix (glass bonded chloroapatite), thus utilizing the immobilizing efficiency of both the ceramic phase and glass. In the present study, the synthesis, characterization and thermo-physical property measurements of the Sr-chloroapatite (SrApCI) and some glass-bonded composites based on it have been investigated. The Sr-chloroapatite glass-ceramics were prepared by solid state reactions among stoichiometric concentrations of apatite forming reagents, 20 wt. % borosilicate glass (BSG), and known concentrations (10, 13 and 16 wt. %) of a simulated waste in chloride form. The products were characterized by XRD to confirm the formation of Sr 10 (PO 4 ) 6 Cl 2 and glass bonded-chloroapatite composites. The surface morphology and qualitative chemical composition of the powders were examined by SEM and EDX. Thermal expansion and glass transition temperature of the matrices were measured by dilatometry. Glass transition temperature of the glass-bonded composites was also examined by differential scanning calorimetry and differential thermal analysis. The 10-16 wt.% waste loaded matrices showed similar thermal expansion as that of SrApCI, indicating the thermal stability of the matrix to chloride waste immobilization. The glass transition temperature of the waste loaded matrices decreases on increasing the

  5. Glass as a waste form for the immobilization of plutonium

    International Nuclear Information System (INIS)

    Bates, J.K.; Ellison, A.J.G.; Emery, J.W.; Hoh, J.C.

    1995-01-01

    Several alternatives for disposal of surplus plutonium are being considered. One method is incorporating Pu into glass and in this paper we discuss the development and corrosion behavior of an alkali-tin-silicate glass and update results in testing Pu doped Defense Waste Processing Facility (DWPF) reference glasses. The alkali-tin-silicate glass was engineered to accommodate a high Pu loading and to be durable under conditions likely to accelerate glass reaction. The glass dissolves about 7 wt% Pu together with the neutron absorber Gd, and under test conditions expected to accelerate the glass reaction with water, is resistant to corrosion. The Pu and the Gd are released from the glass at nearly the same rate in static corrosion tests in water, and are not segregated into surface alteration phases when the glass is reacted in water vapor. Similar results for the behavior of Pu and Gd are found for the DWPF reference glasses, although the long-term rate of reaction for the reference glasses is more rapid than for the alkali-tin-silicate glass

  6. Optimization of waste loading in high-level glass in the presence of uncertainty

    International Nuclear Information System (INIS)

    Hoza, M.; Fann, G.I.; Hopkins, D.F.

    1995-02-01

    Hanford high-level liquid waste will be converted into a glass form for long-term storage. The glass must meet certain constraints on its composition and properties in order to have desired properties for processing (e.g., electrical conductivity, viscosity, and liquidus temperature) and acceptable durability for long-term storage. The Optimal Waste Loading (OWL) models, based on rigorous mathematical optimization techniques, have been developed to minimize the number of glass logs required and determine glass-former compositions that will produce a glass meeting all relevant constraints. There is considerable uncertainty in many of the models and data relevant to the formulation of high-level glass. In this paper, we discuss how we handle uncertainty in the glass property models and in the high-level waste composition to the vitrification process. Glass property constraints used in optimization are inequalities that relate glass property models obtained by regression analysis of experimental data to numerical limits on property values. Therefore, these constraints are subject to uncertainty. The sampling distributions of the regression models are used to describe the uncertainties associated with the constraints. The optimization then accounts for these uncertainties by requiring the constraints to be satisfied within specified confidence limits. The uncertainty in waste composition is handled using stochastic optimization. Given means and standard deviations of component masses in the high-level waste stream, distributions of possible values for each component are generated. A series of optimization runs is performed; the distribution of each waste component is sampled for each run. The resultant distribution of solutions is then statistically summarized. The ability of OWL models to handle these forms of uncertainty make them very useful tools in designing and evaluating high-level waste glasses formulations

  7. Preparation of durable hydrophobic cellulose fabric from water glass and mixed organosilanes

    Science.gov (United States)

    Shang, Song-Min; Li, Zhengxiong; Xing, Yanjun; Xin, John H.; Tao, Xiao-Ming

    2010-12-01

    Durable superhydrophobic cellulose fabric was prepared from water glass and n-octadecyltriethoxysilane (ODTES) with 3-glycidyloxypropyltrimethoxysilane (GPTMS) as crosslinker by sol-gel method. The result showed that the addition of GPTMS could result in a better fixation of silica coating from water glass on cellulose fabric. The silanization of hydrolyzed ODTES at different temperatures and times was studied and optimized. The results showed that silanization time was more important than temperature in forming durable hydrophobic surface. The durability of superhydrophobicity treatment was analyzed by XPS. As a result, the superhydrophobic cotton treated under the optimal condition still remained hydrophobic properties after 50 washing cycles.

  8. Physical properties and chemical durability of selected zirconia containing silicate glasses

    Directory of Open Access Journals (Sweden)

    ELEONÓRA GAŠPÁREKOVÁ

    2011-12-01

    Full Text Available Density, thermal expansion, glass transition temperature, refractive index, molar refractivity and chemical durability of five- and six-component glasses with as weighted composition xNa2O·(15-xK2O·yCaO∙(10-yZnO∙zZrO2∙(75-zSiO2 (x = 0, 7.5, 15; y = 0, 5, 10; z = 5, 7 were measured. The obtained experimental data were merged together with the previous results obtained for analogous glasses with lower zirconia content. The full set of glasses enabled the quantitative statistical estimation of possible mixed-oxide effects. The results of the multilinear regression analysis pointed out the ideal behavior of molar volume and molar refractivity. The strongest influence of mutual oxide interactions was found for chemical durability and glass transition temperature. The regression analysis of compositional dependence of metastable melt thermal expansion coefficient practically failed. The need of property-composition study based on the thermodynamic model was pointed out. Qualitatively the obtained results confirmed those previously obtained for the analogous glasses with zirconia content reaching up to 3 mol. %.

  9. Candidate Low-Temperature Glass Waste Forms for Technetium-99 Recovered from Hanford Effluent Management Facility Evaporator Concentrate

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Mei [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tang, Ming [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rim, Jung Ho [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chamberlin, Rebecca M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-07-24

    Alternative treatment and disposition options may exist for technetium-99 (99Tc) in secondary liquid waste from the Hanford Direct-Feed Low-Activity Waste (DFLAW) process. One approach includes development of an alternate glass waste form that is suitable for on-site disposition of technetium, including salts and other species recovered by ion exchange or precipitation from the EMF evaporator concentrate. By recovering the Tc content from the stream, and not recycling the treated concentrate, the DFLAW process can potentially be operated in a more efficient manner that lowers the cost to the Department of Energy. This report provides a survey of candidate glass formulations and glass-making processes that can potentially incorporate technetium at temperatures <700 °C to avoid volatilization. Three candidate technetium feed streams are considered: (1) dilute sodium pertechnetate loaded on a non-elutable ion exchange resin; (2) dilute sodium-bearing aqueous eluent from ion exchange recovery of pertechnetate, or (3) technetium(IV) oxide precipitate containing Sn and Cr solids in an aqueous slurry. From the technical literature, promising candidate glasses are identified based on their processing temperatures and chemical durability data. The suitability and technical risk of three low-temperature glass processing routes (vitrification, encapsulation by sintering into a glass composite material, and sol-gel chemical condensation) for the three waste streams was assessed, based on available low-temperature glass data. For a subset of candidate glasses, their long-term thermodynamic behavior with exposure to water and oxygen was modeled using Geochemist’s Workbench, with and without addition of reducing stannous ion. For further evaluation and development, encapsulation of precipitated TcO2/Sn/Cr in a glass composite material based on lead-free sealing glasses is recommended as a high priority. Vitrification of pertechnetate in aqueous anion exchange eluent solution

  10. INTERNATIONAL STUDY OF ALUMINUM IMPACTS ON CRYSTALLIZATION IN U.S. HIGH LEVEL WASTE GLASS

    International Nuclear Information System (INIS)

    Fox, K; David Peeler, D; Tommy Edwards, T; David Best, D; Irene Reamer, I; Phyllis Workman, P; James Marra, J

    2008-01-01

    The objective of this task was to develop glass formulations for (Department of Energy) DOE waste streams with high aluminum concentrations to avoid nepheline formation while maintaining or meeting waste loading and/or waste throughput expectations as well as satisfying critical process and product performance related constraints. Liquidus temperatures and crystallization behavior were carefully characterized to support model development for higher waste loading glasses. The experimental work, characterization, and data interpretation necessary to meet these objectives were performed among three partnering laboratories: the V.G. Khlopin Radium Institute (KRI), Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL). Projected glass compositional regions that bound anticipated Defense Waste Processing Facility (DWPF) and Hanford high level waste (HLW) glass regions of interest were developed and used to generate glass compositions of interest for meeting the objectives of this study. A thorough statistical analysis was employed to allow for a wide range of waste glass compositions to be examined while minimizing the number of glasses that had to be fabricated and characterized in the laboratory. The glass compositions were divided into two sets, with 45 in the test matrix investigated by the U.S. laboratories and 30 in the test matrix investigated by KRI. Fabrication and characterization of the US and KRI-series glasses were generally handled separately. This report focuses mainly on the US-series glasses. Glasses were fabricated and characterized by SRNL and PNNL. Crystalline phases were identified by X-ray diffraction (XRD) in the quenched and canister centerline cooled (CCC) glasses and were generally iron oxides and spinels, which are not expected to impact durability of the glass. Nepheline was detected in five of the glasses after the CCC heat treatment. Chemical composition measurements for each of the glasses were conducted

  11. INTERNATIONAL STUDY OF ALUMINUM IMPACTS ON CRYSTALLIZATION IN U.S. HIGH LEVEL WASTE GLASS

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K; David Peeler, D; Tommy Edwards, T; David Best, D; Irene Reamer, I; Phyllis Workman, P; James Marra, J

    2008-09-23

    The objective of this task was to develop glass formulations for (Department of Energy) DOE waste streams with high aluminum concentrations to avoid nepheline formation while maintaining or meeting waste loading and/or waste throughput expectations as well as satisfying critical process and product performance related constraints. Liquidus temperatures and crystallization behavior were carefully characterized to support model development for higher waste loading glasses. The experimental work, characterization, and data interpretation necessary to meet these objectives were performed among three partnering laboratories: the V.G. Khlopin Radium Institute (KRI), Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL). Projected glass compositional regions that bound anticipated Defense Waste Processing Facility (DWPF) and Hanford high level waste (HLW) glass regions of interest were developed and used to generate glass compositions of interest for meeting the objectives of this study. A thorough statistical analysis was employed to allow for a wide range of waste glass compositions to be examined while minimizing the number of glasses that had to be fabricated and characterized in the laboratory. The glass compositions were divided into two sets, with 45 in the test matrix investigated by the U.S. laboratories and 30 in the test matrix investigated by KRI. Fabrication and characterization of the US and KRI-series glasses were generally handled separately. This report focuses mainly on the US-series glasses. Glasses were fabricated and characterized by SRNL and PNNL. Crystalline phases were identified by X-ray diffraction (XRD) in the quenched and canister centerline cooled (CCC) glasses and were generally iron oxides and spinels, which are not expected to impact durability of the glass. Nepheline was detected in five of the glasses after the CCC heat treatment. Chemical composition measurements for each of the glasses were conducted

  12. Influence of curing conditions on durability of alkali-resistant glass ...

    Indian Academy of Sciences (India)

    Glass fibres in concrete material often increase the flexural strength. However, these fibres when in contact with cement are altered by alkali reactions due to the presence of portlandite. This study presents the results of investigation to show the effect of curing conditions on the durability of alkali-resistant glass fibres in ...

  13. Vitrification development for mixed wastes

    International Nuclear Information System (INIS)

    Merrill, R.; Whittington, K.; Peters, R.

    1995-02-01

    Vitrification is a promising approach to waste-form immobilization. It destroys hazardous organic compounds and produces a durable and highly stable glass. Vitrification tests were performed on three surrogate wastes during fiscal year 1994; 183-H Solar Evaporation Basin waste from Hanford, bottom ash from the Oak Ridge TSCA incinerator, and saltcrete from Rocky Flats. Preliminary glass development involved melting trials followed by visual homogeneity examination, short-duration leach tests on glass specimens, and long-term leach tests on selected glasses. Viscosity and electrical conductivity measurements were taken for the most durable glass formulations. Results for the saltcrete are presented in this paper and demonstrate the applicability of vitrification technology to this mixed waste

  14. Effects of heat treatment and formulation on the phase composition and chemical durability of the EBR-II ceramic waste form

    International Nuclear Information System (INIS)

    Ebert, W. E.; Dietz, N. L.; Janney, D. E.

    2006-01-01

    High-level radioactive waste salts generated during the electrometallurgical treatment of spent sodium-bonded nuclear fuel from the Experimental Breeder Reactor-II will be immobilized in a ceramic waste form (CWF). Tests are being conducted to evaluate the suitability of the CWF for disposal in the planned federal high-level radioactive waste repository at Yucca Mountain. In this report, the results of laboratory tests and analyses conducted to address product consistency and thermal stability issues called out in waste acceptance requirements are presented. The tests measure the impacts of (1) variations in the amounts of salt and binder glass used to make the CWF and (2) heat treatments on the phase composition and chemical durability of the waste form. A series of CWF materials was made to span the ranges of salt and glass contents that could be used during processing: between 5.0 and 15 mass% salt loaded into the zeolite (the nominal salt loading is 10.7%, and the process control range is 10.6 to 11.2 mass%), and between 20 and 30 mass% binder glass mixed with the salt-loaded zeolite (the nominal glass content is 25% and the process control range is 20 to 30 mass%). In another series of tests, samples of two CWF products made with the nominal salt and glass contents were reheated to measure the impact on the phase composition and durability: long-term heat treatments were conducted at 400 and 500 C for durations of 1 week, 4 weeks, 3 months, 6 months, and 1 year; short-term heat treatments were conducted at 600, 700, 800, and 850 C for durations of 4, 28, 52, and 100 hours. All of the CWF products that were made with different amounts of salt, zeolite, and glass and all of the heat-treated CWF samples were analyzed with powder X-ray diffraction to measure changes in phase compositions and subjected to 7-day product consistency tests to measure changes in the chemical durability. The salt loading had the greatest impact on phase composition and durability. A

  15. NMR characterization of simulated Hanford low-activity waste glasses and its use in understanding waste form chemical durability

    International Nuclear Information System (INIS)

    Darab, J.G.; Linehan, J.C.; McGrail, B.P.

    1999-01-01

    Magic Angle Spinning Nuclear Magnetic Resonance (MAS-NMR) spectroscopy has been used to characterize the structural and chemical environments of B, Al, and Si in model Hanford low-activity waste glasses. The average 29 Si NMR peak position was found to systematically change with changing glass composition and structure. From an understanding of the structural roles of Al and B obtained from MAS-NMR experiments, the authors first developed a model that reliably predicts the distribution of structural units and the average 29 Si chemical shift value, δ, based purely on glass composition. A product consistency test (PCT) was used to determine the normalized elemental release (NL) from the prepared glasses. Comparison of the NMR and PCT data obtained from sodium boro-aluminosilicate glasses indicates that a rudimentary exponential relationship exists between the 29 Si chemical shift value, and the boron NL value

  16. Glass containing radioactive nuclear waste

    International Nuclear Information System (INIS)

    Boatner, L.A.; Sales, B.C.

    1985-01-01

    Lead-iron phosphate glasses containing a high level of Fe 2 O 3 for use as a storage medium for high-level-radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90 C, with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe 2 O 3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800 C, since they exhibit very low melt viscosities in the 800 to 1050 C temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550 C and are not adversely affected by large doses of gamma radiation in H 2 O at 135 C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear waste forms. (author)

  17. Performance of borosilicate glass, Synroc and spent fuel as nuclear waste forms

    International Nuclear Information System (INIS)

    Lutze, W.; Grambow, B.; Ewing, R.C.

    1990-01-01

    Presently, there are three prominent waste forms under consideration for the disposal of high-level waste: Borosilicate glass and Synroc for high-level radioactive waste from fuel reprocessing and spent fuel as the waste form for non-reprocessed fuel. Using the present experimental data base, one may compare the performance of these three waste forms under repository relevant conditions. In low water flow regimes and at temperatures less than 100 degree C, the fractional release rates of all three waste forms are low, on the order of 10-7/d or less and may decrease with time. Under these conditions the three waste forms behave similarly. At elevated temperatures or in high flow regimes, the durability of borosilicate glass will be much less than that of Synroc, and thus, for certain disposal schemes (e.g., deep burial) Synroc is preferable. All predictions of the long-term behavior are based on the extrapolation of short term experimental data, we point out that appropriate and useful natural analogues are available for each of these waste forms and should be used in the performance assessment of each waste form's long-term behavior. 14 refs

  18. Elucidating the effects of solar panel waste glass substitution on the physical and mechanical characteristics of clay bricks.

    Science.gov (United States)

    Lin, Kae-Long; Huang, Long-Sheng; Shie, Je-Lueng; Cheng, Ching-Jung; Lee, Ching-Hwa; Chang, Tien-Chin

    2013-01-01

    This study deals with the effect of solar panel waste glass on fired clay bricks. Brick samples were heated to temperatures which varied from 700-1000 degrees C for 6 h, with a heating rate of 10 degrees C min(-1). The material properties of the resultant material were then determined, including speciation variation, loss on ignition, shrinkage, bulk density, 24-h absorption rate, compressive strength and salt crystallization. The results indicate that increasing the amount of solar panel waste glass resulted in a decrease in the water absorption rate and an increase in the compressive strength of the solar panel waste glass bricks. The 24-h absorption rate and compressive strength of the solar panel waste glass brick made from samples containing 30% solar panel waste glass sintered at 1000 degrees C all met the Chinese National Standard (CNS) building requirements for first-class brick (compressive strengths and water absorption of the bricks were 300 kg cm(-2) and 10% of the brick, respectively). The addition of solar panel waste glass to the mixture reduced the degree of firing shrinkage. The salt crystallization test and wet-dry tests showed that the addition of solar panel waste glass had highly beneficial effects in that it increased the durability of the bricks. This indicates that solar panel waste glass is indeed suitable for the partial replacement of clay in bricks.

  19. Radiation effects in silicate glasses

    International Nuclear Information System (INIS)

    Bibler, N.E.; Howitt, D.G.

    1988-01-01

    The study of radiation effects in complex silicate glasses has received renewed attention because of their use in special applications such as high level nuclear waste immobilization and fiber optics. Radiation changes the properties of these glasses by altering their electronic and atomic configurations. These alterations or defects may cause dilatations or microscopic phase changes along with absorption centers that limit the optical application of the glasses. Atomic displacements induced in the already disordered structure of the glasses may affect their use where heavy irradiating particles such as alpha particles, alpha recoils, fission fragments, or accelerated ions are present. Large changes (up to 1%) in density may result. In some cases the radiation damage may be severe enough to affect the durability of the glass in aqueous solutions. In the paper, the authors review the literature concerning radiation effects on density, durability, stored energy, microstructure and optical properties of silicate glasses. Both simple glasses and complex glasses used for immobilization of nuclear waste are considered

  20. Leaching and mechanical properties of cabal glasses developed as matrices for immobilization high-level wastes

    International Nuclear Information System (INIS)

    Ezz-Eldin, F.M.

    2001-01-01

    This paper discusses the leaching behavior of simulated high-level-waste cabal glass (CaO-B 2 O 3 -Al 2 O 3 ) as a bulk specimen. During leach tests, the glass is immersed in either deionized water or in groundwater for up to 57 days at 70 deg. C. Based on the results, mechanisms observed with the leaching of the glass in deionized water or groundwater are discussed. Three factors, i.e., time of immersion, type of leaching solution and irradiation effect, are extensively studied. The corrosion was found to be linear with time in the limit of investigation (1-57 days) but with different rates depending on the type of solution and glass composition. Effects of γ-irradiation on the glass together with groundwater were found to decrease the glass durability. The evolution of the damage on mechanical and physical properties of the glass before and after leaching or irradiation was also discussed. The addition of waste oxide changes the properties of the glass matrix, so the influence of the guest oxides on the properties of host materials is also discussed

  1. Durability of adhesive glass-metal connections for structural applications

    NARCIS (Netherlands)

    Van Lancker, B.; Dispersyn, J.; De Corte, W.; Belis, J.

    2016-01-01

    The use of adhesive bonds for structural glass-metal connections in the building envelope has increased in recent years. Despite the multiple advantages compared to more traditional bolted connections, long-term behaviour and durability of the adhesives have to be investigated accurately. Because,

  2. Study on dissolution behavior of molten solidified waste

    International Nuclear Information System (INIS)

    Mizuno, Tsuyoshi; Maeda, Toshikatsu

    2005-01-01

    Radioactive molten solidified waste (slag) has been generated by melting non-metallic low-level radioactive wastes (LLW). Slag is expected to immobilize radionuclides in the waste repository. The chemical durability of slag is an important factor for the safety assessment of the disposal in that the durability provides the source term in the assessment. Since a chemical characteristic of slag is similar to that of glass, the general information on the chemical durability of slag might be provided from previous studies on nuclear waste glass. We have investigated effects of chemical compositions of slag and alkaline environments of repository on the chemical durability of slag. (author)

  3. Control of radioactive waste-glass melters

    International Nuclear Information System (INIS)

    Bickford, D.F.; Smith, P.K.; Hrma, P.; Bowan, B.W.

    1987-01-01

    Radioactive waste-glass melters require physical control limits and redox control of glass to assure continuous operation, and maximize production rates. Typical waste-glass melter operating conditions, and waste-glass chemical reaction paths are discussed. Glass composition, batching and melter temperature control are used to avoid the information of phases which are disruptive to melting or reduce melter life. The necessity and probable limitations of control for electric melters with complex waste feed compositions are discussed. Preliminary control limits, their bases, and alternative control methods are described for use in the Defense Waste Processing Facility (DWPF) at the US Department of Energy's Savannah River Plant (SRP), and at the West Valley Demonstration Project (WVDP). Slurries of simulated high level radioactive waste and ground glass frit or glass formers have been isothermally reacted and analyzed to identify the sequence of the major chemical reactions in waste vitrification, and their effect on waste-glass production rates. Relatively high melting rates of waste batches containing mixtures of reducing agents (formic acid, sucrose) and nitrates are attributable to exothermic reactions which occur at critical stages in the vitrification process. The effect of foaming on waste glass production rates is analyzed, and limits defined for existing waste-glass melters, based upon measurable thermophysical properties. Through balancing the high nitrate wastes of the WVDP with reducing agents, the high glass melting rates and sustained melting without foaming required for successful WVDP operations have been demonstrated. 65 refs., 4 figs., 15 tabs

  4. Optical Basicity and Nepheline Crystallization in High Alumina Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, Carmen P.; McCloy, John S.; Schweiger, M. J.; Crum, Jarrod V.; Winschell, Abigail E.

    2011-02-25

    The purpose of this study was to find compositions that increase waste loading of high-alumina wastes beyond what is currently acceptable while avoiding crystallization of nepheline (NaAlSiO4) on slow cooling. Nepheline crystallization has been shown to have a large impact on the chemical durability of high-level waste glasses. It was hypothesized that there would be some composition regions where high-alumina would not result in nepheline crystal production, compositions not currently allowed by the nepheline discriminator. Optical basicity (OB) and the nepheline discriminator (ND) are two ways of describing a given complex glass composition. This report presents the theoretical and experimental basis for these models. They are being studied together in a quadrant system as metrics to explore nepheline crystallization and chemical durability as a function of waste glass composition. These metrics were calculated for glasses with existing data and also for theoretical glasses to explore nepheline formation in Quadrant IV (passes OB metric but fails ND metric), where glasses are presumed to have good chemical durability. Several of these compositions were chosen, and glasses were made to fill poorly represented regions in Quadrant IV. To evaluate nepheline formation and chemical durability of these glasses, quantitative X-ray diffraction (XRD) analysis and the Product Consistency Test were conducted. A large amount of quantitative XRD data is collected here, both from new glasses and from glasses of previous studies that had not previously performed quantitative XRD on the phase assemblage. Appendix A critically discusses a large dataset to be considered for future quantitative studies on nepheline formation in glass. Appendix B provides a theoretical justification for choice of the oxide coefficients used to compute the OB criterion for nepheline formation.

  5. Waste form development for a DC arc furnace

    Energy Technology Data Exchange (ETDEWEB)

    Feng, X.; Bloomer, P.E.; Chantaraprachoom, N.; Gong, M.; Lamar, D.A.

    1996-09-01

    A laboratory crucible study was conducted to develop waste forms to treat nonradioactive simulated {sup 238}Pu heterogeneous debris waste from Savannah River, metal waste from the Idaho National Engineering Laboratory (INEL), and nominal waste also from INEL using DC arc melting. The preliminary results showed that the different waste form compositions had vastly different responses for each processing effect. The reducing condition of DC arc melting had no significant effects on the durability of some waste forms while it decreased the waste form durability from 300 to 700% for other waste forms, which resulted in the failure of some TCLP tests. The right formulations of waste can benefit from devitrification and showed an increase in durability by 40%. Some formulations showed no devitrification effects while others decreased durability by 200%. Increased waste loading also affected waste form behavior, decreasing durability for one waste, increasing durability by 240% for another, and showing no effect for the third waste. All of these responses to the processing and composition variations were dictated by the fundamental glass chemistry and can be adjusted to achieve maximal waste loading, acceptable durability, and desired processing characteristics if each waste formulation is designed for the result according to the glass chemistry.

  6. Low-Li2O Frits: Selecting Glasses that Support the Melt Rate Studies and Challenge the Current Durability Model

    International Nuclear Information System (INIS)

    Peeler, D. K.; Edwards, T. B.

    2005-01-01

    During the progressive development of the cold cap model (as it applies to a potential melt rate predictive tool), the formation of an Al-Li-silicate phase was identified as an intermediate reaction phase that could possibly hinder melt rate for SB4. To test this theory, six glasses were designed (using Frit 320's composition as the baseline) to maintain a constant 20 wt% sum of alkali content (in frit) by varying Na 2 O to Li 2 O ratios. The Li 2 O concentration ranged from 8 wt% down to 0% in either 2% or 1% increments with the differences being accounted for by an increase in Na 2 O concentration. Although the primary objective of the ''lower Li 2 O'' frits was to evaluate the potential for melt rate improvements, assessments of durability (as measured by the Product Consistency Test (PCT)) were also performed. The results suggest that durable glasses can be produced with these ''lower Li 2 O'' frits should it be necessary to pursue this option for improving melt rate. In addition to the series of glasses to support melt rate assessments, a series of frits were also developed to challenge the current durability model based on the limits proposed by Edwards et al. (2004). Although the ''new'' limits allow access into compositional regions of interest (i.e., higher alkali systems) which can improve melt rate and/or waste loading, there may still be ''additional'' conservatism. In this report, two series of glasses were developed to challenge the ''new'' durability limits for the SB4 system. In the first series, the total alkali of the Frit 320-based glasses (designed to support the melt rate program) was increased from 20 wt% to 21 wt% (in the frit), but the series also evaluated the possible impact of various Na 2 O and Li 2 O mass ratio differences. The second series pushed the alkali limit in the frit even further with frits containing either 22 or 24 wt% total alkali as well as various Na 2 O and Li 2 O mass ratios. The results of the PCT evaluation indicated

  7. GLASS COMPOSITION-TCLP RESPONSE MODEL FOR WASTE GLASSES

    International Nuclear Information System (INIS)

    Kim, Dong-Sang; Vienna, John D.

    2004-01-01

    A first-order property model for normalized Toxicity Characteristic Leaching Procedure (TCLP) release as a function of glass composition was developed using data collected from various studies. The normalized boron release is used to estimate the release of toxic elements based on the observation that the boron release represents the conservative release for those constituents of interest. The current TCLP model has two targeted application areas: (1) delisting of waste-glass product as radioactive (not mixed) waste and (2) designating the glass wastes generated from waste-glass research activities as hazardous or non-hazardous. This paper describes the data collection and model development for TCLP releases and discusses the issues related to the application of the model

  8. Chemical and mechanical performance properties for various final waste forms -- PSPI scoping study

    International Nuclear Information System (INIS)

    Farnsworth, R.K.; Larsen, E.D.; Sears, J.W.; Eddy, T.L.; Anderson, G.L.

    1996-09-01

    The US DOE is obtaining data on the performance properties of the various final waste forms that may be chosen as primary treatment products for the alpha-contaminated low-level and transuranic waste at the INEL's Transuranic Storage Area. This report collects and compares selected properties that are key indicators of mechanical and chemical durability for Portland cement concrete, concrete formed under elevated temperature and pressure, sulfur polymer cement, borosilicate glass, and various forms of alumino-silicate glass, including in situ vitrification glass and various compositions of iron-enriched basalt (IEB) and iron-enriched basalt IV (IEB4). Compressive strength and impact resistance properties were used as performance indicators in comparative evaluation of the mechanical durability of each waste form, while various leachability data were used in comparative evaluation of each waste form's chemical durability. The vitrified waste forms were generally more durable than the non-vitrified waste forms, with the iron-enriched alumino-silicate glasses and glass/ceramics exhibiting the most favorable chemical and mechanical durabilities. It appears that the addition of zirconia and titania to IEB (forming IEB4) increases the leach resistance of the lanthanides. The large compositional ranges for IEB and IEB4 more easily accommodate the compositions of the waste stored at the INEL than does the composition of borosilicate glass. It appears, however, that the large potential variation in IEB and IEB4 compositions resulting from differing waste feed compositions can impact waste form durability. Further work is needed to determine the range of waste stream feed compositions and rates of waste form cooling that will result in acceptable and optimized IEB or IEB4 waste form performance. 43 refs

  9. Standard test method for determining liquidus temperature of immobilized waste glasses and simulated waste glasses

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    1.1 These practices cover procedures for determining the liquidus temperature (TL) of nuclear waste, mixed nuclear waste, simulated nuclear waste, or hazardous waste glass in the temperature range from 600°C to 1600°C. This method differs from Practice C829 in that it employs additional methods to determine TL. TL is useful in waste glass plant operation, glass formulation, and melter design to determine the minimum temperature that must be maintained in a waste glass melt to make sure that crystallization does not occur or is below a particular constraint, for example, 1 volume % crystallinity or T1%. As of now, many institutions studying waste and simulated waste vitrification are not in agreement regarding this constraint (1). 1.2 Three methods are included, differing in (1) the type of equipment available to the analyst (that is, type of furnace and characterization equipment), (2) the quantity of glass available to the analyst, (3) the precision and accuracy desired for the measurement, and (4) candi...

  10. Mechanical properties of nuclear waste glasses

    International Nuclear Information System (INIS)

    Connelly, A.J.; Hand, R.J.; Bingham, P.A.; Hyatt, N.C.

    2011-01-01

    The mechanical properties of nuclear waste glasses are important as they will determine the degree of cracking that may occur either on cooling or following a handling accident. Recent interest in the vitrification of intermediate level radioactive waste (ILW) as well as high level radioactive waste (HLW) has led to the development of new waste glass compositions that have not previously been characterised. Therefore the mechanical properties, including Young's modulus, Poisson's ratio, hardness, indentation fracture toughness and brittleness of a series of glasses designed to safely incorporate wet ILW have been investigated. The results are presented and compared with the equivalent properties of an inactive simulant of the current UK HLW glass and other nuclear waste glasses from the literature. The higher density glasses tend to have slightly lower hardness and indentation fracture toughness values and slightly higher brittleness values, however, it is shown that the variations in mechanical properties between these different glasses are limited, are well within the range of published values for nuclear waste glasses, and that the surveyed data for all radioactive waste glasses fall within relatively narrow range.

  11. Alteration of medieval stained-glasses. Contribution to the long-term behaviour of vitrified wastes

    International Nuclear Information System (INIS)

    Sterpenich, J.

    1998-01-01

    In this work, the behaviour of glasses during alteration have been studied in two different ways: 1)study of the alteration of medieval stained-glasses 2)experimental leaching of modelled glasses. Medieval stained-glasses have a silico-calcic and alkaline composition. It appears three different alteration modes for these glasses: 1)by condensation waters 2)by atmospheric agents 3)by porosity waters and humic acids. A chemical study of the altered areas has allowed to understand the alteration behaviour of a lot of elements: in particular transition elements, heavy metals and some rare earths. On the other hand, two vitrified wastes and a glass having the same composition of the potassic medieval stained-glasses have been leached in a static mode (pH=1 to 10, T=20 to 80 degrees Celsius, T=12 hours to 6 months). These experiments have revealed that the alteration mechanisms depend on the pH of the solution and on the chemical composition of the glass. An increasing durability of glasses in terms of the global polymerization degree has been revealed too. At last, the behaviours of glasses during alteration, observed with natural and experimental conditions, show that it is necessary to study natural analogous for predicting the long-term behaviour of vitrified wastes. (O.M.)

  12. Current Understanding and Remaining Challenges in Modeling Long-Term Degradation of Borosilicate Nuclear Waste Glasses

    International Nuclear Information System (INIS)

    Vienna, John D.; Ryan, Joseph V.; Gin, Stephane; Inagaki, Yaohiro

    2013-01-01

    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

  13. PNL vitrification technology development project glass formulation strategy for LLW vitrification

    International Nuclear Information System (INIS)

    Kim, D.; Hrma, P.R.; Westsik, J.H. Jr.

    1996-03-01

    This Glass Formulation Strategy describes development approaches to optimize glass compositions for Hanford's low-level waste vitrification between now and the projected low-level waste facility start-up in 2005. The objectives of the glass formulation task are to develop optimized glass compositions with satisfactory long-term durability, acceptable processing characteristics, adequate flexibility to handle waste variations, maximize waste loading to practical limits, and to develop methodology to respond to further waste variations

  14. Achievement report for fiscal 2000 on research and development of high level waste glass utilization system of CO2 emission suppression type; 2000 nendo CO2 haishutsu yokuseigata hai glass kodo riyo system no kenkyu kaihatsu seika hokokusho (kokaiyo)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    With an objective to establish a waste glass recycling system of low cost and requiring less CO2 emission amount and energy consumption, research and development has been made on classification of waste glasses into particle composition that is required in regenerated commercial products. This paper summarizes the achievements in fiscal 2000. In the research of a waste glass reusing and supplying system designed by using LCA, discussions were given on items required in circulating and recycling waste glass resources, and quality control on raw materials and products. Evaluations of product quality control items were made on crystallized glass, sintered glasswool, automotive window glass, electric bulbs, fluorescent lamp glass, and quartz glass. Utilization tests were carried out for Mashiko porcelain china clay with an intention of expanding the application of waste glass, whereas the relationship between waste glass addition amount and optimal sintering temperature range was verified, disclosing that the limit of the waste glass addition is 10%. In the research on multi-functional hybrid materials, discussions were given on light-weight tiles and water permeating blocks with regard to the manufacturing technology, facility specifications, product quality, effects of the functions, and durability. (NEDO)

  15. Product consistency testing of three reference glasses in stainless steel and perfluoroalkoxy resin vessels

    International Nuclear Information System (INIS)

    Olson, K.M.; Smith, G.L.; Marschman, S.C.

    1995-03-01

    Because of their chemical durability, silicate glasses have been proposed and researched since the mid-1950s as a medium for incorporating high-level radioactive waste (HLW) generated from processing of nuclear materials. A number of different waste forms were evaluated and ranked in the early 1980s; durability (leach resistance) was the highest weighted factor. Borosilicate glass was rated the best waste form available for incorporation of HLW. Four different types of vessels and three different glasses were used to study the possible effect of vessel composition on durability test results from the Production Consistency Test (PCT). The vessels were 45-m 304 stainless steel vessels, 150-m 304 L stainless steel vessels, and 60-m perfluoroalkoxy (PFA) fluoropolymer resin vessels. The three glasses were the Environmental Assessment glass manufactured by Corning Incorporated and supplied by Westinghouse Savannah River company, and West Valley Nuclear Services reference glasses 5 and 6, manufactured and supplied by Catholic University of America. Within experimental error, no differences were found in durability test results using the 3 different glasses in the 304L stainless steel or PFA fluoropolymer resin vessels over the seven-day test period

  16. Sodalite as a vehicle to increase Re retention in waste glass simulant during vitrification

    Energy Technology Data Exchange (ETDEWEB)

    Luksic, Steven A., E-mail: steven.luksic@pnnl.gov; Riley, Brian J.; Parker, Kent E.; Hrma, Pavel

    2016-10-15

    Technetium (Tc) retention during Hanford waste vitrification can be increased if the volatility can be controlled. Incorporating Tc into a thermally stable mineral phase, such as sodalite, is one way to achieve increased retention. Here, rhenium (Re)-bearing sodalite was tested as a vehicle to transport perrhenate (ReO{sub 4}{sup −}), a nonradioactive surrogate for pertechnetate (TcO{sub 4}{sup −}), into high-level (HLW) and low-activity waste (LAW) glass simulants. After melting HLW and LAW simulant feeds, the retention of Re in the glass was measured and compared with the Re retention in glass prepared from a feed containing Re{sub 2}O{sub 7}. Phase analysis of sodalite in both these glasses across a profile of temperatures describes the durability of Re-sodalite during the feed-to-glass transition. The use of Re sodalite improved the Re retention by 21% for HLW glass and 85% for LAW glass, demonstrating the potential improvement in Tc-retention if TcO{sub 4}{sup −} were to be encapsulated in a Tc-sodalite prior to vitrification. - Highlights: • Re retention is improved by incorporation into sodalite structure. • LAW-type glass shows lower retention but larger improvement with Re-sodalite. • Sodalite is stable to higher temperatures in high-alumina glass melts.

  17. The long-term acceleration of waste glass corrosion: A preliminary review

    International Nuclear Information System (INIS)

    Kielpinski, A.L.

    1995-07-01

    Whereas a prior conception of glass dissolution assumed a relatively rapid initial dissolution which then slowed to a smaller, fairly constant longer-term rate, some recent work suggests that these two stages are followed by a third phase of dissolution, in which the dissolution rate is accelerated with respect to what had previously been thought of as the final long-term rate. The goals of the present study are to compile experimental data which may have a bearing on this phenomena, and to provide an initial assessment of these data. The Savannah River Technology Center (SRTC) is contracted to develop glass formulation models for vitrification of Hanford low-level waste (LLW), in support of the Hanford Tank Waste Remediation System Technology Development Program. The phenomenon of an increase in corrosion rate, following a period characterized by a low corrosion rate, has been observed by a number of researchers on a number of waste glass compositions. Despite inherent ambiguities arising from SA/V (glass surface area to solution volume ratio) and other effects, valid comparisons can be made in which accelerated corrosion was observed in one test, but not in another. Some glass compositions do not appear to attain a plateau region; it may be that the observation of continued, non-negligible corrosion in these glasses represents a passage from the initial rate to the accelerated rate. The long-term corrosion is a function of the interaction between the glass and its environment, including the leaching solution and the surrounding materials. Reaction path modeling and stability field considerations have been used with some success to predict the changes in corrosion rate over time, due to these interactions. The accelerated corrosion phenomenon highlights the need for such integrated corrosion modeling and the scenario-specific nature of a particular glass composition's durability

  18. Technetium Incorporation in Glass for the Hanford Tank Waste Treatment and Immobilization Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Kim, Dong Sang

    2015-01-14

    A priority of the United States Department of Energy (U.S. DOE) is to dispose of nuclear wastes accumulated in 177 underground tanks at the Hanford Nuclear Reservation in eastern Washington State. These nuclear wastes date from the Manhattan Project of World War II and from plutonium production during the Cold War. The DOE plans to separate high-level radioactive wastes from low activity wastes and to treat each of the waste streams by vitrification (immobilization of the nuclides in glass) for disposal. The immobilized low-activity waste will be disposed of here at Hanford and the immobilized high-level waste at the national geologic repository. Included in the inventory of highly radioactive wastes is large volumes of 99Tc (~9 × 10E2 TBq or ~2.5 × 104 Ci or ~1500 kg). A problem facing safe disposal of Tc-bearing wastes is the processing of waste feed into in a chemically durable waste form. Technetium incorporates poorly into silicate glass in traditional glass melting. It readily evaporates during melting of glass feeds and out of the molten glass, leading to a spectrum of high-to-low retention (ca. 20 to 80%) in the cooled glass product. DOE-ORP currently has a program at Pacific Northwest National Laboratory (PNNL), in the Department of Materials Science and Engineering at Rutgers University and in the School of Mechanical and Materials Engineering at Washington State University that seeks to understand aspects of Tc retention by means of studying Tc partitioning, molten salt formation, volatilization pathways, and cold cap chemistry. Another problem involves the stability of Tc in glass in both the national geologic repository and on-site disposal after it has been immobilized. The major environmental concern with 99Tc is its high mobility in addition to a long half-life (2.1×105 yrs). The pertechnetate ion (TcO4-) is highly soluble in water and does not adsorb well onto the surface of minerals and so migrates nearly at the same velocity as groundwater

  19. Defining the Glass Composition Limits for SRS Contaminated Soils

    International Nuclear Information System (INIS)

    Cicero, C.A.; Bickford, D.F.; Crews, W.O.

    1995-01-01

    Contaminated soil resulting from the excavation, repair, and decommissioning of facilities located at the Savannah River Site (SRS) is currently being disposed of by shallow land burial or is being stored when considered only hazardous. Vitrification of this waste is being investigated, since it will bind the hazardous and radioactive species in a stable and durable glass matrix, which will reduce the risk of ground water contamination. However, the composition limits for producing durable glass have to be determined before the technology can be applied. Glass compositions, consisting of SRS soil and glass forming additives, were tested on a crucible-scale in three ternary phase systems. Nine different glass compositions were produced, with waste loadings ranging from 43 to 58 weight percent. These were characterized using varoius chemical methods and tested for durability in both alkaline and acidic environments. All nine performed well in alkaline environments, but only three met the strictest criteria for the acidic environment tests. Although the glasses did not meet all of the limits for the acidic tests, the test was performed on very conservative size samples, so the results were also conservative. Therefore, enough evidence was found to provide proof that SRS soil can be vitrified in a durable glass matrix

  20. Waste glass melting stages

    International Nuclear Information System (INIS)

    Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

    1993-04-01

    Three different simulated nuclear waste glass feeds, consisting of dried waste and glass frit, were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600 degrees C--1000 degrees C. Simulated melter feeds from the Hanford Waste Vitrification Plant (HWVP), the Defense Waste Processing Facility (DWPF), and Kernforschungszentrum Karlsruhe (KfK) in Germany were used. The samples were thin-sectioned and examined by optical microscopy to investigate the stages of the conversion from feed to glass. Various phenomena were seen, such as frit softening, bubble formation, foaming, bubble motion and removal, convective mixing, and homogenization. Behavior of different feeds was similar, although the degree of gas generation and melt homogenization varied

  1. Durability-Based Design Criteria for a Chopped-Glass-Fiber Automotive Structural Composite; TOPICAL

    International Nuclear Information System (INIS)

    Battiste, R.L.; Corum, J.M.; Ren, W.; Ruggles, M.B.

    1999-01-01

    This report provides recommended durability-based design criteria for a chopped-glass-fiber reinforced polymeric composite for automotive structural applications. The criteria closely follow the framework of an earlier criteria document for a continuous-strand-mat (CSM) glass-fiber reference composite. Together these design criteria demonstrate a framework that can be adapted for future random-glass-fiber composites for automotive structural applications

  2. Minimum additive waste stabilization using vitreous ceramics. Progress report, October 1994--September 1995

    International Nuclear Information System (INIS)

    Feng, X.; Hahn, W.K.; Gong, M.; Gong, W.; Wang, L.; Ewing, R.C.

    1995-01-01

    Vitreous ceramic waste forms are being developed at Pacific Northwest Laboratory to complement glass waste forms in implementing the Minimum Additive Waste Stabilization (MAWS) Program to support the US Department of Energy's environmental restoration efforts. These vitreous ceramics are composed of various metal-oxide crystalline phases embedded in a silicate-glass phase. This work extends the success of vitreous ceramic waste forms to treat wastes with both high metal and high alkali contents. Two successful approaches are discussed: developing high-durability alkali-binding crystals in a durable glassy matrix, and developing water-soluble crystals in a durable and continuous glassy matrix. Nepheline-vitreous ceramics were demonstrated for the immobilization of high-alkali wastes with alkali contents up to 21 wt%. The chemical durability of the nepheline-vitreous ceramics is better than the corresponding glasses, especially in over longer times. Vitreous ceramics with Cs 2 O loading up to 35.4 wt% have been developed. Vitreous ceramic waste forms were developed from 90 and 100% Oak Ridge National Laboratory K-25 pond sludge. Heat treatment resulted in targeted crystal formation of spinels, potassium feldspar, and Ca-P phases. The K-25 pond sludge vitreous ceramics were up to 42 times more durable than high-level environmental assessments (EA) glass. The toxicity characteristics leach procedure (TCLP) concentration of LVC-6 is at least 2,000 times lower than US Environmental Protection Agency limits. Idaho Chemical Process Plant (ICPP) calcined wastes were immobilized into vitreous ceramics with calcine loading up to 88%. These ICPP-vitreous ceramics were more durable than the EA glass by factors of 5 to 30. Vitreous ceramic waste forms are being developed to complement, not to replace, glass waste forms

  3. An investigation of waste glass-based geopolymers supplemented with alumina

    Science.gov (United States)

    Christiansen, Mary U.

    An increased consideration of sustainability throughout society has resulted in a surge of research investigating sustainable alternatives to existing construction materials. A new binder system, called a geopolymer, is being investigated to supplement ordinary portland cement (OPC) concrete, which has come under scrutiny because of the CO2 emissions inherent in its production. Geopolymers are produced from the alkali activation of a powdered aluminosilicate source by an alkaline solution, which results in a dense three-dimensional matrix of tetrahedrally linked aluminosilicates. Geopolymers have shown great potential as a building construction material, offering similar mechanical and durability properties to OPC. Additionally, geopolymers have the added value of a considerably smaller carbon footprint than OPC. This research considered the compressive strength, microstructure and composition of geopolymers made from two types of waste glass with varying aluminum contents. Waste glass shows great potential for mainstream use in geopolymers due to its chemical and physical homogeneity as well as its high content of amorphous silica, which could eliminate the need for sodium silicate. However, the lack of aluminum is thought to negatively affect the mechanical performance and alkali stability of the geopolymer system. 39 Mortars were designed using various combinations of glass and metakaolin or fly ash to supplement the aluminum in the system. Mortar made from the high-Al glass (12% Al2O3) reached over 10,000 psi at six months. Mortar made from the low-Al glass (use in geopolymers, when care is given to consider the compositional and physical properties of the glass in mixture design.

  4. Nuclear waste glass corrosion mechanisms

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1987-04-01

    Dissolution of nuclear waste glass occurs by corrosion mechanisms similar to those of other solids, e.g., metallurgical and mineralogic systems. Metallurgical phenomena such as active corrosion, passivation and immunity have been observed to be a function of the glass composition and the solution pH. Hydration thermodynamics was used to quantify the role of glass composition and its effect on the solution pH during dissolution. A wide compositional range of natural, lunar, medieval, and nuclear waste glasses, as well as some glass-ceramics were investigated. The factors observed to affect dissolution in deionized water are pertinent to the dissolution of glass in natural environments such as the groundwaters anticipated to interact with nuclear waste glass in a geologic repository. The effects of imposed pH and oxidation potential (Eh) conditions existing in natural environments on glass dissolution is described in the context of Pourbaix diagrams, pH potential diagrams, for glass

  5. Review of glass ceramic waste forms

    International Nuclear Information System (INIS)

    Rusin, J.M.

    1981-01-01

    Glass ceramics are being considered for the immobilization of nuclear wastes to obtain a waste form with improved properties relative to glasses. Improved impact resistance, decreased thermal expansion, and increased leach resistance are possible. In addition to improved properties, the spontaneous devitrification exhibited in some waste-containing glasses can be avoided by the controlled crystallization after melting in the glass-ceramic process. The majority of the glass-ceramic development for nuclear wastes has been conducted at the Hahn-Meitner Institute (HMI) in Germany. Two of their products, a celsian-based (BaAl 3 Si 2 O 8 ) and a fresnoite-based (Ba 2 TiSi 2 O 8 ) glass ceramic, have been studied at Pacific Northwest Laboratory (PNL). A basalt-based glass ceramic primarily containing diopsidic augite (CaMgSi 2 O 6 ) has been developed at PNL. This glass ceramic is of interest since it would be in near equilibrium with a basalt repository. Studies at the Power Reactor and Nuclear Fuel Development Corporation (PNC) in Japan have favored a glass-ceramic product based upon diopside (CaMgSi 2 O 6 ). Compositions, processing conditions, and product characterization of typical commercial and nuclear waste glass ceramics are discussed. In general, glass-ceramic waste forms can offer improved strength and decreased thermal expansion. Due to typcially large residual glass phases of up to 50%, there may be little improvement in leach resistance

  6. Product consistency leach tests of Savannah River Site radioactive waste glasses

    International Nuclear Information System (INIS)

    Bibler, N.E.; Bates, J.K.

    1990-01-01

    The product consistency test (PCT) is a glass leach test developed at the Savannah River Site (SRS) to confirm the durability of radioactive nuclear waste glasses that will be produced in the Defense Waste Processing Facility. The PCT is a seven day, crushed glass leach test in deionized water at 90C. Final leachates are filtered and acidified prior to analysis. To demonstrate the reproducibility of the PCT when performed remotely, SRS and Argonne National Laboratory have performed the PCT on samples of two radioactive glasses. The tests were also performed to compare the releases of the radionuclides with the major nonradioactive glass components and to determine if radiation from the glass was affecting the results of the PCT. The test was performed in triplicate at each laboratory. For the major soluble elements, B, Li, Na, and Si, in the glass, each investigator obtained relative precisions in the range 2-5% in the triplicate tests. This range indicates good precision for the PCT when performed remotely with master slave manipulators in a shielded cell environment. When the results of the two laboratories were compared to each other, the agreement was within 20%. Normalized concentrations for the nonradioactive and radioactive elements in the PCT leachates measured at both facilities indicated that the radionuclides were released from the glass slower than the major soluble elements in the glass. For both laboratories, the normalized releases for both glasses were in the general order Li ∼ B ∼ Na > Si > Cs - 137 > Sb - 125 < Sr - 90. The normalized releases for the major soluble elements and the final pH values in the tests with radioactive glass are consistent with those for nonradioactive glasses with similar compositions. This indicates that there was no significant effect of radiation on the results of the PCT

  7. Waste glass melting stages

    International Nuclear Information System (INIS)

    Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

    1994-01-01

    Three simulated nuclear waste glass feeds, consisting of dried waste and glass frit, were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600 degrees C to 1000 degrees C. Simulated melter feeds from the Hanford Waste Vitrification Plant (HWVP), the Defense Waste Processing Facility (DWPF), and Kernforschungszentru Karlsruhe (KfK) in Germany were used. The samples were thin sectioned and examined by optical microscopy to investigate the stages of the conversion from feed to glass. Various phenomena were seen, such as frit softening, bubble formation, foaming, bubble motion and removal, convective mixing, and homogenization. The behavior of different feeds was similar, although the degree of gas generation and melt homogenization varied. 2 refs., 8 tabs

  8. Glass Property Data and Models for Estimating High-Level Waste Glass Volume

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, John D.; Fluegel, Alexander; Kim, Dong-Sang; Hrma, Pavel R.

    2009-10-05

    This report describes recent efforts to develop glass property models that can be used to help estimate the volume of high-level waste (HLW) glass that will result from vitrification of Hanford tank waste. The compositions of acceptable and processable HLW glasses need to be optimized to minimize the waste-form volume and, hence, to save cost. A database of properties and associated compositions for simulated waste glasses was collected for developing property-composition models. This database, although not comprehensive, represents a large fraction of data on waste-glass compositions and properties that were available at the time of this report. Glass property-composition models were fit to subsets of the database for several key glass properties. These models apply to a significantly broader composition space than those previously publised. These models should be considered for interim use in calculating properties of Hanford waste glasses.

  9. Glass Property Data and Models for Estimating High-Level Waste Glass Volume

    International Nuclear Information System (INIS)

    Vienna, John D.; Fluegel, Alexander; Kim, Dong-Sang; Hrma, Pavel R.

    2009-01-01

    This report describes recent efforts to develop glass property models that can be used to help estimate the volume of high-level waste (HLW) glass that will result from vitrification of Hanford tank waste. The compositions of acceptable and processable HLW glasses need to be optimized to minimize the waste-form volume and, hence, to save cost. A database of properties and associated compositions for simulated waste glasses was collected for developing property-composition models. This database, although not comprehensive, represents a large fraction of data on waste-glass compositions and properties that were available at the time of this report. Glass property-composition models were fit to subsets of the database for several key glass properties. These models apply to a significantly broader composition space than those previously publised. These models should be considered for interim use in calculating properties of Hanford waste glasses.

  10. Comparative waste forms study

    International Nuclear Information System (INIS)

    Wald, J.W.; Lokken, R.O.; Shade, J.W.; Rusin, J.M.

    1980-12-01

    A number of alternative process and waste form options exist for the immobilization of nuclear wastes. Although data exists on the characterization of these alternative waste forms, a straightforward comparison of product properties is difficult, due to the lack of standardized testing procedures. The characterization study described in this report involved the application of the same volatility, mechanical strength and leach tests to ten alternative waste forms, to assess product durability. Bulk property, phase analysis and microstructural examination of the simulated products, whose waste loading varied from 5% to 100% was also conducted. The specific waste forms investigated were as follows: Cold Pressed and Sintered PW-9 Calcine; Hot Pressed PW-9 Calcine; Hot Isostatic Pressed PW-9 Calcine; Cold Pressed and Sintered SPC-5B Supercalcine; Hot Isostatic pressed SPC-5B Supercalcine; Sintered PW-9 and 50% Glass Frit; Glass 76-68; Celsian Glass Ceramic; Type II Portland Cement and 10% PW-9 Calcine; and Type II Portland Cement and 10% SPC-5B Supercalcine. Bulk property data were used to calculate and compare the relative quantities of waste form volume produced at a spent fuel processing rate of 5 metric ton uranium/day. This quantity ranged from 3173 L/day (5280 Kg/day) for 10% SPC-5B supercalcine in cement to 83 L/day (294 Kg/day) for 100% calcine. Mechanical strength, volatility, and leach resistance tests provide data related to waste form durability. Glass, glass-ceramic and supercalcine ranked high in waste form durability where as the 100% PW-9 calcine ranked low. All other materials ranked between these two groupings

  11. Minimum Additive Waste Stabilization (MAWS)

    International Nuclear Information System (INIS)

    1994-02-01

    In the Minimum Additive Waste Stabilization(MAWS) concept, actual waste streams are utilized as additive resources for vitrification, which may contain the basic components (glass formers and fluxes) for making a suitable glass or glassy slag. If too much glass former is present, then the melt viscosity or temperature will be too high for processing; while if there is too much flux, then the durability may suffer. Therefore, there are optimum combinations of these two important classes of constituents depending on the criteria required. The challenge is to combine these resources in such a way that minimizes the use of non-waste additives yet yields a processable and durable final waste form for disposal. The benefit to this approach is that the volume of the final waste form is minimized (waste loading maximized) since little or no additives are used and vitrification itself results in volume reduction through evaporation of water, combustion of organics, and compaction of the solids into a non-porous glass. This implies a significant reduction in disposal costs due to volume reduction alone, and minimizes future risks/costs due to the long term durability and leach resistance of glass. This is accomplished by using integrated systems that are both cost-effective and produce an environmentally sound waste form for disposal. individual component technologies may include: vitrification; thermal destruction; soil washing; gas scrubbing/filtration; and, ion-exchange wastewater treatment. The particular combination of technologies will depend on the waste streams to be treated. At the heart of MAWS is vitrification technology, which incorporates all primary and secondary waste streams into a final, long-term, stabilized glass wasteform. The integrated technology approach, and view of waste streams as resources, is innovative yet practical to cost effectively treat a broad range of DOE mixed and low-level wastes

  12. Natural analogues of nuclear waste glass corrosion

    International Nuclear Information System (INIS)

    Abrajano, T.A. Jr.; Ebert, W.L.; Luo, J.S.

    1999-01-01

    This report reviews and summarizes studies performed to characterize the products and processes involved in the corrosion of natural glasses. Studies are also reviewed and evaluated on how well the corrosion of natural glasses in natural environments serves as an analogue for the corrosion of high-level radioactive waste glasses in an engineered geologic disposal system. A wide range of natural and experimental corrosion studies has been performed on three major groups of natural glasses: tektite, obsidian, and basalt. Studies of the corrosion of natural glass attempt to characterize both the nature of alteration products and the reaction kinetics. Information available on natural glass was then compared to corresponding information on the corrosion of nuclear waste glasses, specifically to resolve two key questions: (1) whether one or more natural glasses behave similarly to nuclear waste glasses in laboratory tests, and (2) how these similarities can be used to support projections of the long-term corrosion of nuclear waste glasses. The corrosion behavior of basaltic glasses was most similar to that of nuclear waste glasses, but the corrosion of tektite and obsidian glasses involves certain processes that also occur during the corrosion of nuclear waste glasses. The reactions and processes that control basalt glass dissolution are similar to those that are important in nuclear waste glass dissolution. The key reaction of the overall corrosion mechanism is network hydrolysis, which eventually breaks down the glass network structure that remains after the initial ion-exchange and diffusion processes. This review also highlights some unresolved issues related to the application of an analogue approach to predicting long-term behavior of nuclear waste glass corrosion, such as discrepancies between experimental and field-based estimates of kinetic parameters for basaltic glasses

  13. Natural analogues of nuclear waste glass corrosion.

    Energy Technology Data Exchange (ETDEWEB)

    Abrajano, T.A. Jr.; Ebert, W.L.; Luo, J.S.

    1999-01-06

    This report reviews and summarizes studies performed to characterize the products and processes involved in the corrosion of natural glasses. Studies are also reviewed and evaluated on how well the corrosion of natural glasses in natural environments serves as an analogue for the corrosion of high-level radioactive waste glasses in an engineered geologic disposal system. A wide range of natural and experimental corrosion studies has been performed on three major groups of natural glasses: tektite, obsidian, and basalt. Studies of the corrosion of natural glass attempt to characterize both the nature of alteration products and the reaction kinetics. Information available on natural glass was then compared to corresponding information on the corrosion of nuclear waste glasses, specifically to resolve two key questions: (1) whether one or more natural glasses behave similarly to nuclear waste glasses in laboratory tests, and (2) how these similarities can be used to support projections of the long-term corrosion of nuclear waste glasses. The corrosion behavior of basaltic glasses was most similar to that of nuclear waste glasses, but the corrosion of tektite and obsidian glasses involves certain processes that also occur during the corrosion of nuclear waste glasses. The reactions and processes that control basalt glass dissolution are similar to those that are important in nuclear waste glass dissolution. The key reaction of the overall corrosion mechanism is network hydrolysis, which eventually breaks down the glass network structure that remains after the initial ion-exchange and diffusion processes. This review also highlights some unresolved issues related to the application of an analogue approach to predicting long-term behavior of nuclear waste glass corrosion, such as discrepancies between experimental and field-based estimates of kinetic parameters for basaltic glasses.

  14. Laboratory testing of LITCO glasses

    International Nuclear Information System (INIS)

    Ellison, A.; Wolf, S.; Buck, E.; Luo, J.S.; Dietz, N.; Bates, J.K.; Ebert, W.L.

    1995-01-01

    The purpose of this program is to measure, the intermediate and long-term durability of glasses developed by Lockheed Idaho Technology Co. (LITCO) for the immobilization of calcined radioactive wastes. The objective is to use accelerated corrosion tests as an aid in developing durable waste form compositions. This is a report of tests performed on two LITCO glass compositions, Formula 127 and Formula 532. The main avenue for release of radionuclides into the environment in a geologic repository is the reaction of a waste glass with ground water, which alters the glass and releases its components into solution. These stages in glass corrosion are analyzed by using accelerated laboratory tests in which the ratio of sample surface area to solution volume, SA/V, is varied. At low SA/V, the solution concentrations of glass corrosion products remain low and the reaction approaches the forward rate. At higher SA/V the solution approaches saturation levels for glass corrosion products. At very high SA/V the solution is rapidly saturated in glass corrosion products and secondary crystalline phases precipitate. Tests at very high SA/V provide information about the composition of the solution at saturation or, when no solution is recovered, the identities and the order of appearance of secondary crystalline phases. Tests were applied to Formula 127 and Formula 532 glasses to provide information about the interim and long-term stages in glass corrosion

  15. Silicate glasses. Chapter 1

    International Nuclear Information System (INIS)

    Lutze, W.

    1988-01-01

    This chapter is a survey of world-wide research and development efforts in nuclear waste glasses and its production technology. The principal glasses considered are silicate glasses which contain boron, i.e. borosilicate glass. A historical overview of waste form development programs in nine countries is followed by a summary of the design criteria for borosilicate glass compositions glass compositions. In the sections on glass properties the waste form is characterized in terms of potential alterations under the influence of heat, thermal gradients, radiation, aqueous solutions and combinations thereof. The topics are phase transformations, mechanical properties, radiation effects and chemical durability. The results from studies of volcanic glasses, as natural analogues for borosilicate nuclear waste glasses in order to verify predictions obtained from short-term tests in the laboratory, have been compiled in a special section on natural analogues. A special section on advanced vitrification techniques summarizes the various actual and potential processing schemes and describes the facilities. The literature has been considered until 1985. (author). 430 refs.; 68 figs.; 29 tabs

  16. Characterization of glass and glass ceramic nuclear waste forms

    International Nuclear Information System (INIS)

    Lutze, W.; Borchardt, J.; De, A.K.

    1979-01-01

    Characteristics of solidified nuclear waste forms, glass and glass ceramic compositions and the properties (composition, thermal stability, crystallization, phase behavior, chemical stability, mechanical stability, and radiation effects) of glasses and glass ceramics are discussed. The preparation of glass ceramics may be an optional step for proposed vitrification plants if tailored glasses are used. Glass ceramics exhibit some improved properties with respect to glasses. The overall leach resistance is similar to that of glasses. An increased leach resistance may become effective for single radionuclides being hosted in highly insoluble crystal phases mainly when higher melting temperatures are applicable in order to get more leach resistant residual glass phases. The development of glass ceramic is going on. The technological feasibility is still to be demonstrated. The potential gain of stability when using glass ceramics qualifies the material as an alternative nuclear waste form

  17. Glass-bonded iodosodalite waste form for immobilization of 129 I

    Energy Technology Data Exchange (ETDEWEB)

    Chong, Saehwa; Peterson, Jacob A.; Riley, Brian J.; Tabada, Diana; Wall, Donald; Corkhill, Claire L.; McCloy, John S.

    2018-06-01

    Immobilization of radioiodine (e.g., 129I, 131I) is an important need for current and future nuclear fuel cycles. For the current work, iodosodalite [Na8(AlSiO4)6I2] was synthesized hydrothermally from metakaolin, NaI, and NaOH. Following hydrothermal treatment, dried unwashed powders were used to make glass-bonded iodosodalite waste forms by heating pressed pellets at 650, 750, or 850 °C with two different types of sodium borosilicate glass binders, i.e., NBS-4 and SA-800. These heat-treated specimens were characterized with X-ray diffraction, Fourier-transform infrared spectroscopy, scanning electron microscopy, energy dispersive spectroscopy, thermal analysis, porosity and density measurements, neutron activation analysis, and inductively-coupled plasma mass spectrometry. The pellets mixed with 10 mass% of NBS-4 or SA-800 and heat-treated at 750 °C contained relatively high percentage iodine retention (~44-47 % of the maximum iodine loading) with relatively low porosities, while other pellets with higher percentages iodine retention either contained higher porosity or were not completely sintered. ASTM C1308 chemical durability tests of monolithic specimens showed a large initial release of Na, Al, Si, and I on the first day, possibly from water-soluble salt crystals or non-durable amorphous phases. Release rates of Na and Si were higher than for Al and I, probably due to a poorly durable Na-Si-O phase from the glass bonding matrix. The cumulative normalized release of iodine was 12.5 g m-2 for the first 10 1-d exchanges, suggestive of coherent dissolution. The average release rate from 10-24 days during the 7-d exchange intervals was 0.2336 g m-2 d-1.

  18. The borosilicate glass for 'PAMELA'

    International Nuclear Information System (INIS)

    Schiewer, E.

    1986-01-01

    The low enriched waste concentrate (LEWC) stored at Mol, Belgium, will be solidified in the vitrification plant 'PAMELA'. An alkali-borosilicate glass was developed by the Hahn-Meitner-Institut, Berlin, which dissolves (11 +- 3)wt% waste oxides while providing sufficient flexibility for changes in the process parameters. The development of the glass labelled SM513LW11 is described. Important properties of the glass melt (viscosity, resistivity, formation of yellow phase) and of the glass (corrosion in aqueous solutions, crystallization) are reported. The corrosion data of this glass are similar to those of other HLW-glasses. Less than five wt% of crystalline material are produced upon cooling of large glass blocks. Crystallization does not affect the chemical durability. (Auth.)

  19. Vitrification of TRU wastes at Rocky Flats Plant

    International Nuclear Information System (INIS)

    Williams, P.M.; Johnson, A.J.; Ledford, J.A.

    1979-01-01

    Immobilization of incinerator ash and various noncombustible TRU wastes was investigated. In three different research projects borosilicate glass proved to be the best candidate for TRU waste fixation. This glass has excellent chemical durability, long-term stability in the presence of radiation, and will withstand continuous temperatures up to 400 0 C without devitrification. In addition, wastes prepared in the form of glass will attain densities of approximately 2500 kg/m 3 (2.5 g/cc). The free forming method of producing glass buttons provides a very simple, consistent, low maintenance way of producing a final waste form for transporting and either retrievable or permanent storage for TRU waste. The vitrification process produces a durable glass from the low density ash generated by the fluidized bed incinerator process and provides volume and weight reductions that are superior to other fixation processes. This results in decreased transportation and storage costs

  20. Volumetric change of simulated radioactive waste glass irradiated by electron accelerator. [Silica glass

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Seichi; Furuya, Hirotaka; Inagaki, Yaohiro; Kozaka, Tetsuo; Sugisaki, Masayasu

    1987-11-01

    Density changes of simulated radioactive waste glasses, silica glass and Pyrex glass irradiated by an electron accelerator were measured by a ''sink-float'' technique. The density changes of the waste and silica glasses were less than 0.05 %, irradiated at 2.0 MeV up to the fluence of 1.7 x 10/sup 17/ ecm/sup 2/, while were remarkably smaller than that of Pyrex glass of 0.18 % shrinkage. Precision of the measurements in the density changes of the waste glass was lower than that of Pyrex glass possibly because of the inhomogeneity of the waste glass

  1. Laboratory testing of glasses for Lockheed Idaho Technology Co. - fiscal year 1994 report

    International Nuclear Information System (INIS)

    Ellison, A.J.G.; Wolf, S.F.; Bates, J.K.

    1995-04-01

    The purpose of this project is to measure the intermediate and long-term durability of vitrified waste forms developed by Lockheed Idaho Technology Co. (LITCO) for the immobilization of calcined radioactive wastes at Idaho National Engineering Laboratory. Two vitreous materials referred to as Formula 127 and Formula 532, have been subjected to accelerated durability tests to measure their long-term performance. Formula 127 consists of a glass matrix containing 5-10 vol % fluorite (CaF 2 ) as a primary crystalline phase. It shows low releases of glass components to solution in 7-, 28-, 70-, and 140-day Product Consistency Tests performed at 2000 m -1 at 90 degrees C. In these tests, release rates for glass-forming components were similar to those found for durable waste glasses. The Ca and F released by the glass as it corrodes appear to reprecipitate as fluorite. Formula 532 consists of a glass matrix containing 5-10 vol % of an Al-Si-rich primary crystalline phase. The release rates for components other than aluminum are relatively low, but aluminum is released at a much higher rate than is typical for durable waste glasses. Secondary crystalline phases form relatively early during the corrosion of Formula 532 and appear to consist almost entirely of the Al-Si-rich primary phase (or a crystal with the same Al:Si ratio) and a sodium-bearing zeolite. Future test results are expected to highlight the relative importance of primary and secondary crystalline phases to the rate of corrosion of Formula 127 and Formula 532

  2. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1981-06-01

    The process which will be used to decontaminate waste glass canisters at the Savannah River Plant consists of: decontamination (slurry blasting); rinse (high-pressure water); and spot decontamination (high-pressure water plus slurry). No additional waste will be produced by this process because glass frit used in decontamination will be mixed with the radioactive waste and fed into the glass melter. Decontamination of waste glass canisters with chemical and abrasive blasting techniques was investigated. The ability of a chemical technique with HNO 3 -HF and H 2 C 2 O 4 to remove baked-on contamination was demonstrated. A correlation between oxide removal and decontamination was observed. Oxide removal and, thus, decontamination by abrasive blasting techniques with glass frit as the abrasive was proposed and demonstrated

  3. Corrosion of simulated nuclear waste glass

    International Nuclear Information System (INIS)

    Music, S.; Ristic, M.; Gotic, M.; Foric, J.

    1988-01-01

    In this study the preparation and characterization of borosilicate glasses of different chemical composition were investigated. Borosilicate glasses were doped with simulated nuclear waste oxides. The chemical corrosion in water of these glasses was followed by measuring the leach rates as a function of time. It was found that a simulated nuclear waste glass with the chemical composition (weight %), 15.61% Na 2 O, 10.39% B 2 O 3 , 45.31% SiO 2 , 13.42% ZnO, 6.61% TiO 2 and 8.66% waste oxides, is characterized by low melting temperature and with good corrosion resistance in water. Influence of passive layers on the leaching behaviour of nuclear waste glasses is discussed. (author) 20 refs.; 7 figs.; 4 tabs

  4. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1981-01-01

    The process which will be used to decontaminate waste glass canisters at the Savannah River Plant consists of: decontamination (slurry blasting); rinse (high-pressure water); and spot decontamination (high-pressure water plus slurry). No additional waste will be produced by this process because glass frit used in decontamination will be mixed with the radioactive waste and fed into the glass melter. Decontamination of waste glass canisters with chemical and abrasive blasting techniques was investigated. The ability of a chemical technique with HNO 3 -HF and H 2 C 2 O 4 to remove baked-on contamination was demonstrated. A correlation between oxide removal and decontamination was observed. Oxide removal and, thus, decontamination by abrasive blasting techniques with glass frit as the abrasive was proposed and demonstrated

  5. Immobilization of high level nuclear wastes in sintered glasses. Devitrification evaluation produced with different thermal treatments

    International Nuclear Information System (INIS)

    Messi de Bernasconi, N.B.; Russo, D.O.; Bevilacqua, M.E.; Sterba, M.E.; Heredia, A.D.; Audero, M.A.

    1990-01-01

    This work describes immobilization of high level nuclear wastes in sintered glass, as alternative way to melting glass. Different chemical compositions of borosilicate glass with simulate waste were utilized and satisfactory results were obtained at laboratory scale. As another contribution to the materials studies by X ray powder diffraction analysis, the devitrification produced with different thermal treatments, was evaluated. The effect of the thermal history on the behaviour of fission products containing glasses has been studied by several working groups in the field of high level waste fixation. When the glass is cooled through the temperature range from 800 deg C down to less than 400 deg C (these temperatures are approximates) nucleation and crystal growth can take place. The rate of crystallization will be maximum near the transformation point but through this rate may be low at lower temperatures, devitrification can still occur over long periods of time, depending on the glass composition. It was verified that there can be an appreciable increase in leaching in some waste glass compositions owing to the presence of crystalline phases. On the other hand, other compositions show very little change in leachability and the devitrified product is often preferable as there is less tendency to cracking, particularly in massive blocks of glass. A borosilicate glass, named SG7, which was developed specially in the KfK for the hot pressing of HLW with glass frit was studied. It presents a much enhanced chemical durability than borosolicate glass developed for the melting process. The crystallization behaviour of SG7 glass products was investigated in our own experiments by annealing sintered samples up to 3000 h at temperatures between 675 and 825 deg C. The samples had contained simulated waste with noble metals, since these might act as foreign nuclei for crystallization. Results on the extent of devitrification and time- temperature- transformation curves are

  6. Database for waste glass composition and properties

    International Nuclear Information System (INIS)

    Peters, R.D.; Chapman, C.C.; Mendel, J.E.; Williams, C.G.

    1993-09-01

    A database of waste glass composition and properties, called PNL Waste Glass Database, has been developed. The source of data is published literature and files from projects funded by the US Department of Energy. The glass data have been organized into categories and corresponding data files have been prepared. These categories are glass chemical composition, thermal properties, leaching data, waste composition, glass radionuclide composition and crystallinity data. The data files are compatible with commercial database software. Glass compositions are linked to properties across the various files using a unique glass code. Programs have been written in database software language to permit searches and retrievals of data. The database provides easy access to the vast quantities of glass compositions and properties that have been studied. It will be a tool for researchers and others investigating vitrification and glass waste forms

  7. Characterization of the Italian glasses and their interaction with clay Task 3 Characterization of radioactive waste forms a series of final reports (1985-89) No 23

    International Nuclear Information System (INIS)

    Cantale, C.; Castelli, S.; Donato, A.; Traverso, D.M.

    1991-01-01

    The objective of this research work was the selection of a borosilicate glass composition suitable for the solidification of the HLW stream coming from the treatment of all the high-level wastes stored in Italy (MTR, Candu and Elk River) and the characterization of this glass with reference to the geological disposal. This research work was part of an Italian research project named 'Ulisse project', whose goal was the development and the demonstration of an integrated treatment of all the HLW stored in Italy, after their mixing (resulting waste: MCE waste). The main concept is to carry out a pre-treatment of the wastes, in order to concentrate the HLW fraction and to simplify the vitrification process, separating the most part of the inert salts. The research work concerning the separation process and pilot plant demonstration of the pre-treatment process were carried out in the framework of the CEC R and D programme (Contract No Fl1W-0011-lS). The laboratory studies concerning the vitrification of the resulting HLW streams and the vitrification demonstration in the Italian full-scale, inactive IVET plant complete the 'Ulisse project'. Some glass compositions were prepared and preliminarily characterized. The glass named BAZ was finally selected. A complete characterization of this glass was carried out in order to evaluate its mechanical, physical and physico-chemical properties. The chemical durability was evaluated by the MCC-1 static leach test at 90 0 C, using three different leachants and two surface-area to leachant-volume ratios. The same characterization programme was applied to the BAZ glass produced in the IVET plant during the plant vitrification demonstration programme. A comparison between the two glasses and a critical evaluation of their performances with respect to other nuclear waste glasses' durability was performed. 25 refs.; 46 figs.; 20 tabs

  8. Production of a High-Level Waste Glass from Hanford Waste Samples

    International Nuclear Information System (INIS)

    Crawford, C.L.; Farrara, D.M.; Ha, B.C.; Bibler, N.E.

    1998-09-01

    The HLW glass was produced from a HLW sludge slurry (Envelope D Waste), eluate waste streams containing high levels of Cs-137 and Tc-99, solids containing both Sr-90 and transuranics (TRU), and glass-forming chemicals. The eluates and Sr-90/TRU solids were obtained from ion-exchange and precipitation pretreatments, respectively, of other Hanford supernate samples (Envelopes A, B and C Waste). The glass was vitrified by mixing the different waste streams with glass-forming chemicals in platinum/gold crucibles and heating the mixture to 1150 degree C. Resulting glass analyses indicated that the HLW glass waste form composition was close to the target composition. The targeted waste loading of Envelope D sludge solids in the HLW glass was 30.7 wt percent, exclusive of Na and Si oxides. Condensate samples from the off-gas condenser and off-gas dry-ice trap indicated that very little of the radionuclides were volatilized during vitrification. Microstructure analysis of the HLW glass using Scanning Electron Microscopy (SEM) and Energy Dispersive X-Ray Analysis (EDAX) showed what appeared to be iron spinel in the HLW glass. Further X-Ray Diffraction (XRD) analysis confirmed the presence of nickel spinel trevorite (NiFe2O4). These crystals did not degrade the leaching characteristics of the glass. The HLW glass waste form passed leach tests that included a standard 90 degree C Product Consistency Test (PCT) and a modified version of the United States Environmental Protection Agency Toxicity Characteristic Leaching Procedure (TCLP)

  9. Inhibitory Effect of Waste Glass Powder on ASR Expansion Induced by Waste Glass Aggregate

    Directory of Open Access Journals (Sweden)

    Shuhua Liu

    2015-10-01

    Full Text Available Detailed research is carried out to ascertain the inhibitory effect of waste glass powder (WGP on alkali-silica reaction (ASR expansion induced by waste glass aggregate in this paper. The alkali reactivity of waste glass aggregate is examined by two methods in accordance with the China Test Code SL352-2006. The potential of WGP to control the ASR expansion is determined in terms of mean diameter, specific surface area, content of WGP and curing temperature. Two mathematical models are developed to estimate the inhibitory efficiency of WGP. These studies show that there is ASR risk with an ASR expansion rate over 0.2% when the sand contains more than 30% glass aggregate. However, WGP can effectively control the ASR expansion and inhibit the expansion rate induced by the glass aggregate to be under 0.1%. The two mathematical models have good simulation results, which can be used to evaluate the inhibitory effect of WGP on ASR risk.

  10. Effects of waste content of glass waste forms on Savannah River high-level waste disposal costs

    International Nuclear Information System (INIS)

    McDonell, W.R.; Jantzen, C.M.

    1985-01-01

    Effects of the waste content of glass waste forms of Savannah River high-level waste disposal costs are evaluated by their impact on the number of waste canisters produced. Changes in waste content affect onsite Defense Waste Processing Facility (DWPF) costs as well as offsite shipping and repository emplacement charges. A nominal 1% increase over the 28 wt % waste loading of DWPF glass would reduce disposal costs by about $50 million for Savannah River wastes generated to the year 2000. Waste form modifications under current study include adjustments of glass frit content to compensate for added salt decontamination residues and increased sludge loadings in the DWPF glass. Projected cost reductions demonstrate significant incentives for continued optimization of the glass waste loadings. 13 refs., 3 figs., 3 tabs

  11. The immobilization of High Level Waste Into Glass

    International Nuclear Information System (INIS)

    Aisyah; Martono, H.

    1998-01-01

    High level liquid waste is generated from the first step extraction in the nuclear fuel reprocessing. The waste is immobilized with boro-silicate glass. A certain composition of glass is needed for a certain type of waste, so that the properties of waste glass would meet the requirement either for further process or for disposal. The effect of waste loading on either density, thermal expansion, softening point and leaching rate has been studied. The composition of the high level liquid waste has been determined by ORIGEN 2 and the result has been used to prepare simulated high level waste. The waste loading in the waste glass has been set to be 19.48; 22.32; 25.27; and 26.59 weight percent. The result shows that increasing the waste loading has resulted in the higher density with no thermal expansion and softening point significant change. The increase in the waste loading increase that leaching rate. The properties of the waste glass in this research have not shown any deviation from the standard waste glass properties

  12. Effects of container material on PCT leach test results for high-level nuclear waste glasses

    International Nuclear Information System (INIS)

    Xing, S.B.; Pegg, I.L.

    1994-01-01

    A glass-based waste form used for the immobilization of high-level nuclear wastes should exhibit good resistance to aqueous corrosion since typically this is the primary process by which radionucleides could be released into the environment upon failure of other barriers. In the USA, the Waste Acceptance Product Specifications (WAPS) provides a set of requirements to ensure the consistency of the waste forms produced and specifies the Product Consistency Test (PCT) as a measure of relative chemical durability. While the PCT procedure permits usage of both Teflon and stainless steel vessels for testing of simulated development glasses, Teflon is not permitted for testing of production glasses due to radiative degradation. The results presented in this paper indicate that there are very significant differences between tests conducted in the two types of vessels due to the well-known permeability of Teflon to atmospheric carbon dioxide which results in lowering of the solution pH and a consequent reduction in the leach rate of silicate glasses. A wide range of nuclear waste glass compositions was subjected to the PCT procedure using both Teflon and stainless steel vessels. The magnitude of the effect (up to a factor of four for B, Na, Li concentrations) depends strongly on glass composition, therefore the isolated checks performed previously were inconclusive. The permeability to CO, of two types of Teflon vessels specified in the PCT procedure was directly measured using buffer solutions: ingress of CO, is linear in time, strongly pH-dependent, and was as high as 100 ppm after 7 days. In actual PCT tests in Teflon vessels, the total CO, content was 560 ppm after 87 days and 1930 ppm after one year

  13. SUMMARY OF FY11 SULFATE RETENTION STUDIES FOR DEFENSE WASTE PROCESSING FACILITY GLASS

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K.; Edwards, T.

    2012-05-08

    This report describes the results of studies related to the incorporation of sulfate in high level waste (HLW) borosilicate glass produced at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). A group of simulated HLW glasses produced for earlier sulfate retention studies was selected for full chemical composition measurements to determine whether there is any clear link between composition and sulfate retention over the compositional region evaluated. In addition, the viscosity of several glasses was measured to support future efforts in modeling sulfate solubility as a function of predicted viscosity. The intent of these studies was to develop a better understanding of sulfate retention in borosilicate HLW glass to allow for higher loadings of sulfate containing waste. Based on the results of these and other studies, the ability to improve sulfate solubility in DWPF borosilicate glasses lies in reducing the connectivity of the glass network structure. This can be achieved, as an example, by increasing the concentration of alkali species in the glass. However, this must be balanced with other effects of reduced network connectivity, such as reduced viscosity, potentially lower chemical durability, and in the case of higher sodium and aluminum concentrations, the propensity for nepheline crystallization. Future DWPF processing is likely to target higher waste loadings and higher sludge sodium concentrations, meaning that alkali concentrations in the glass will already be relatively high. It is therefore unlikely that there will be the ability to target significantly higher total alkali concentrations in the glass solely to support increased sulfate solubility without the increased alkali concentration causing failure of other Product Composition Control System (PCCS) constraints, such as low viscosity and durability. No individual components were found to provide a significant improvement in sulfate retention (i.e., an increase of the magnitude

  14. Relative leach behavior of waste glasses and naturally occurring glasses

    International Nuclear Information System (INIS)

    Adams, P.B.

    1979-01-01

    Simulated nuclear waste glasses of the sodium-borosilicate type with a low waste loading and of the zinc-borosilicate type with a high waste loading have been compared with obsidians. The resuls indicate that the waste glasses would corrode in normal natural environments at a rate of about 0.1 μm per year at 30 0 C and about 5 μm per year at 90 0 C, compared with obsidians which seem to corrode at, or less than, about 0.01 μm per year at 30 0 C and less than 1 μm per year at 90 0 C. Activation energies for reactions of the two waste glasses with pure water are about 20 kcal/g-mol. 3 figures, 7 tables

  15. Towards optimization of nuclear waste glass: Constraints, property models, and waste loading

    International Nuclear Information System (INIS)

    Hrma, P.

    1994-04-01

    Vitrification of both low- and high-level wastes from 177 tanks at Hanford poses a great challenge to glass makers, whose task is to formulate a system of glasses that are acceptable to the federal repository for disposal. The enormous quantity of the waste requires a glass product of the lowest possible volume. The incomplete knowledge of waste composition, its variability, and lack of an appropriate vitrification technology further complicates this difficult task. A simple relationship between the waste loading and the waste glass volume is presented and applied to the predominantly refractory (usually high-activity) and predominantly alkaline (usually low-activity) waste types. Three factors that limit waste loading are discussed, namely product acceptability, melter processing, and model validity. Glass formulation and optimization problems are identified and a broader approach to uncertainties is suggested

  16. Retention of Halogens in Waste Glass

    Energy Technology Data Exchange (ETDEWEB)

    Hrma, Pavel R.

    2010-05-01

    In spite of their potential roles as melting rate accelerators and foam breakers, halogens are generally viewed as troublesome components for glass processing. Of five halogens, F, Cl, Br, I, and At, all but At may occur in nuclear waste. A nuclear waste feed may contain up to 10 g of F, 4 g of Cl, and ≤100 mg of Br and I per kg of glass. The main concern is halogen volatility, producing hazardous fumes and particulates, and the radioactive iodine 129 isotope of 1.7x10^7-year half life. Because F and Cl are soluble in oxide glasses and tend to precipitate on cooling, they can be retained in the waste glass in the form of dissolved constituents or as dispersed crystalline inclusions. This report compiles known halogen-retention data in both high-level waste (HLW) and low-activity waste (LAW) glasses. Because of its radioactivity, the main focus is on I. Available data on F and Cl were compiled for comparison. Though Br is present in nuclear wastes, it is usually ignored; no data on Br retention were found.

  17. The effects of uranium oxide high-level waste on the structure of iron phosphate glasses

    International Nuclear Information System (INIS)

    Badyal, Y.

    1998-01-01

    Because of their unusually good chemical durability, iron phosphate glasses are a natural candidate for a nuclear waste disposal glass. We have studied the effects of UO 2 high-level waste on the structure of iron phosphate glasses with both neutron and high-energy x-ray diffraction using the GLAD instrument of the Intense Pulsed Neutron Source and the 1-BM bending magnet beamline of the Advanced Photon Source, respectively. The results of neutron scattering, which is mostly sensitive to correlations involving light atoms i.e. O-O, Fe-O and P-O, suggest the main structural features of the base glass are largely unaffected by the addition of UO 2 . The nearest-neighbor P-O, Fe-O and O-O peaks remain at the same position in real space and their intensities scale approximately with concentration. These findings are consistent with the earlier results of Raman scattering and EXAFS on the Fe-K edge wherein both cases the spectra remain similar to the base glass. High-energy x-ray scattering which is sensitive to correlations involving the heavier atoms and thus complements the neutron measurements, is also consistent with uranium occupying interstitial sites in the relatively undisturbed base glass structure. However, important questions remain as to the precise local structure and oxidation state of uranium in these glasses

  18. Demonstration of the Defense Waste Processing Facility vitrification process for Tank 42 radioactive sludge -- Glass preparation and characterization

    International Nuclear Information System (INIS)

    Bibler, N.E.; Fellinger, T.L.; Marshall, K.M.; Crawford, C.L.; Cozzi, A.D.; Edwards, T.B.

    1999-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) is currently processing and immobilizing the radioactive high level waste sludge at SRS into a durable borosilicate glass for final geological disposal. The DWPF has recently finished processing the first radioactive sludge batch, and is ready for the second batch of radioactive sludge. The second batch is primarily sludge from Tank 42. Before processing this batch in the DWPF, the DWPF process flowsheet has to be demonstrated with a sample of Tank 42 sludge to ensure that an acceptable melter feed and glass can be made. This demonstration was recently completed in the Shielded Cells Facility at SRS. An earlier paper in these proceedings described the sludge composition and processes necessary for producing an acceptable melter fee. This paper describes the preparation and characterization of the glass from that demonstration. Results substantiate that Tank 42 sludge after mixing with the proper amount of glass forming frit (Frit 200) can be processed to make an acceptable glass

  19. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1982-01-01

    A Defense Waste Processing Facility (DWPF) is currently being designed to convert Savannah River Plant liquid, high-level radioactive waste into a solid form, such as borosilicate glass. To prevent the spread of radioactivity, the outside of the canisters of waste glass must have very low levels of smearable radioactive contamination before they are removed from the DWPF. Several techniques were considered for canister decontamination: high-pressure water spray, electropolishing, chemical dissolution, and abrasive blasting. An abrasive blasting technique using a glass frit slurry has been selected for use in the DWPF. No additional equipment is needed to process waste generated from decontamination. Frit used as the abrasive will be mixed with the waste and fed to the glass melter. In contrast, chemical and electrochemical techniques require more space in the DWPF, and produce large amounts of contaminated by-products, which are difficult to immobilize by vitrification

  20. Glass and nuclear wastes

    International Nuclear Information System (INIS)

    Sombret, C.

    1982-10-01

    Glass shows interesting technical and economical properties for long term storage of solidified radioactive wastes by vitrification or embedding. Glass composition, vitrification processes, stability under irradiation, thermal stability and aqueous corrosion are studied [fr

  1. Prediction of waste glass melt rates

    International Nuclear Information System (INIS)

    Lee, L.

    1987-01-01

    Under contract to the Department of Energy, the Du Pont Company has begun construction of a Defense Waste Processing Facility to immobilize radioactive wastes now stored as liquids at the Department of Energy's Savannah River Plant. The immobilization process solidifies waste sludge by vitrification into a leach-resistant borosilicate glass. Development of this process has been the responsibility of the Savannah River Laboratory. As part of the development, a simple model was developed to predict the melt rates for the waste glass melter. This model is based on an energy balance for the cold cap and gives very good agreement with melt rate data obtained from experimental campaigns in smaller scale waste glass melters

  2. From glass structure to its chemical durability

    International Nuclear Information System (INIS)

    Angeli, F.

    2009-01-01

    The author gives an overview of his research activities. He more precisely reports studies related to glass structure based on nuclei observed by NMR and present in glasses of interest for nuclear activities. He discusses the influence of chemical composition on structure, and discusses information which can be extracted from network formers (Al, B) and modifiers (Na, Ca), and from oxygen present in the network linkages of oxide glasses. He discusses the different experimental and modelling approaches which enable structural and morphological information to be obtained at a mesoscopic scale. The last part deals with the investigation of the long term behaviour of confinement matrices (glassy matrix for medium-activity wastes, ceramic matrix)

  3. Effect of different glass and zeolite A compositions on the leach resistance of ceramic waste forms

    International Nuclear Information System (INIS)

    Lewis, M.A.; Hash, M.; Glandorf, D.

    1996-01-01

    A ceramic waste form is being developed for waste generated during electrometallurgical treatment of spent nuclear fuel. The waste is generated when fission products are removed from the electrolyte, LiCl-KCl eutectic. The waste form is a composite fabricated by hot isostatic pressing a mixture of glass frit and zeolite occluded with fission products and salt. Normalized release rate is less than 1 g/m 2 d for all elements in MCC-1 leach test run for 28 days in deionized water at 90 C. This leach resistance is comparable to that of early Savannah River glasses. We are investigating how leach resistance is affected by changes in cationic form of zeolite and in glass composition. Composites were made with 3 forms of zeolite A and 6 glasses. We used 3-day ASTM C1220-92 (formerly MCC-1) leach tests to screen samples for development purposes only. The leach test results show that the glass composites of zeolites 5A and 4A retain fission products equally well. Loss of Cs is small (0.1-0.5 wt%), while the loss of divalent and trivalent fission products is one or more orders of magnitude smaller. Composites of 5A retain chloride ion better in these short-term screens than 4A and 3A. The more leach resistant composites were made with durable glasses rich in silica and poor in alkaline earth oxides. XRD show that a salt phase was absent in the leach resistant composites of 5A and the better glasses but was present in the other composites with poorer leach performance. Thus, absence of salt phase corresponds to improved leach resistance. Interactions between zeolite and glass depend on composition of both

  4. Complexing Agents and pH Influence on Chemical Durability of Type I Molded Glass Containers.

    Science.gov (United States)

    Biavati, Alberto; Poncini, Michele; Ferrarini, Arianna; Favaro, Nicola; Scarpa, Martina; Vallotto, Marta

    2017-01-01

    Among the factors that affect the glass surface chemical durability, pH and complexing agents present in aqueous solution have the main role. Glass surface attack can be also related to the delamination issue causing glass particles' appearance in the pharmaceutical preparation. A few methods to check for glass containers delamination propensity and some control guidelines have been proposed. The present study emphasizes the possible synergy between a few complexing agents with pH on borosilicate glass chemical durability.Hydrolytic attack was performed in small-volume 23 mL type I glass containers autoclaved according to the European Pharmacopoeia or United States Pharmacopeia for 1 h at 121 °C, in order to enhance the chemical attack due to time, temperature, and the unfavorable surface/volume ratio. Solutions of 0.048 M or 0.024 M (M/L) of the acids citric, glutaric, acetic, EDTA (ethylenediaminetetraacetic acid), together with sodium phosphate with water for comparison, were used for the trials. The pH was adjusted ±0.05 units at fixed values 5.5, 6.6, 7, 7.4, 8, and 9 by LiOH diluted solution.Because silicon is the main glass network former, silicon release into the attack solutions was chosen as the main index of the glass surface attack and analysed by inductively coupled plasma atomic emission spectrophotometry. The work was completed by the analysis of the silicon release in the worst attack conditions of molded glass, soda lime type II glass, and tubing borosilicate glass vials to compare different glass compositions and forming technologies. Surface analysis by scanning electron microscopy was finally performed to check for the surface status after the worst chemical attack condition by citric acid. LAY ABSTRACT: Glass, like every packaging material, can have some usage limits, mainly in basic pH solutions. The issue of glass surface degradation particles that appear in vials (delamination) has forced a number of drug product recalls in recent years

  5. High level waste at Hanford: Potential for waste loading maximization

    International Nuclear Information System (INIS)

    Hrma, P.R.; Bailey, A.W.

    1995-09-01

    The loading of Hanford nuclear waste in borosilicate glass is limited by phase-related phenomena, such as crystallization or formation of immiscible liquids, and by breakdown of the glass structure because of an excessive concentration of modifiers. The phase-related phenomena cause both processing and product quality problems. The deterioration of product durability determines the ultimate waste loading limit if all processing problems are resolved. Concrete examples and mass-balance based calculations show that a substantial potential exists for increasing waste loading of high-level wastes that contain a large fraction of refractory components

  6. Immobilization of radioactive waste in glass matrices

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1978-01-01

    A promising process for long-term management of high-level radioactive waste is to immobilize the waste in a borosilicate glass matrix. Among the most important criteria characterizing the integrity of the large-scale glass-waste forms are that they possess good chemical stability (including low leachability), thermal stability, mechanical integrity, and high radiation stability. Fulfillment of these criteria ensures the maximum margin of safety of glass-waste products, following solidification, handling, transportation, and long-term storage

  7. Glass-Ceramic Waste Forms for Uranium and Plutonium Residues Wastes - 13164

    International Nuclear Information System (INIS)

    Stewart, Martin W.A.; Moricca, Sam A.; Zhang, Yingjie; Day, R. Arthur; Begg, Bruce D.; Scales, Charlie R.; Maddrell, Ewan R.; Hobbs, Jeff

    2013-01-01

    A program of work has been undertaken to treat plutonium-residues wastes at Sellafield. These have arisen from past fuel development work and are highly variable in both physical and chemical composition. The principal radiological elements present are U and Pu, with small amounts of Th. The waste packages contain Pu in amounts that are too low to be economically recycled as fuel and too high to be disposed of as lower level Pu contaminated material. NNL and ANSTO have developed full-ceramic and glass-ceramic waste forms in which hot-isostatic pressing is used as the consolidation step to safely immobilize the waste into a form suitable for long-term disposition. We discuss development work on the glass-ceramic developed for impure waste streams, in particular the effect of variations in the waste feed chemistry glass-ceramic. The waste chemistry was categorized into actinides, impurity cations, glass formers and anions. Variations of the relative amounts of these on the properties and chemistry of the waste form were investigated and the waste form was found to be largely unaffected by these changes. This work mainly discusses the initial trials with Th and U. Later trials with larger variations and work with Pu-doped samples further confirmed the flexibility of the glass-ceramic. (authors)

  8. Vitrified waste form performance modeling applied to the treatment and disposal of mixed-waste sludge at the Savannah River Site

    International Nuclear Information System (INIS)

    Whited, A.R.; Fjeld, R.A.

    1998-01-01

    Vitrification, the conversion of source components into a solid amorphous glass matrix, has emerged as a viable treatment technology for low-level radioactive waste and mixed waste. To dispose of vitrified low-level waste at US Department of Energy facilities, site-specific radiological performance assessments must be conducted to demonstrate that waste glass satisfies performance objectives for environmental protection. More than 2,500 m 3 of F0006-listed low-level mixed-waste sludge stored in the Reactor Materials Department (M-Area) at the Savannah River Site (SRS) is scheduled for vitrification. This study evaluates the feasibility of on-site disposal of vitrified M-Area waste at SRS. Laboratory leaching tests that accelerate the glass corrosion process are currently the best indicators of vitrified waste form durability. A method to incorporate laboratory leaching data into performance assessments is presented. A screening-level performance assessments code is used to model trench disposal of M-Area waste glass. The allowable leach rate for vitrified M-Area waste is determined based on both a maximum radiological dose equivalent of 4 mrem/yr for the drinking water pathway and a maximum uranium concentration of 20 microg/ell in groundwater. The allowable leach rate is compared with published long-term leaching data for a wide range of waste glass compositions and test conditions. This analysis demonstrates that trench disposal of the waste glass is likely to meet applicable performance objectives if the glass is of above average durability compared with the reference set of glasses

  9. Diffusion processes in nuclear waste glasses

    International Nuclear Information System (INIS)

    Serruys, Y.; Limoge, Y.; Brebec, G.

    1992-01-01

    Problems concerning the containment of nuclear wastes are presented. Different materials which have been considered for this purpose are briefly reviewed and we see why glass is one of the favorite candidates. It is focussed on what is known about diffusion in 'simple enough' glasses. After a recall concerning the structure and possible defects, the main results on diffusion in 'simple' glasses are given and it is shown what these results involve for the mechanisms of diffusion. The diffusion models are presented which can account for transport in random media: percolation and random walk models. Specific phenomena for the nuclear waste glasses are considered: the effect of irradiation on diffusion and leaching (i.e. corrosion by water). Finally diffusion data in nuclear waste glasses are presented. (author). 199 refs., 6 figs., 1 tab

  10. Recommendations for Glass Durability Test Criteria

    International Nuclear Information System (INIS)

    Strachan, D.M.

    1999-01-01

    The objective of this short report is to define a set of activities that should lead to a specification for a test that can be used as one of the acceptance criteria and as an indicator of acceptable long-term behavior in contact with water. Since the glass composition is not yet defined and is likely to change as the composition of the waste changes, a strategy for developing criteria for what is acceptable involves a series of tests and modeling activities. The results of these activities lead to a criterion for an acceptable product and the data that are needed to reliably determine the behavior of the glass in the storage environment

  11. Glass-ceramics: Their production from wastes - a review

    Energy Technology Data Exchange (ETDEWEB)

    Rawlings, R.D.; Wu, J.P.; Boccaccini, A.R. [University of London, London (United Kingdom). Imperial College of Science & Technology, Dept. of Medicine

    2006-02-15

    Glass-ceramics are polycrystalline materials of fine microstructure that are produced by the controlled crystallisation (devitrification) of a glass. Numerous silicate based wastes, such as coal combustion ash, slag from steel production, fly ash and filter dusts from waste incinerators, mud from metal hydrometallurgy, different types of sludge as well as glass cullet or mixtures of them have been considered for the production of glass-ceramics. Developments of glass-ceramics from waste using different processing methods are described comprehensively in this review, covering R&D work carried out worldwide in the last 40 years. Properties and applications of the different glass-ceramics produced are discussed. The review reveals that considerable knowledge and expertise has been accumulated on the process of transformation of silicate waste into useful glass-ceramic products. These glass-ceramics are attractive as building materials for usage as construction and architectural components or for other specialised technical applications requiring a combination of suitable thermo-mechanical properties. Previous attempts to commercialise glass-ceramics from waste and to scale-up production for industrial exploitation are also discussed.

  12. DEFENSE HIGH LEVEL WASTE GLASS DEGRADATION

    International Nuclear Information System (INIS)

    Ebert, W.

    2001-01-01

    The purpose of this Analysis/Model Report (AMR) is to document the analyses that were done to develop models for radionuclide release from high-level waste (HLW) glass dissolution that can be integrated into performance assessment (PA) calculations conducted to support site recommendation and license application for the Yucca Mountain site. This report was developed in accordance with the ''Technical Work Plan for Waste Form Degradation Process Model Report for SR'' (CRWMS M andO 2000a). It specifically addresses the item, ''Defense High Level Waste Glass Degradation'', of the product technical work plan. The AP-3.15Q Attachment 1 screening criteria determines the importance for its intended use of the HLW glass model derived herein to be in the category ''Other Factors for the Postclosure Safety Case-Waste Form Performance'', and thus indicates that this factor does not contribute significantly to the postclosure safety strategy. Because the release of radionuclides from the glass will depend on the prior dissolution of the glass, the dissolution rate of the glass imposes an upper bound on the radionuclide release rate. The approach taken to provide a bound for the radionuclide release is to develop models that can be used to calculate the dissolution rate of waste glass when contacted by water in the disposal site. The release rate of a particular radionuclide can then be calculated by multiplying the glass dissolution rate by the mass fraction of that radionuclide in the glass and by the surface area of glass contacted by water. The scope includes consideration of the three modes by which water may contact waste glass in the disposal system: contact by humid air, dripping water, and immersion. The models for glass dissolution under these contact modes are all based on the rate expression for aqueous dissolution of borosilicate glasses. The mechanism and rate expression for aqueous dissolution are adequately understood; the analyses in this AMR were conducted to

  13. Laboratory testing of waste glass aqueous corrosion; effects of experimental parameters

    International Nuclear Information System (INIS)

    Ebert, W.L.; Mazer, J.J.

    1993-01-01

    A literature survey has been performed to assess the effects of the temperature, glass surface area/leachate volume ratio, leachant composition, leachant flow rate, and glass composition (actual radioactive vs. simulated glass) used in laboratory tests on the measured glass reaction rate. The effects of these parameters must be accounted for in mechanistic models used to project glass durability over long times. Test parameters can also be utilized to highlight particular processes in laboratory tests. Waste glass corrosion results as water diffusion, ion-exchange, and hydrolysis reactions occur simultaneously to devitrify the glass and release soluble glass components into solution. The rates of these processes are interrelated by the affects of the solution chemistry and glass alteration phases on each process, and the dominant (fastest) process may change as the reaction progresses. Transport of components from the release sites into solution may also affect the observed corrosion rate. The reaction temperature will affect the rate of each process, while other parameters will affect the solution chemistry and which processes are observed during the test. The early stages of corrosion will be observed under test conditions which maintain dilute leachates and the later stages will be observed under conditions that generate more concentrated leachate solutions. Typically, water diffusion and ion-exchange reactions dominate the observed glass corrosion in dilute solutions while hydrolysis reactions dominant in more concentrated solutions. Which process(es) controls the long-term glass corrosion is not fully understood, and the long-term corrosion rate may be either transport- or reaction-limited

  14. The encapsulation of nuclear waste in a magnesium aluminosilicate glass-ceramic

    International Nuclear Information System (INIS)

    Luk, K.M.

    1999-07-01

    The use of Magnesium aluminosilicate (MAS) glass-ceramics for the immobilisation of nuclear waste has been investigated. Nuclear waste is currently immobilised in a borosilicate glass. It is possible that immobilisation in an MAS glass-ceramic will reduce processing temperature of the waste, offer greater thermal and chemical stabilities and chemical durabilities. The primary reason for investigating sintered glass-ceramics is the possible advent of wastes containing high levels of refractory elements such as zirconia from the future reprocessing techniques such as electrochemical dissolution. In the first instance zirconia was used as a simulated waste with the principal of encapsulating zirconia with the minimum of porosity. Attempts were made to encapsulate 0, 20 and 40 volume % of zirconia in MAS sintering at temperatures of around 950 deg. C. It was found that the main cause of porosity was the agglomeration of fine zirconia powder. Three Taguchi experiments to optimise conditions for encapsulation of zirconia in MAS were carried out. In each case 10 volume % of zirconia was encapsulated. A Taguchi L 8 was carried out to optimise thermal conditions and powder characteristics. A Taguchi L 9 was carried out to improve knowledge of the thermal characteristics and an L 16 was carried out to provide information on curvature of thermal parameters and powder particle sizes. The conditions predicted to be optimum from these Taguchi experiments were a temperature of 940 - 960 deg. C, a heating rate of 30 deg. C/min, a hold time of 30 - 50 minutes and particle sizes of 2-4 and ∼ 15μm respectively. Densifications of up to 99% have been observed. Tapping experiments were carried out in an attempt to remove the pressing stage from processing. MAS was tapped into an alumina crucible with and without the addition of a dead weight. Almost fully dense MAS pellets were produced. This is an indication that it may be possible to process glass-ceramic waste forms in their final

  15. The role of frit in nuclear waste vitrification

    International Nuclear Information System (INIS)

    Vienna, J.D.; Smith, P.A.; Dorn, D.A.; Hrma, P.

    1994-04-01

    Vitrification of nuclear waste requires additives which are often vitrified independently to form a frit. Frit composition is formulated to meet the needs of glass composition and processing. The effects of frit on melter feed and melt processing, glass acceptance, and waste loading is of practical interest in understanding the trade-offs associated with the competing demands placed on frit composition. Melter feed yield stress, viscosity and durability of frits and corresponding waste glasses as well as the kinetics of elementary melting processes have been measured. The results illustrate the competing requirements on frit. Four frits (FY91, FY93, HW39-4, and SR202) and simulated neutralized current acid waste (NCAW) were used in this study. The experimental evidence shows that optimization of frit for one processing related property often results in poorer performance for the remaining properties. The difficulties associated with maximum waste loading and durability are elucidated for glasses which could be processed using technology available for the previously proposed Hanford Waste Vitrification Plant

  16. Insertion of marble waste in the production chain of glass wool; Insercao do residuo de marmore na cadeia produtiva da la de vidro

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, G.F.; Alves, J.O.; Espinosa, D.C.R.; Tenorio, J.A.S., E-mail: girleyf@gmail.co [Universidade de Sao Paulo (USP), SP (Brazil). Escola Politecnica. Dept. de Engenharia Metalurgica e de Materiais

    2010-07-01

    The work aimed the study of the recycle of the waste from marble cutting, aiming the reuse as partial raw material in the production of glass wool. Glass wool are materials with chemical and mechanical resistance, durability and lightness, and also important thermo-acoustic properties. A mixture of the waste with chemical additives was melted in a laboratory electric furnace using temperature of 1450 deg C. The melted material was directly poured in a water-filled recipient aiming the rapidly cooling. Samples of the produced material were characterized by XRD, SEM and DTA. The results showed that the residue from marble cutting can be inserted into the productive chain of glass wool, providing a decrease in the extraction of mineral resources, a profitable destination for this waste, and a economy for the companies producer of thermo-acoustic insulators. (author)

  17. Iron phosphate glass containing simulated fast reactor waste: Characterization and comparison with pristine iron phosphate glass

    International Nuclear Information System (INIS)

    Joseph, Kitheri; Asuvathraman, R.; Venkata Krishnan, R.; Ravindran, T.R.; Govindaraj, R.; Govindan Kutty, K.V.; Vasudeva Rao, P.R.

    2014-01-01

    Detailed characterization was carried out on an iron phosphate glass waste form containing 20 wt.% of a simulated nuclear waste. High temperature viscosity measurement was carried out by the rotating spindle method. The Fe 3+ /Fe ratio and structure of this waste loaded iron phosphate glass was investigated using Mössbauer and Raman spectroscopy respectively. Specific heat measurement was carried out in the temperature range of 300–700 K using differential scanning calorimeter. Isoconversional kinetic analysis was employed to understand the crystallization behavior of the waste loaded iron phosphate glass. The glass forming ability and glass stability of the waste loaded glass were also evaluated. All the measured properties of the waste loaded glass were compared with the characteristics of pristine iron phosphate glass

  18. Radionuclides containment in nuclear glasses. An overview

    Energy Technology Data Exchange (ETDEWEB)

    Gin, Stephane; Jollivet, Patrick; Tribet, Magaly; Peuget, Sylvain; Schuller, Sophie [CEA Marcoule, Bagnols-sur-Ceze (France). DE2D SEVT

    2017-07-01

    Radioactive waste vitrification has been carried out industrially in several countries for nearly 40 years. Research into the formulation and long term behavior of high and intermediate level waste glasses, mainly borosilicate compositions, is still continuing in order to (i) safely condition new types of wastes and (ii) design and demonstrate the safety of the disposal of these long-lived waste forms in a deep geological repository. This article presents a summary of current knowledge on the formulation, irradiation resistance and the chemical durability of these conditioning materials, with a special focus on the fate of radionuclides during glass processing and aging. It is shown that, apart from the situation for certain elements with very low incorporation rate in glass matrices, vitrification in borosilicate glass can enable waste loadings of up to ∝20 wt% while maintaining the glass homogeneity for geological time scales and guaranteeing a high stability level in spite of irradiation and water contact.

  19. Glass Ceramics Composites Fabricated from Coal Fly Ash and Waste Glass

    International Nuclear Information System (INIS)

    Angjusheva, B.; Jovanov, V.; Srebrenkoska, V.; Fidancevska, E.

    2014-01-01

    Great quantities of coal ash are produced in thermal power plants which present a double problem to the society: economical and environmental. This waste is a result of burning of coal at temperatures between 1100-14500C. Fly ash available as fine powder presents a source of important oxides SiO2, Al2O3, Fe2O3, MgO, Na2O, but also consist of small amount of ecologically hazardous oxides such as Cr2O3, NiO, MnO. The combination of the fly ash with waste glass under controlled sintering procedure gave bulk glass-ceramics composite material. The principle of this procedure is presented as a multi barrier concept. Many researches have been conducted the investigations for utilization of fly ash as starting material for various glass–ceramics production. Using waste glass ecologically hazardous components are fixed at the molecular level in the silicate phase and the fabricated new glass-ceramic composites possess significantly higher mechanical properties. The aim of this investigation was to fabricate dense glass ceramic composites using fly ash and waste glass with the potential for its utilization as building material

  20. Ancient Glass: A Literature Search and its Role in Waste Management

    Energy Technology Data Exchange (ETDEWEB)

    Strachan, Denis M.; Pierce, Eric M.

    2010-07-01

    When developing a performance assessment model for the long-term disposal of immobilized low-activity waste (ILAW) glass, it is desirable to determine the durability of glass forms over very long periods of time. However, testing is limited to short time spans, so experiments are performed under conditions that accelerate the key geochemical processes that control weathering. Verification that models currently being used can reliably calculate the long term behavior ILAW glass is a key component of the overall PA strategy. Therefore, Pacific Northwest National Laboratory was contracted by Washington River Protection Solutions, LLC to evaluate alternative strategies that can be used for PA source term model validation. One viable alternative strategy is the use of independent experimental data from archaeological studies of ancient or natural glass contained in the literature. These results represent a potential independent experiment that date back to approximately 3600 years ago or 1600 before the current era (bce) in the case of ancient glass and 106 years or older in the case of natural glass. The results of this literature review suggest that additional experimental data may be needed before the result from archaeological studies can be used as a tool for model validation of glass weathering and more specifically disposal facility performance. This is largely because none of the existing data set contains all of the information required to conduct PA source term calculations. For example, in many cases the sediments surrounding the glass was not collected and analyzed; therefore having the data required to compare computer simulations of concentration flux is not possible. This type of information is important to understanding the element release profile from the glass to the surrounding environment and provides a metric that can be used to calibrate source term models. Although useful, the available literature sources do not contain the required information

  1. Ancient Glass: A Literature Search and its Role in Waste Management

    International Nuclear Information System (INIS)

    Strachan, Denis M.; Pierce, Eric M.

    2010-01-01

    When developing a performance assessment model for the long-term disposal of immobilized low-activity waste (ILAW) glass, it is desirable to determine the durability of glass forms over very long periods of time. However, testing is limited to short time spans, so experiments are performed under conditions that accelerate the key geochemical processes that control weathering. Verification that models currently being used can reliably calculate the long term behavior ILAW glass is a key component of the overall PA strategy. Therefore, Pacific Northwest National Laboratory was contracted by Washington River Protection Solutions, LLC to evaluate alternative strategies that can be used for PA source term model validation. One viable alternative strategy is the use of independent experimental data from archaeological studies of ancient or natural glass contained in the literature. These results represent a potential independent experiment that date back to approximately 3600 years ago or 1600 before the current era (bce) in the case of ancient glass and 106 years or older in the case of natural glass. The results of this literature review suggest that additional experimental data may be needed before the result from archaeological studies can be used as a tool for model validation of glass weathering and more specifically disposal facility performance. This is largely because none of the existing data set contains all of the information required to conduct PA source term calculations. For example, in many cases the sediments surrounding the glass was not collected and analyzed; therefore having the data required to compare computer simulations of concentration flux is not possible. This type of information is important to understanding the element release profile from the glass to the surrounding environment and provides a metric that can be used to calibrate source term models. Although useful, the available literature sources do not contain the required information

  2. Effects of beta/gamma radiation on nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Weber, W.J. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-07-01

    A key challenge in the disposal of high-level nuclear waste (HLW) in glass waste forms is the development of models of long-term performance based on sound scientific understanding of relevant phenomena. Beta decay of fission products is one source of radiation that can impact the performance of HLW glasses through the interactions of the emitted {beta}-particles and g-rays with the atoms in the glass by ionization processes. Fused silica, alkali silicate glasses, alkali borosilicate glasses, and nuclear waste glasses are all susceptible to radiation effects from ionization. In simple glasses, defects (e.g., non-bridging oxygen and interstitial molecular oxygen) are observed experimentally. In more complex glasses, including nuclear waste glasses, similar defects are expected, and changes in microstructure, such as the formation of bubbles, have been reported. The current state of knowledge regarding the effects of {beta}/{gamma} radiation on the properties and microstructure of nuclear waste glasses are reviewed. (author)

  3. Effects of beta/gamma radiation on nuclear waste glasses

    International Nuclear Information System (INIS)

    Weber, W.J.

    1997-01-01

    A key challenge in the disposal of high-level nuclear waste (HLW) in glass waste forms is the development of models of long-term performance based on sound scientific understanding of relevant phenomena. Beta decay of fission products is one source of radiation that can impact the performance of HLW glasses through the interactions of the emitted β-particles and g-rays with the atoms in the glass by ionization processes. Fused silica, alkali silicate glasses, alkali borosilicate glasses, and nuclear waste glasses are all susceptible to radiation effects from ionization. In simple glasses, defects (e.g., non-bridging oxygen and interstitial molecular oxygen) are observed experimentally. In more complex glasses, including nuclear waste glasses, similar defects are expected, and changes in microstructure, such as the formation of bubbles, have been reported. The current state of knowledge regarding the effects of β/γ radiation on the properties and microstructure of nuclear waste glasses are reviewed. (author)

  4. Radiation effects in glass and glass-ceramic waste forms for the immobilization of CANDU UO2 fuel reprocessing waste

    International Nuclear Information System (INIS)

    Tait, J.C.

    1993-05-01

    AECL has investigated three waste forms for the immobilization of high-level liquid wastes that would arise if used CANDU fuels were reprocessed at some time in the future to remove fissile materials for the fabrication of new power reactor fuel. These waste forms are borosilicate glasses, aluminosilicate glasses and titanosilicate glass-ceramics. This report discusses the potential effects of alpha, beta and gamma radiation on the releases of radionuclides from these waste forms as a result of aqueous corrosion by groundwaters that would be present in an underground waste disposal vault. The report discusses solid-state damage caused by radiation-induced atomic displacements in the waste forms as well as irradiation of groundwater solutions (radiolysis), and their potential effects on waste-form corrosion and radionuclide release. The current literature on radiation effects on borosilicate glasses and in ceramics is briefly reviewed, as are potential radiation effects on specialized waste forms for the immobilization of 129 I, 85 Kr and 14 C. (author). 104 refs., 9 tabs., 5 figs

  5. Vitrification chemistry and nuclear waste

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1985-01-01

    The vitrification of nuclear waste offers unique challenges to the glass technologist. The waste contains 50 or 60 elements, and often varies widely in composition. Most of these elements are seldom encountered in processing commercial glasses. The melter to vitrify the waste must be able to tolerate these variations in composition, while producing a durable glass. This glass must be produced without releasing hazardous radionuclides to the environment during any step of the vitrification process. Construction of a facility to convert the nearly 30 million gallons of high-level nuclear waste at the Savannah River Plant into borosilicate glass began in late 1983. In developing the vitrification process, the Savannah River Laboratory has had to overcome all of these challenges to the glass technologist. Advances in understanding in three areas have been crucial to our success: oxidation-reduction phenomena during glass melting; the reaction between glass and natural wastes; and the causes of foaming during glass melting

  6. Glasses obtained from industrial wastes

    International Nuclear Information System (INIS)

    Bortoluzzi, D.; Oliveira Fillho, J.; Uggioni, E.; Bernardin, A.M.

    2009-01-01

    This paper deals with the study of the vitrification mechanism as an inertization method for industrial wastes contaminated with heavy metals. Ashes from coal (thermoelectric), wastes from mining (fluorite and feldspar) and plating residue were used to compose vitreous systems planed by mixture design. The chemical composition of the wastes was determined by XRF and the formulations were melted at 1450 deg C for 2h using 10%wt of CaCO 3 (fluxing agent). The glasses were poured into a mold and annealed (600 deg C). The characteristic temperatures were determined by thermal analysis (DTA, air, 20 deg C/min) and the mechanical behavior by Vickers microhardness. As a result, the melting temperature is strongly dependent on silica content of each glass, and the fluorite residue, being composed mainly by silica, strongly affects Tm. The microhardness of all glasses is mainly affected by the plating residue due to the high iron and zinc content of this waste. (author)

  7. Storage and disposal of radioactive waste as glass in canisters

    International Nuclear Information System (INIS)

    Mendel, J.E.

    1978-12-01

    A review of the use of waste glass for the immobilization of high-level radioactive waste glass is presented. Typical properties of the canisters used to contain the glass, and the waste glass, are described. Those properties are used to project the stability of canisterized waste glass through interim storage, transportation, and geologic disposal

  8. Leach behavior of high-level borosilicate glasses under deep geological environment

    International Nuclear Information System (INIS)

    Kim, Seung Soo; Chun, Kwan Sik; Park, Hyun Soo

    1998-02-01

    This report presents an overview of the activities in high-level radioactive waste glass which is considered as the most practicable form of waste, and also is intended to be used in the disposal of national high-level radioactive waste in future. Leach theory of waste glass and the leach effects of ground water, metal barrier, buffer materials and rocks on the waste glass were reviewed. The leach of waste glass was affected by various factors such as composition, pH and Eh of ground water, temperature, pressure, radiation and humic acid. The crystallization, crack, weathering and the formation of altered phases of waste glass which is expected to occur in real disposal site were reviewed. The results of leaching in laboratory and in-situ were compared. The behaviors of radioactive elements leached from waste glass and the use of basalt glass for the long-term natural analogue of waste glass were also written in this report. The appraisal of durability of borosilicate waste glass as a waste media was performed from the known results of leach test and international in-situ tests were introduced. (author). 134 refs., 15 tabs., 24 figs

  9. Method and apparatus for glass solidification porcessing for radioactive liquid waste

    International Nuclear Information System (INIS)

    Torada, Shin-ichiro; Masaki, Toshio; Sakai, Akira.

    1989-01-01

    Glass material supplied to a glass melting furnace is made in the form of a glass container. Then, radioactive liquid wastes are directly injected into the glass vessel and the glass vessel injected with the radioactive liquid wastes is charged into the glass melting furnace. The glass material and the radioactive liquid wastes are supplied simultaneously to the glass melting furnace. Then, corresponding to the amount of the glass material used for the glass vessel, the amount of the radioactive liquid wastes injected to the inside thereof is controlled to thereby set the mixing ratio between the glass material and the radioactive liquid wastes. Further, by controlling the number of the glass vessels injected with the radioactive liquid wastes to be charged into the glass melting furnace, the amount of supplying the radioactive liquid wastes and the glass material is controlled. This can easily maintain constant the amount of the glass material and the radioacative liquid wastes supplied to the glass melting furnace and the mixing ratio thereof. (T.M.)

  10. Development Of Glass Matrices For HLW Radioactive Wastes

    International Nuclear Information System (INIS)

    Jantzen, C.

    2010-01-01

    Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either borosilicate glass or phosphate glass. One of the primary reasons that glass has become the most widely used immobilization media is the relative simplicity of the vitrification process, e.g. melt waste plus glass forming frit additives and cast. A second reason that glass has become widely used for HLW is that the short range order (SRO) and medium range order (MRO) found in glass atomistically bonds the radionuclides and governs the melt properties such as viscosity, resistivity, sulphate solubility. The molecular structure of glass controls contaminant/radionuclide release by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. The molecular structure is flexible and hence accounts for the flexibility of glass formulations to waste variability. Nuclear waste glasses melt between 1050-1150 C which minimizes the volatility of radioactive components such as Tc 99 , Cs 137 , and I 129 . Nuclear waste glasses have good long term stability including irradiation resistance. Process control models based on the molecular structure of glass have been mechanistically derived and have been demonstrated to be accurate enough to control the world's largest HLW Joule heated ceramic melter in the US since 1996 at 95% confidence.

  11. Complexing agents and pH influence on chemical durability of type I moulded glass containers.

    Science.gov (United States)

    Biavati, Alberto; Poncini, Michele; Ferrarini, Arianna; Favaro, Nicola; Scarpa, Martina; Vallotto, Marta

    2017-06-16

    Among the factors that affect the glass surface chemical durability, pH and complexing agents presence in aqueous solution have the main role (1). Glass surface attack can be also related to the delamination issue with glass particles appearance in the pharmaceutical preparation. A few methods to check for glass containers delamination propensity and some control guidelines have been proposed (2,3). The present study emphasizes the possible synergy between a few complexing agents with pH on the borosilicate glass chemical durability. Hydrolytic attack was performed in small volume 23 ml type I glass containers autoclaved according to EP or USP for 1 hour at 121°C, in order to enhance the chemical attack due to time, temperature and the unfavourable surface/volume ratio. 0,048 M or 0.024 M (moles/liter) solutions of the acids citric, glutaric, acetic, EDTA (ethylenediaminetetraacetic acid) and sodium phosphate with water for comparison, were used for the trials. The pH was adjusted ± 0,05 units at fixed values 5,5-6,6-7-7,4-8-9 by LiOH diluted solution. Since silicon is the main glass network former, silicon release into the attack solutions was chosen as the main index of the glass surface attack and analysed by ICPAES. The work was completed by the analysis of the silicon release in the worst attack conditions, of moulded glass, soda lime type II and tubing borosilicate glass vials to compare different glass compositions and forming technologies. Surface analysis by SEM was finally performed to check for the surface status after the worst chemical attack condition by citric acid. Copyright © 2017, Parenteral Drug Association.

  12. Waste glass weathering

    International Nuclear Information System (INIS)

    Bates, J.K.; Buck, E.C.

    1994-01-01

    The weathering of glass is reviewed by examining processes that affect the reaction of commercial, historical, natural, and nuclear waste glass under conditions of contact with humid air and slowly dripping water, which may lead to immersion in nearly static solution. Radionuclide release data from weathered glass under conditions that may exist in an unsaturated environment are presented and compared to release under standard leaching conditions. While the comparison between the release under weathering and leaching conditions is not exact, due to variability of reaction in humid air, evidence is presented of radionuclide release under a variety of conditions. These results suggest that both the amount and form of radionuclide release can be affected by the weathering of glass

  13. Ceramic and glass radioactive waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Readey, D.W.; Cooley, C.R. (comps.)

    1977-01-01

    This report contains 14 individual presentations and 6 group reports on the subject of glass and polycrystalline ceramic radioactive waste forms. It was the general consensus that the information available on glass as a waste form provided a good basis for planning on the use of glass as an initial waste form, that crystalline ceramic forms could also be good waste forms if much more development work were completed, and that prediction of the chemical and physical stability of the waste form far into the future would be much improved if the basic synergistic effects of low temperature, radiation and long times were better understood. Continuing development of the polycrystalline ceramic forms was recommended. It was concluded that the leach rate of radioactive species from the waste form is an important criterion for evaluating its suitability, particularly for the time period before solidified waste is permanently placed in the geologic isolation of a Federal repository. Separate abstracts were prepared for 12 of the individual papers; the remaining two were previously abstracted.

  14. Thermal and physicochemical properties important for the long term behavior of nuclear waste glasses

    International Nuclear Information System (INIS)

    Vernaz, E.; Matzke, H.J.

    1992-01-01

    High level nuclear waste from reprocessing of spent nuclear fuel has to be solidified in a stable matrix for safe long-time storage. Vitrification in borosilicate glasses is the technique accepted worldwide. A number of different glasses was developed in different national programs. The criteria and the reasons for selecting the final compositions are briefly described. Emphasis is placed on the French product R7T7 and on thermal and physicochemical properties though glasses developed in other national projects (e.g. the German product GP 98/12 etc.) are also treated. The basic physical and mechanical properties and the chemical durability of the glass in contact with water or other aqueous solutions are described. The basic mechanisms of aqueous corrosion are discussed and the evolving modelling of the leaching process is dealt with, as well as effects of container material, backfill, etc. The thermal behavior has also been studied and extensive data exist on diffusion of glass constituents (Na) and of interesting elements of the waste such as the alkalis Rb and Cs or the actinides U and Pu, as well as on crystallization processes in the glass during storage at elevated temperatures. Emphasis is placed on the radiation stability of the glasses, based on extensive studies using short-lived actinides (e.g. Cm-244) or ion-implantation to produce the damage expected during long storage at an accelerated rate. The radiation stability is shown to be very good, if realistic damage conditions are used. The knowledge accumulated in the past years is used to evaluate and predict the long-term evolution of the glass under storage conditions

  15. Glassy slags as novel waste forms for remediating mixed wastes with high metal contents

    International Nuclear Information System (INIS)

    Feng, X.; Wronkiewicz, D.J.; Bates, J.K.; Brown, N.R.; Buck, E.C.; Gong, M.; Ebert, W.L.

    1994-01-01

    Argonne National Laboratory (ANL) is developing a glassy slag final waste form for the remediation of low-level radioactive and mixed wastes with high metal contents. This waste form is composed of various crystalline and metal oxide phases embedded in a silicate glass phase. This work indicates that glassy slag shows promise as final waste form because (1) it has similar or better chemical durability than high-level nuclear waste (HLW) glasses, (2) it can incorporate large amounts of metal wastes, (3) it can incorporate waste streams having low contents of flux components (boron and alkalis), (4) it has less stringent processing requirements (e.g., viscosity and electric conductivity) than glass waste forms, (5) its production can require little or no purchased additives, which can result in greater reduction in waste volume and overall treatment costs. By using glassy slag waste forms, minimum additive waste stabilization approach can be applied to a much wider range of waste streams than those amenable only to glass waste forms

  16. Eco-efficient waste glass recycling: Integrated waste management and green product development through LCA

    International Nuclear Information System (INIS)

    Blengini, Gian Andrea; Busto, Mirko; Fantoni, Moris; Fino, Debora

    2012-01-01

    Highlights: ► A new eco-efficient recycling route for post-consumer waste glass was implemented. ► Integrated waste management and industrial production are crucial to green products. ► Most of the waste glass rejects are sent back to the glass industry. ► Recovered co-products give more environmental gains than does avoided landfill. ► Energy intensive recycling must be limited to waste that cannot be closed-loop recycled. - Abstract: As part of the EU Life + NOVEDI project, a new eco-efficient recycling route has been implemented to maximise resources and energy recovery from post-consumer waste glass, through integrated waste management and industrial production. Life cycle assessment (LCA) has been used to identify engineering solutions to sustainability during the development of green building products. The new process and the related LCA are framed within a meaningful case of industrial symbiosis, where multiple waste streams are utilised in a multi-output industrial process. The input is a mix of rejected waste glass from conventional container glass recycling and waste special glass such as monitor glass, bulbs and glass fibres. The green building product is a recycled foam glass (RFG) to be used in high efficiency thermally insulating and lightweight concrete. The environmental gains have been contrasted against induced impacts and improvements have been proposed. Recovered co-products, such as glass fragments/powders, plastics and metals, correspond to environmental gains that are higher than those related to landfill avoidance, whereas the latter is cancelled due to increased transportation distances. In accordance to an eco-efficiency principle, it has been highlighted that recourse to highly energy intensive recycling should be limited to waste that cannot be closed-loop recycled.

  17. Synroc tailored waste forms for actinide immobilization

    Energy Technology Data Exchange (ETDEWEB)

    Gregg, Daniel J.; Vance, Eric R. [Australian Nuclear Science and Technology Organisation, Kirrawee (Australia). ANSTOsynroc, Inst. of Materials Engineering

    2017-07-01

    Since the end of the 1970s, Synroc at the Australian Nuclear Science and Technology Organisation (ANSTO) has evolved from a focus on titanate ceramics directed at PUREX waste to a platform waste treatment technology to fabricate tailored glass-ceramic and ceramic waste forms for different types of actinide, high- and intermediate level wastes. The particular emphasis for Synroc is on wastes which are problematic for glass matrices or existing vitrification process technologies. In particular, nuclear wastes containing actinides, notably plutonium, pose a unique set of requirements for a waste form, which Synroc ceramic and glass-ceramic waste forms can be tailored to meet. Key aspects to waste form design include maximising the waste loading, producing a chemically durable product, maintaining flexibility to accommodate waste variations, a proliferation resistance to prevent theft and diversion, and appropriate process technology to produce waste forms that meet requirements for actinide waste streams. Synroc waste forms incorporate the actinides within mineral phases, producing products which are much more durable in water than baseline borosilicate glasses. Further, Synroc waste forms can incorporate neutron absorbers and {sup 238}U which provide criticality control both during processing and whilst within the repository. Synroc waste forms offer proliferation resistance advantages over baseline borosilicate glasses as it is much more difficult to retrieve the actinide and they can reduce the radiation dose to workers compared to borosilicate glasses. Major research and development into Synroc at ANSTO over the past 40 years has included the development of waste forms for excess weapons plutonium immobilization in collaboration with the US and for impure plutonium residues in collaboration with the UK, as examples. With a waste loading of 40-50 wt.%, Synroc would also be considered a strong candidate as an engineered waste form for used nuclear fuel and highly

  18. Initial demonstration of DWPF process and product control strategy using actual radioactive waste

    International Nuclear Information System (INIS)

    Andrews, M.K.; Bibler, N.E.; Jantzen, C.M.; Beam, D.C.

    1991-01-01

    The Defense Waste Processing Facility at the Savannah River Site (SRS) will vitrify high-level nuclear waste into borosilicate glass. The waste will be mixed with properly formulated glass-making frit and fed to a melter at 1150 degrees C. Process control and product quality are ensured by proper control of the melter feed composition. Algorithms have been developed to predict the processability of the melt and the durability of the final glass based on this feed composition. To test these algorithms, an actual radioactive waste contained in a shielded facility at SRS was analyzed and a frit composition formulated using a simple computer spreadsheet which contained the algorithms. This frit was then mixed with the waste and the resulting slurry fed to a research scale joule-heated melter operated remotely. Approximately 24 kg of glass were successfully prepared. This paper will describe the frit formulation, the vitrification process, and the glass durability

  19. Glass to contain wastes

    International Nuclear Information System (INIS)

    Moncouyoux, M.; Jacquet-Francillon, M.

    1994-01-01

    Here are the tables and figures presented during the conference on the glass to confine high level radioactive wastes: definition, fabrication, storage and disposal. The composition of glasses are detailed, their properties and the vitrification proceeding. The behaviour of these glasses in front of water, irradiation and heat are shown. The characteristics of parcels are given according to the radiation protection rule, ALARA principle, the concept of multi-barriers and the geological stability

  20. Nuclear hazardous waste cost control management

    International Nuclear Information System (INIS)

    Selg, R.A.

    1991-01-01

    The effects of the waste content of glass waste forms on Savannah River high-level waste disposal costs are currently under study to adjust the glass frit content to optimize the glass waste loadings and therefore significantly reduce the overall waste disposal cost. Changes in waste content affect onsite Defense Waste Changes in waste contents affect onsite Defense Waste Processing Facility (DWPF) costs as well as offsite shipping and repository emplacement charges. A nominal 1% increase over the 28 wt% waste loading of DWPF glass would reduce disposal costs by about $50 million for Savannah River wastes generated to the year 2000. Optimization of the glass waste forms to be produced in the SWPF is being supported by economic evaluations of the impact of the forms on waste disposal costs. Glass compositions are specified for acceptable melt processing and durability characteristics, with economic effects tracked by the number of waste canisters produced. This paper presents an evaluation of the effects of variations in waste content of the glass waste forms on the overall cost of the disposal, including offsite shipment and repository emplacement, of the Savannah River high-level wastes

  1. DEVELOPMENT OF GLASS MATRICES FOR HLW RADIOACTIVE WASTES

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C.

    2010-03-18

    Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either borosilicate glass or phosphate glass. One of the primary reasons that glass has become the most widely used immobilization media is the relative simplicity of the vitrification process, e.g. melt waste plus glass forming frit additives and cast. A second reason that glass has become widely used for HLW is that the short range order (SRO) and medium range order (MRO) found in glass atomistically bonds the radionuclides and governs the melt properties such as viscosity, resistivity, sulphate solubility. The molecular structure of glass controls contaminant/radionuclide release by establishing the distribution of ion exchange sites, hydrolysis sites, and the access of water to those sites. The molecular structure is flexible and hence accounts for the flexibility of glass formulations to waste variability. Nuclear waste glasses melt between 1050-1150 C which minimizes the volatility of radioactive components such as Tc{sup 99}, Cs{sup 137}, and I{sup 129}. Nuclear waste glasses have good long term stability including irradiation resistance. Process control models based on the molecular structure of glass have been mechanistically derived and have been demonstrated to be accurate enough to control the world's largest HLW Joule heated ceramic melter in the US since 1996 at 95% confidence.

  2. Glass forms for immobilization of Hanford wastes

    International Nuclear Information System (INIS)

    Schulz, W.W.; Dressen, A.L.; Hobbick, C.W.; Babad, H.

    1975-03-01

    Approximately 140 million liters of solid salt cake (mainly NaNO 3 ), produced by evaporation of aged alkaline high-level liquid wastes, will be stored in underground tanks when the present Hanford Waste Management Program is completed in the early 1980's. At this time also, large volumes of various other solid radioactive wastes (sludges, excavated Pu-contaminated soil, and doubly encapsulated 137 CsCl and 90 SrF 2 ) will be stored on the Hanford Reservation. All these solid wastes can be converted to immobile silicate and aluminosilicate glasses of low water leachability by melting them at 1100 0 to 1400 0 C with appropriate amounts of basalt (or sand) and other glass-formers such as B 2 O 3 or CaO. Reviewed in this paper are formulations and other melt conditions used successfully in batch tests to make glasses from actual and synthetic wastes; leachability and other properties of these glasses show them to be satisfactory vehicles for immobilization of the Hanford wastes. (U.S.)

  3. Radiation effects in glass and glass-ceramic waste forms for the immobilization of CANDU UO{sub 2} fuel reprocessing waste

    Energy Technology Data Exchange (ETDEWEB)

    Tait, J C

    1993-05-01

    AECL has investigated three waste forms for the immobilization of high-level liquid wastes that would arise if used CANDU fuels were reprocessed at some time in the future to remove fissile materials for the fabrication of new power reactor fuel. These waste forms are borosilicate glasses, aluminosilicate glasses and titanosilicate glass-ceramics. This report discusses the potential effects of alpha, beta and gamma radiation on the releases of radionuclides from these waste forms as a result of aqueous corrosion by groundwaters that would be present in an underground waste disposal vault. The report discusses solid-state damage caused by radiation-induced atomic displacements in the waste forms as well as irradiation of groundwater solutions (radiolysis), and their potential effects on waste-form corrosion and radionuclide release. The current literature on radiation effects on borosilicate glasses and in ceramics is briefly reviewed, as are potential radiation effects on specialized waste forms for the immobilization of {sup 129}I, {sup 85}Kr and {sup 14}C. (author). 104 refs., 9 tabs., 5 figs.

  4. HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASSES FOR HANFORD'S WTP (WASTE TREATMENT PROJECT)

    International Nuclear Information System (INIS)

    Kruger, A.A.; Bowan, B.W.; Joseph, I.; Gan, H.; Kot, W.K.; Matlack, K.S.; Pegg, I.L.

    2010-01-01

    This paper presents the results of glass formulation development and melter testing to identify high waste loading glasses to treat high-Al high level waste (HLW) at Hanford. Previous glass formulations developed for this HLW had high waste loadings but their processing rates were lower that desired. The present work was aimed at improving the glass processing rate while maintaining high waste loadings. Glass formulations were designed, prepared at crucible-scale and characterized to determine their properties relevant to processing and product quality. Glass formulations that met these requirements were screened for melt rates using small-scale tests. The small-scale melt rate screening included vertical gradient furnace (VGF) and direct feed consumption (DFC) melter tests. Based on the results of these tests, modified glass formulations were developed and selected for larger scale melter tests to determine their processing rate. Melter tests were conducted on the DuraMelter 100 (DMIOO) with a melt surface area of 0.11 m 2 and the DuraMelter 1200 (DMI200) HLW Pilot Melter with a melt surface area of 1.2 m 2 . The newly developed glass formulations had waste loadings as high as 50 wt%, with corresponding Al 2 O 3 concentration in the glass of 26.63 wt%. The new glass formulations showed glass production rates as high as 1900 kg/(m 2 .day) under nominal melter operating conditions. The demonstrated glass production rates are much higher than the current requirement of 800 kg/(m 2 .day) and anticipated future enhanced Hanford Tank Waste Treatment and Immobilization Plant (WTP) requirement of 1000 kg/(m 2 .day).

  5. Production of highly porous glass-ceramics from metallurgical slag, fly ash and waste glass

    OpenAIRE

    Mangutova Bianka V.; Fidancevska Emilija M.; Milosevski Milosav I.; Bossert Joerg H.

    2004-01-01

    Glass-ceramics composites were produced based on fly-ash obtained from coal power stations, metallurgical slag from ferronickel industry and waste glass from TV monitors, windows and flasks. Using 50% waste flask glass in combination with fly ash and 20% waste glass from TV screens in combination with slag, E-modulus and bending strength values of the designed systems are increased (system based on fly ash: E-modulus from 6 to 29 GPa, and bending strength from 9 to 75 MPa). The polyurethane f...

  6. The effect of vitrification technology on waste loading

    International Nuclear Information System (INIS)

    Hrma, P.R.; Smith, P.A.

    1994-08-01

    Radioactive wastes on the Hanford Site are going to be permanently disposed of by incorporation into a durable glass. These wastes will be separated into low and high-level portions, and then vitrified. The low-level waste (LLW) is water soluble. Its vitrifiable part (other than off-gas) contains approximately 80 wt% Na 2 O, the rest being Al 2 O 3 , P 2 O 5 , K 2 O, and minor components. The challenge is to formulate durable LLW glasses with as high Na 2 O content as possible by optimizing the additions of SiO 2 , Al 2 O 3 , B 2 O 3 , CaO, and ZrO 2 . This task will not be simple, considering the non-linear and interactive nature of glass properties as a function of composition. Once developed, the LLW glass, being similar in composition to commercial glasses, is unlikely to cause major processing problems, such as crystallization or molten salt segregation. For example, inexpensive LLW glass can be produced in a high-capacity Joule-heated melter with a cold cap to minimize volatilization. The high-level waste (HLW) consists of water-insoluble sludge (Fe 2 O 3 , Al 2 O 3 , ZrO 2 , Cr 2 O 3 , NiO, and others) and a substantial water-soluble residue (Na 2 O). Most of the water-insoluble components are refractory; i.e., their melting points are above the glass melting temperature. With regard to product acceptability, the maximum loading of Hanford HLW in the glass is limited by product durability, not by radiolytic heat generation. However, this maximum may not be achievable because of technological constraints imposed by melter feed rheology, frit properties, and glass melter limits. These restrictions are discussed in this paper. 38 refs

  7. Leaching behavior of glass ceramic nuclear waste forms

    International Nuclear Information System (INIS)

    Lokken, R.O.

    1981-11-01

    Glass ceramic waste forms have been investigated as alternatives to borosilicate glasses for the immobilization of high-level radioactive waste at Pacific Northwest Laboratory (PNL). Three glass ceramic systems were investigated, including basalt, celsian, and fresnoite, each containing 20 wt % simulated high-level waste calcine. Static leach tests were performed on seven glass ceramic materials and one parent glass (before recrystallization). Samples were leached at 90 0 C for 3 to 28 days in deionized water and silicate water. The results, expressed in normalized elemental mass loss, (g/m 2 ), show comparable releases from celsian and fresnoite glass ceramics. Basalt glass ceramics demonstrated the lowest normalized elemental losses with a nominal release less than 2 g/m 2 when leached in polypropylene containers. The releases from basalt glass ceramics when leached in silicate water were nearly identical with those in deionized water. The overall leachability of celsian and fresnoite glass ceramics was improved when silicate water was used as the leachant

  8. The Impact of the Proposed delta Gp Limits on Glass Formulation Efforts: Part II. Experimental Results

    International Nuclear Information System (INIS)

    PEELER, DAVID

    2004-01-01

    The Savannah River National Laboratory (SRNL) has initiated studies to assess alternative durability options that may provide access to compositional regions of interest in support of the accelerated cleanup mission at the Defense Waste Processing Facility (DWPF). One of the options being pursued is the redefinition of the durability model acceptability limits. Some of the conservative steps used in establishing the current limits without compromising the high confidence required for meeting the specification on the waste form quality were identified and eliminated. The results led to a set of three new Property Acceptability Region (PAR) values for the preliminary glass dissolution estimator that has the potential to allow access to compositional regions of interest to improve melt rate or waste loading. Although these limits are available for implementation, there is no driving force to do so with the current sludge batch (i.e., the current Frit 418 - Sludge Batch 3 (SB3) system is TL limited). The objectives of this task were to investigate (and generate) the incentive of applying the proposed durability limits in the Product Composition Control System (PCCS) from a glass formulation perspective. Glass compositions were identified or developed to transition into and through the region of GP acceptability as defined by the current and proposed durability limits. The progression through the newly defined acceptability region was accomplished by increasing the total alkali in the glass via higher alkali frits and/or waste loading (WL). The focus of this report is on the measured durability response as it compares to model predictions to assess the applicability and/or potential conservatism of the various limits or durability approaches. The normalized boron release values (NL [B] g/L) for the study glasses ranged from approximately 1.0 g/L to 2.0 g/L. The Product Consistency Test (PCT) responses provide evidence that implementation of the proposed GP limits will

  9. The durability of fired brick incorporating textile factory waste ash and basaltic pumice

    Energy Technology Data Exchange (ETDEWEB)

    Binici, Hanifi [Kahramanmaras Sutcu Imam Univ., Kahramanmaras (Turkey). Dept. of Civil Engineering; Yardim, Yavuz [Epoka Univ., Tirana (Albania). Dept. of Civil Engineering

    2012-07-15

    This study investigates the durability of fired brick produced with additives of textile factories' waste ash and basaltic pumice. The effects of incorporating waste ash and basaltic pumice on durability and mechanical properties of the clay bricks were studied. Samples were produced with different ratios of the textile factories' waste ash and basaltic pumice added and at different fire temperatures of 700, 900, and 1 050 C for 8 h. The bricks with additives were produced by adding equal amounts of textile factories' waste ash and basaltic pumice, separately and together, with rates of 5, 10 and 20 wt.%. The produced samples were kept one year in sodium sulphate and sodium nitrate and tested under freezing - unfreezing and drying - wetting conditions. Then compression strength and mass loss of the samples with and without additives were investigated. The test results were compared with standards and results obtained from control specimens. The results showed that incorporations up to 10 wt.% of textile factories' waste ash and basaltic pumice is beneficial to the fired brick. Both textile factories' waste ash and basaltic pumice were suitable additives and could be used for more durable clay brick production at 900 C fire temperature. (orig.)

  10. Advanced High-Level Waste Glass Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Peeler, David K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vienna, John D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schweiger, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fox, Kevin M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-01

    The U.S. Department of Energy Office of River Protection (ORP) has implemented an integrated program to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. The integrated ORP program is focused on providing a technical, science-based foundation from which key decisions can be made regarding the successful operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) facilities. The fundamental data stemming from this program will support development of advanced glass formulations, key process control models, and tactical processing strategies to ensure safe and successful operations for both the low-activity waste (LAW) and high-level waste (HLW) vitrification facilities with an appreciation toward reducing overall mission life. The purpose of this advanced HLW glass research and development plan is to identify the near-, mid-, and longer-term research and development activities required to develop and validate advanced HLW glasses and their associated models to support facility operations at WTP, including both direct feed and full pretreatment flowsheets. This plan also integrates technical support of facility operations and waste qualification activities to show the interdependence of these activities with the advanced waste glass (AWG) program to support the full WTP mission. Figure ES-1 shows these key ORP programmatic activities and their interfaces with both WTP facility operations and qualification needs. The plan is a living document that will be updated to reflect key advancements and mission strategy changes. The research outlined here is motivated by the potential for substantial economic benefits (e.g., significant increases in waste throughput and reductions in glass volumes) that will be realized when advancements in glass formulation continue and models supporting facility operations are implemented. Developing and applying advanced

  11. Glass–water interaction: Effect of high-valence cations on glass structure and chemical durability

    Energy Technology Data Exchange (ETDEWEB)

    Hopf, J.; Kerisit, Sebastien N.; Angeli, F.; Charpentier, Thibault M.; Icenhower, Jonathan P.; McGrail, Bernard P.; Windisch, Charles F.; Burton, Sarah D.; Pierce, Eric M.

    2016-05-15

    Borosilicate glass is a durable solid, but it dissolves when in contact with aqueous fluids. The dissolution mechanism, which involves a variety of sequential reactions that occur at the solid-fluid interface, has important implications for the corrosion resistance of industrial and nuclear waste glasses. In this study, spectroscopic measurements, dissolution experiments, and Monte Carlo simulations were performed to investigate the effect of high–valence cations (HVC) on the mechanisms of glass dissolution under dilute and near-saturated conditions. Raman and NMR spectroscopy were used to determine the structural changes that occur in glass, specifically network formers (e.g., Al, Si, and B), with the addition of the HVC element hafnium in the Na2O–Al2O3–B2O3–HfO2–SiO2 system (e.g., Na/[Al+B] = 1.0 and HfO2/SiO2 from 0.0 to 0.42). Spectroscopic measurements revealed that increasing hafnium content decreases N4 (tetrahedral boron/total boron) and increases the amount of Si—O—Hf moieties in the glass. Results from flow–through experiments conducted under dilute and near–saturated conditions show a decrease of approximately 100× or more in the dissolution rate over the series from 0 to 20 mol% HfO2. Comparing the average steady-state rates obtained under dilute conditions to the rates obtained for near-saturated conditions reveals a divergence in the magnitude between the average steady state rates measured in these different conditions. The reason for this divergence was investigated more thoroughly using Monte Carlo simulations. Simulations indicate that the divergence in glass dissolution behavior under dilute and near-saturated conditions result from the stronger binding of Si sites that deposit on the surface from the influent when Hf is present in the glass. As a result, the residence time at the glass surface of these newly-formed Si sites is longer in the presence of Hf, which increases the density of anchor sites from which altered layers

  12. Eco-efficient waste glass recycling: Integrated waste management and green product development through LCA.

    Science.gov (United States)

    Blengini, Gian Andrea; Busto, Mirko; Fantoni, Moris; Fino, Debora

    2012-05-01

    As part of the EU Life + NOVEDI project, a new eco-efficient recycling route has been implemented to maximise resources and energy recovery from post-consumer waste glass, through integrated waste management and industrial production. Life cycle assessment (LCA) has been used to identify engineering solutions to sustainability during the development of green building products. The new process and the related LCA are framed within a meaningful case of industrial symbiosis, where multiple waste streams are utilised in a multi-output industrial process. The input is a mix of rejected waste glass from conventional container glass recycling and waste special glass such as monitor glass, bulbs and glass fibres. The green building product is a recycled foam glass (RFG) to be used in high efficiency thermally insulating and lightweight concrete. The environmental gains have been contrasted against induced impacts and improvements have been proposed. Recovered co-products, such as glass fragments/powders, plastics and metals, correspond to environmental gains that are higher than those related to landfill avoidance, whereas the latter is cancelled due to increased transportation distances. In accordance to an eco-efficiency principle, it has been highlighted that recourse to highly energy intensive recycling should be limited to waste that cannot be closed-loop recycled. Copyright © 2011 Elsevier Ltd. All rights reserved.

  13. Assessment of Savannah River borosilicate glass in the repository environment

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Wicks, G.G.; Bibler, N.E.

    1982-04-01

    Since 1973, borosilicate glass has been studied as a matrix for the immobilization of high-level radioactive waste generated at the Savannah River Plant (SRP). In 1977, efforts began to develop and test the large-scale equipment necessary to convert the alkaline waste slurries at SRP into a durable borosilicate glass. A process has now been developed for the proposed Defense Waste Processing Facility (DWPF) which will annually produce approximately 500 canisters of SRP waste glass which will be stored on an interim basis on the Savannah River site. Current national policy calls for the permanent disposal of high-level waste in deep geologic repositories. In the repository environment, SRP waste glass will eventually be exposed to such stresses as lithostatic or hydrostatic pressures, radiation fields, and self-heating due to radioactive decay. In addition, producing and handling each canister of glass will also expose the glass to thermal and mechanical stresses. An important objective of the extensive glass characterization and testing programs of the Savannah River Laboratory (SRL) has been to determine how these stresses affect the performance of SRP waste glass. The results of these programs indicate that: these stresses will not significantly affect the performance of borosilicate glass containing SRP waste; and SRP waste glass will effectively immobilize hazardous radionuclides in the repository environment

  14. Direction of CRT waste glass processing: Electronics recycling industry communication

    International Nuclear Information System (INIS)

    Mueller, Julia R.; Boehm, Michael W.; Drummond, Charles

    2012-01-01

    Highlights: ► Given a large flow rate of CRT glass ∼10% of the panel glass stream will be leaded. ► The supply of CRT waste glass exceeded demand in 2009. ► Recyclers should use UV-light to detect lead oxide during the separation process. ► Recycling market analysis techniques and results are given for CRT glass. ► Academic initiatives and the necessary expansion of novel product markets are discussed. - Abstract: Cathode Ray Tube, CRT, waste glass recycling has plagued glass manufacturers, electronics recyclers and electronics waste policy makers for decades because the total supply of waste glass exceeds demand, and the formulations of CRT glass are ill suited for most reuse options. The solutions are to separate the undesirable components (e.g. lead oxide) in the waste and create demand for new products. Achieving this is no simple feat, however, as there are many obstacles: limited knowledge of waste glass composition; limited automation in the recycling process; transportation of recycled material; and a weak and underdeveloped market. Thus one of the main goals of this paper is to advise electronic glass recyclers on how to best manage a diverse supply of glass waste and successfully market to end users. Further, this paper offers future directions for academic and industry research. To develop the recommendations offered here, a combination of approaches were used: (1) a thorough study of historic trends in CRT glass chemistry; (2) bulk glass collection and analysis of cullet from a large-scale glass recycler; (3) conversations with industry members and a review of potential applications; and (4) evaluation of the economic viability of specific uses for recycled CRT glass. If academia and industry can solve these problems (for example by creating a database of composition organized by manufacturer and glass source) then the reuse of CRT glass can be increased.

  15. Turning nuclear waste into glass

    Energy Technology Data Exchange (ETDEWEB)

    Pegg, Ian L.

    2015-02-15

    Vitrification has emerged as the treatment option of choice for the most dangerous radioactive waste. But dealing with the nuclear waste legacy of the Cold War will require state-of-the-art facilities and advanced glass formulations.

  16. Spinel dissolution via addition of glass forming chemicals. Results of preliminary experiments

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States); Johnson, F. C. [Savannah River Site (SRS), Aiken, SC (United States)

    2015-11-01

    Increased loading of high level waste in glass can lead to crystallization within the glass. Some crystalline species, such as spinel, have no practical impact on the chemical durability of the glass, and therefore may be acceptable from both a processing and a product performance standpoint. In order to operate a melter with a controlled amount of crystallization, options must be developed for remediating an unacceptable accumulation of crystals. This report describes preliminary experiments designed to evaluate the ability to dissolve spinel crystals in simulated waste glass melts via the addition of glass forming chemicals (GFCs).

  17. Development and characterization of solidified forms for high-level wastes: 1978. Annual report

    International Nuclear Information System (INIS)

    Ross, W.A.; Mendel, J.E.

    1979-12-01

    Development and characterization of solidified high-level waste forms are directed at determining both process properties and long-term behaviors of various solidified high-level waste forms in aqueous, thermal, and radiation environments. Waste glass properties measured as a function of composition were melt viscosity, melt electrical conductivity, devitrification, and chemical durability. The alkali metals were found to have the greatest effect upon glass properties. Titanium caused a slight decrease in viscosity and a significant increase in chemical durability in acidic solutions (pH-4). Aluminum, nickel and iron were all found to increase the formation of nickel-ferrite spinel crystals in the glass. Four multibarrier advanced waste forms were produced on a one-liter scale with simulated waste and characterized. Glass marbles encapsulated in a vacuum-cast lead alloy provided improved inertness with a minimal increase in technological complexity. Supercalcine spheres exhibited excellent inertness when coated with pyrolytic carbon and alumina and put in a metal matrix, but the processing requirements are quite complex. Tests on simulated and actual high-level waste glasses continue to suggest that thermal devitrification has a relatively small effect upon mechanical and chemical durabilities. Tests on the effects radiation has upon waste forms also continue to show changes to be relatively insignificant. Effects caused by decay of actinides can be estimated to saturate at near 10 19 alpha-events/cm 3 in homogeneous solids. Actually, in solidified waste forms the effects are usually observed around certain crystals as radiation causes amorphization and swelling of th crystals

  18. Iron Phosphate Glasses: An Alternative for Vitrifying Certain Nuclear Wastes

    Energy Technology Data Exchange (ETDEWEB)

    Delbert E. Day; Chandra S. Ray; Cheol-Woon Kim

    2004-12-28

    Vitrification of nuclear waste in a glass is currently the preferred process for waste disposal. DOE currently approves only borosilicate (BS) type glasses for such purposes. However, many nuclear wastes, presently awaiting disposal, have complex and diverse chemical compositions, and often contain components that are poorly soluble or chemically incompatible in BS glasses. Such problematic wastes can be pre-processed and/or diluted to compensate for their incompatibility with a BS glass matrix, but both of these solutions increases the wasteform volume and the overall cost for vitrification. Direct vitrification using alternative glasses that utilize the major components already present in the waste is preferable, since it avoids pre-treating or diluting the waste, and, thus, minimizes the wasteform volume and overall cost.

  19. Iron Phosphate Glasses: An Alternative for Vitrifying Certain Nuclear Wastes

    International Nuclear Information System (INIS)

    Day, Delbert E.; Ray, Chandra S.; Cheol-Woon Kim

    2004-01-01

    Vitrification of nuclear waste in a glass is currently the preferred process for waste disposal. DOE currently approves only borosilicate (BS) type glasses for such purposes. However, many nuclear wastes, presently awaiting disposal, have complex and diverse chemical compositions, and often contain components that are poorly soluble or chemically incompatible in BS glasses. Such problematic wastes can be pre-processed and/or diluted to compensate for their incompatibility with a BS glass matrix, but both of these solutions increases the wasteform volume and the overall cost for vitrification. Direct vitrification using alternative glasses that utilize the major components already present in the waste is preferable, since it avoids pre-treating or diluting the waste, and, thus, minimizes the wasteform volume and overall cost

  20. Decontamination of Savannah River Plant waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1982-01-01

    A Defense Waste Processing Facility (DWPF) is currently being designed to convert Savannah River Plant (SRP) liquid, high-level radioactive waste into a solid form, such as borosilicate glass. The outside of the canisters of waste glass must have very low levels of smearable radioactive contamination before they are removed from the DWPF to prevent the spread of radioactivity. Several techniques were considered for canister decontamination: high-pressure water spray, electropolishing, chemical dissolution, and abrasive blasting. An abrasive blasting technique using a glass frit slurry has been selected for use in the DWPF. No additional equipment is needed to process waste generated from decontamination. Frit used as the abrasive will be mixed with the waste and fed to the glass melter. In contrast, chemical and electrochemical techniques require more space in the DWPF, and produce large amounts of contaminated byproducts which are difficult to immobilize by vitrification

  1. Direction of CRT waste glass processing: electronics recycling industry communication.

    Science.gov (United States)

    Mueller, Julia R; Boehm, Michael W; Drummond, Charles

    2012-08-01

    Cathode Ray Tube, CRT, waste glass recycling has plagued glass manufacturers, electronics recyclers and electronics waste policy makers for decades because the total supply of waste glass exceeds demand, and the formulations of CRT glass are ill suited for most reuse options. The solutions are to separate the undesirable components (e.g. lead oxide) in the waste and create demand for new products. Achieving this is no simple feat, however, as there are many obstacles: limited knowledge of waste glass composition; limited automation in the recycling process; transportation of recycled material; and a weak and underdeveloped market. Thus one of the main goals of this paper is to advise electronic glass recyclers on how to best manage a diverse supply of glass waste and successfully market to end users. Further, this paper offers future directions for academic and industry research. To develop the recommendations offered here, a combination of approaches were used: (1) a thorough study of historic trends in CRT glass chemistry; (2) bulk glass collection and analysis of cullet from a large-scale glass recycler; (3) conversations with industry members and a review of potential applications; and (4) evaluation of the economic viability of specific uses for recycled CRT glass. If academia and industry can solve these problems (for example by creating a database of composition organized by manufacturer and glass source) then the reuse of CRT glass can be increased. Copyright © 2012 Elsevier Ltd. All rights reserved.

  2. Plan for glass waste form testing for NNWSI [Nevada Nuclear Waste Storage Investigations

    International Nuclear Information System (INIS)

    Aines, R.D.

    1987-09-01

    The purpose of glass waste form testing is to determine the rate of release of radionuclides from breached glass waste containers. This information will be used to qualify glass waste forms with respect to the release requirements. It will be the basis of the source term from glass waste for repository performance assessment modeling. This information will also serve as part of the source term in the calculation of cumulative releases after 100,000 years in the site evaluation process. It will also serve as part of the source term input for calculation of cumulative releases to the accessible environment for 10,000 years after disposal, to determine compliance with EPA regulations. This investigation will provide data to resolve information needs. Information about the waste forms which is provided by the producer will be accumulated and evaluated; the waste form will be tested, properties determined, and mechanisms of degradation determined; and models providing long-term evaluation of release rates designed and tested. 23 refs

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

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Fu, E-mail: wangfu@swust.edu.cn; Liao, Qilong, E-mail: liaoqilong@swust.edu.cn; Dai, Yunya; Zhu, Hanzhen

    2016-08-15

    Immobilization of gadolinium (Gd), a nonradioactive surrogate for Pu{sup 3+}, in iron borophosphate glasses/glass-ceramics (IBP glasses/glass-ceramics) has been investigated. The IBP glass containing 4 mol% Gd{sub 2}O{sub 3} is homogeneously amorphous. At higher Gd{sub 2}O{sub 3} concentrations, additional Gd is retained in the glasses as crystalline inclusions of monazite GdPO{sub 4} crystalline phase detected with X-ray diffraction. Moreover, Gd{sub 2}O{sub 3} addition increases the T{sub g} of the IBP glasses in glass formation range, which is consistent with the structural modification of the glasses. The structure of the Gd{sub 2}O{sub 3}-loaded IBP glasses/glass-ceramics is mainly based on pyrophosphate units. The chemical durability of Gd{sub 2}O{sub 3}-loaded IBP glasses/glass-ceramics is comparable to widely used borosilicate glass waste forms and the existence of monazite GdPO{sub 4} 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 Pu{sup 3+} if the formed crystalline phase(s) have high chemical durability. - Highlights: • Monazite GdPO{sub 4} are identified in the IBP glasses containing up to 6 mol% Gd{sub 2}O{sub 3}. • R{sub L} of the Gd{sub 2}O{sub 3}-loaded IBP glasses/glass-ceramics are about 10{sup −2} g m{sup −2} d{sup −1}. • Existence of GdPO{sub 4} does not degrade aqueous chemical durability of the IBP glass. • T{sub g} increases with increasing Gd{sub 2}O{sub 3} content in glass formation range. • IBP glasses are potential hosts for the immobilization of Pu{sup 3+} containing HLWs.

  4. Durability Characteristics Analysis of Plastic Worm Wheel with Glass Fiber Reinforced Polyamide.

    Science.gov (United States)

    Kim, Gun-Hee; Lee, Jeong-Won; Seo, Tae-Il

    2013-05-10

    Plastic worm wheel is widely used in the vehicle manufacturing field because it is favorable for weight lightening, vibration and noise reduction, as well as corrosion resistance. However, it is very difficult for general plastics to secure the mechanical properties that are required for vehicle gears. If the plastic resin is reinforced by glass fiber in the fabrication process of plastic worm wheel, it is possible to achieve the mechanical properties of metallic material levels. In this study, the mechanical characteristic analysis of the glass-reinforced plastic worm wheel, according to the contents of glass fiber, is performed by analytic and experimental methods. In the case of the glass fiber-reinforced resin, the orientation and contents of glass fibers can influence the mechanical properties. For the characteristic prediction of plastic worm wheel, computer-aided engineering (CAE) analysis processes such as structural and injection molding analysis were executed with the polyamide resin reinforcement glass fiber (25 wt %, 50 wt %). The injection mold for fabricating the prototype plastic worm wheel was designed and made to reflect the CAE analysis results. Finally, the durability of prototype plastic worm wheel fabricated by the injection molding process was evaluated by the experimental method and the characteristics according to the glass fiber contents.

  5. NUCLEAR WASTE GLASSES: CONTINUOUS MELTING AND BULK VITRIFICAITON

    International Nuclear Information System (INIS)

    KRUGER, A.A.

    2008-01-01

    This contribution addresses various aspects of nuclear waste vitrification. Nuclear wastes have a variety of components and composition ranges. For each waste composition, the glass must be formulated to possess acceptable processing and product behavior defined in terms of physical and chemical properties that guarantee the glass can be easily made and resist environmental degradation. Glass formulation is facilitated by developing property-composition models, and the strategy of model development and application is reviewed. However, the large variability of waste compositions presents numerous additional challenges: insoluble solids and molten salts may segregate; foam may hinder heat transfer and slow down the process; molten salts may accumulate in container refractory walls; the glass on cooling may precipitate crystalline phases. These problems need targeted exploratory research. Examples of specific problems and their possible solutions are discussed

  6. An empirical modeling tool and glass property database in development of US-DOE radioactive waste glasses

    International Nuclear Information System (INIS)

    Muller, I.; Gan, H.

    1997-01-01

    An integrated glass database has been developed at the Vitreous State Laboratory of Catholic University of America. The major objective of this tool was to support glass formulation using the MAWS approach (Minimum Additives Waste Stabilization). An empirical modeling capability, based on the properties of over 1000 glasses in the database, was also developed to help formulate glasses from waste streams under multiple user-imposed constraints. The use of this modeling capability, the performance of resulting models in predicting properties of waste glasses, and the correlation of simple structural theories to glass properties are the subjects of this paper. (authors)

  7. Enhanced HLW glass formulations for the waste treatment and immobilization plant

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A. [DOE-WTP Project Office, US Department of Energy, Richland, Washington (United States)

    2013-07-01

    Current estimates and glass formulation efforts are conservative vis-a-vis achievable waste loadings. These formulations have been specified to ensure that glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP Contract terms. The WTP's overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum, chromium, bismuth, iron, phosphorous, zirconium, and sulfur compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. DOE has a testing program to develop and characterize HLW glasses with higher waste loadings. This work has demonstrated the feasibility of increases in waste loading from 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected these higher waste loading glasses will reduce the HLW canister production requirement by 25% or more. (authors)

  8. An alternative host matrix based on iron phosphate glasses for the vitrification of specialized nuclear waste forms. Annual progress report, September 15, 1996 - September 14, 1997

    International Nuclear Information System (INIS)

    Day, D.E.; Marasinghe, K.; Ray, C.S.

    1997-01-01

    'Objectives of this project are to: (1) investigate the glass composition and processing conditions that yield optimum properties for iron phosphate glasses for vitrifying radioactive waste, (2) determine the atomic structure of iron phosphate glasses and the structure-property relationships, (3) determine how the physical and structural properties of iron phosphate glasses are affected by the addition of simulated high level nuclear waste components, and (4) investigate the process and products of devitrification of iron phosphate waste forms. The glass forming ability of about 125 iron phosphate melts has been investigated in different oxidizing to reducing atmospheres using various iron oxide raw materials such as Fe 2 O 3 , FeO, Fe 3 O 4 , and FeC 2 O 4 2H 2 O. The chemical durability, redox equilibria between Fe(II) and Fe(III), crystallization behavior and structural features for these glasses and their crystalline forms have been investigated using a variety of techniques including Mossbauer spectroscopy, X-ray absorption spectroscopy (XAS), X-ray photoelectron spectroscopy (XPS), Extended x-ray absorption fine structure (EXAFS) and X-ray absorption near edge structure (XANES) analysis, differential thermal and thermogravimetric analysis (DTA/TGA), and X-ray and neutron diffraction.'

  9. Production of highly porous glass-ceramics from metallurgical slag, fly ash and waste glass

    Directory of Open Access Journals (Sweden)

    Mangutova Bianka V.

    2004-01-01

    Full Text Available Glass-ceramics composites were produced based on fly-ash obtained from coal power stations, metallurgical slag from ferronickel industry and waste glass from TV monitors, windows and flasks. Using 50% waste flask glass in combination with fly ash and 20% waste glass from TV screens in combination with slag, E-modulus and bending strength values of the designed systems are increased (system based on fly ash: E-modulus from 6 to 29 GPa, and bending strength from 9 to 75 MPa. The polyurethane foam was used as a pore creator which gave the material porosity of 70(5% (fly ash-glass composite and a porosity of 65( 5% (slag-glass composite. E-modulus values of the designed porous systems were 3.5(1.2 GPa and 8.1(3 GPa, while the bending strength values were 6.0(2 MPa and 13.2(3.5 MPa, respectively. These materials could be used for the production of tiles, wall bricks, as well as for the construction of air diffusers for waste water aeration.

  10. Introduction to the crystallization phenomenon in nuclear glass

    International Nuclear Information System (INIS)

    Jacquet Francillon, N.

    1997-01-01

    Crystallization is a subject for concern because of its potentially detrimental effects on the technological feasibility of high-temperature melting, and on the chemical durability of the material at intermediate and low temperatures during interim storage or after disposal. The tendency of glass to crystallize depends to a large extent on the composition of the frit and/or of the waste to be solidified. It depends too of the thermal history of the glass generally, the knowledge is mainly upon determination of the time-temperature-transition (TTT) curves, crystal identification and quantification techniques, and their effects on the durability of the glass matrix. French experience is presented. Only a few authors addressed the long-term development of crystalline phases, notably at temperatures below the vitreous transition temperature Tg. Some recommendations for glass crystallization studies are made but glass crystallization after disposal is acceptable provided some conditions are met. (author)

  11. Thermal Predictions of the Cooling of Waste Glass Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2014-11-01

    Radioactive liquid waste from five decades of weapons production is slated for vitrification at the Hanford site. The waste will be mixed with glass forming additives and heated to a high temperature, then poured into canisters within a pour cave where the glass will cool and solidify into a stable waste form for disposal. Computer simulations were performed to predict the heat rejected from the canisters and the temperatures within the glass during cooling. Four different waste glass compositions with different thermophysical properties were evaluated. Canister centerline temperatures and the total amount of heat transfer from the canisters to the surrounding air are reported.

  12. WTP Waste Feed Qualification: Glass Fabrication Unit Operation Testing Report

    Energy Technology Data Exchange (ETDEWEB)

    Stone, M. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Hanford Missions Programs; Newell, J. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Process Technology Programs; Johnson, F. C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Engineering Process Development; Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL). Engineering Process Development

    2016-07-14

    The waste feed qualification program is being developed to protect the Hanford Tank Waste Treatment and Immobilization Plant (WTP) design, safety basis, and technical basis by assuring waste acceptance requirements are met for each staged waste feed campaign prior to transfer from the Tank Operations Contractor to the feed receipt vessels inside the Pretreatment Facility. The Waste Feed Qualification Program Plan describes the three components of waste feed qualification: 1. Demonstrate compliance with the waste acceptance criteria 2. Determine waste processability 3. Test unit operations at laboratory scale. The glass fabrication unit operation is the final step in the process demonstration portion of the waste feed qualification process. This unit operation generally consists of combining each of the waste feed streams (high-level waste (HLW) and low-activity waste (LAW)) with Glass Forming Chemicals (GFCs), fabricating glass coupons, performing chemical composition analysis before and after glass fabrication, measuring hydrogen generation rate either before or after glass former addition, measuring rheological properties before and after glass former addition, and visual observation of the resulting glass coupons. Critical aspects of this unit operation are mixing and sampling of the waste and melter feeds to ensure representative samples are obtained as well as ensuring the fabrication process for the glass coupon is adequate. Testing was performed using a range of simulants (LAW and HLW simulants), and these simulants were mixed with high and low bounding amounts of GFCs to evaluate the mixing, sampling, and glass preparation steps in shielded cells using laboratory techniques. The tests were performed with off-the-shelf equipment at the Savannah River National Laboratory (SRNL) that is similar to equipment used in the SRNL work during qualification of waste feed for the Defense Waste Processing Facility (DWPF) and other waste treatment facilities at the

  13. Impact of Alkali Source on Vitrification of SRS High Level Waste

    International Nuclear Information System (INIS)

    LAMBERT, D. P.; MILLER, D. H.; PEELER, D. K.; SMITH, M. E.; STONE, M. E.

    2005-01-01

    The Defense Waste Processing Facility (DWPF) Savannah River Site is currently immobilizing high level nuclear waste sludge by vitrification in borosilicate glass. The processing strategy involves blending a large batch of sludge into a feed tank, washing the sludge to reduce the amount of soluble species, then processing the large ''sludge batch'' through the DWPF. Each sludge batch is tested by the Savannah River National Laboratory (SRNL) using simulants and tests with samples of the radioactive waste to ''qualify'' the batch prior to processing in the DWPF. The DWPF pretreats the sludge by first acidifying the sludge with nitric and formic acid. The ratio of nitric to formic acid is adjusted as required to target a final glass composition that is slightly reducing (the target is for ∼20% of the iron to have a valence of two in the glass). The formic acid reduces the mercury in the feed to elemental mercury which is steam stripped from the feed. After a concentration step, the glass former (glass frit) is added as a 50 wt% slurry and the batch is concentrated to approximately 50 wt% solids. The feed slurry is then fed to a joule heated melter maintained at 1150 C. The glass must meet both processing (e.g., viscosity and liquidus temperature) and product performance (e.g., durability) constraints The alkali content of the final waste glass is a critical parameter that affects key glass properties (such as durability) as well as the processing characteristics of the waste sludge during the pretreatment and vitrification processes. Increasing the alkali content of the glass has been shown to improve the production rate of the DWPF, but the total alkali in the final glass is limited by constraints on glass durability and viscosity. Two sources of alkali contribute to the final alkali content of the glass: sodium salts in the waste supernate and sodium and lithium oxides in the glass frit added during pretreatment processes. Sodium salts in the waste supernate can

  14. Thermochemical modeling of nuclear waste glass

    International Nuclear Information System (INIS)

    Spear, K.E.; Besmann, T.M.; Beahm, E.C.

    1998-06-01

    The development of assessed and consistent phase equilibria and thermodynamic data for major glass constituents used to incorporate high-level nuclear waste is discussed in this paper. The initial research has included the binary Na 2 O-SiO 2 , Na 2 O-Al 2 O 3 , and SiO 2 -Al 2 O 3 systems. The nuclear waste glass is assumed to be a supercooled liquid containing the constituents in the glass at temperatures of interest for nuclear waste storage. Thermodynamic data for the liquid solutions were derived from mathematical comparisons of phase diagram information and the thermodynamic data available for crystalline solid phases. An associate model is used to describe the liquid solution phases. Utilizing phase diagram information provides very stringent limits on the relative thermodynamic stabilities of all phases which exist in a given system

  15. Raman and X-ray absorption spectroscopic studies of hydrothermally altered alkali-borosilicate nuclear waste glass

    Energy Technology Data Exchange (ETDEWEB)

    McKeown, David A., E-mail: davidm@vsl.cua.ed [Vitreous State Laboratory, Catholic University of America, 620 Michigan Ave., N.E., Washington, DC 20064 (United States); Buechele, Andrew C.; Viragh, Carol; Pegg, Ian L. [Vitreous State Laboratory, Catholic University of America, 620 Michigan Ave., N.E., Washington, DC 20064 (United States)

    2010-04-01

    Raman spectroscopy and X-ray absorption spectroscopy (XAS) are used to characterize structural changes that took place in hydrothermally altered (Na,K)-alumina-borosilicate glasses with different Na/K ratios, formulated as part of a durability study to investigate the behavior of glasses for nuclear waste storage. The hydrothermal experiments, or vapor hydration tests (VHT), were performed on each glass for 3 and 20 days at 200 deg. C to accelerate and approximate long-term alteration processes that may occur in a nuclear waste repository. Results found for both glasses and their VHT altered counterparts show little, if any, structural influence from the different starting Na/K ratios. X-ray diffraction, differential scanning calorimetry, scanning electron microscopy, and Raman spectroscopy indicate that the altered samples are mostly amorphous with small amounts of analcime-like and leucite-like crystals within 200 mum of the sample surface and contain up to 9.7 wt.% water or OH. The Raman data are nearly identical for the amorphous portions of all altered VHT samples investigated, and indicate that two glass structural changes took place during alteration: one, partial depolymerization of the alumina-borosilicate network, and two, introduction of water or OH. Al and Si XAS data indicate tetrahedral AlO{sub 4} and SiO{sub 4} environments in the original glasses as well as in the altered samples. Small energy shifts of the Si K-edge also show that the altered VHT samples have less polymerized networks than the original glass. Na XAS data indicate expanded Na environments in the VHT samples with longer Na-O distances and more nearest-neighbor oxygen atoms, compared with the original glasses, which may be due to hydrous species introduced into the expanding Na-sites.

  16. Durability and degradation of HT9 based alloy waste forms with variable Ni and Cr content

    Energy Technology Data Exchange (ETDEWEB)

    Olson, L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-12-31

    Short-term electrochemical and long-term hybrid electrochemical corrosion tests were performed on alloy waste forms in reference aqueous solutions that bound postulated repository conditions. The alloy waste forms investigated represent candidate formulations that can be produced with advanced electrochemical treatment of used nuclear fuel. The studies helped to better understand the alloy waste form durability with differing concentrations of nickel and chromium, species that can be added to alloy waste forms to potentially increase their durability and decrease radionuclide release into the environment.

  17. Results of testing the Grambow rate law for use in HWVP glass durability correlations

    International Nuclear Information System (INIS)

    Kuhn, W.L.; Bunnell, L.R.

    1996-03-01

    A theory based on Grambow's work on hydration of glass as linear function of solution composition was evaluated. Use of Grambow's linear rate law for correlation of durability with glass composition is not recommended. Dissolution rate of the glass was determined using the rate of release of sodium with an ion selective electrode. This method was tested first applying it to initial dissolution rate of several glasses at several temperatures with zero initial concentration of silicic acid. HW39-2, HW39-4, and SRL-202 from Savannah River were tested; there was significant scatter in the data, with the dissolution rates of HW39 glasses and the SRL glass being comparable within this scatter. The dissolution rate of SRL-202 at 80 C and pH 7 for silicic acid concentrations 0, 25, 50, and 100% saturation, was found to decrease dramatically at only 25% of the saturated silicic acid concentration, which does not conform to the linear theory

  18. Glassceramics obtained from industrial waste

    Energy Technology Data Exchange (ETDEWEB)

    Cimdins, R.; Rozenstrauha, I.; Berzina, L. [Riga Technical University, Faculty of Chemical Technology, Biomaterials R and D Laboratory, 14/24 Azenes St., LV-1048 Riga (Latvia); Bossert, J.; Buecker, M. [Technisches Institut Materialwissenschaft, Friedrich-Schiller Universitaet, Loebdegraben 32, 07743 Jena (Germany)

    2000-06-01

    Large areas of Latvia are contaminated with industrial waste: metallurgical slag, fly-ash, etching refuse, peat, and coal ash as well as glass waste which often contain dangerous substances. From the environmental point of view this waste should be neutralised. As this waste also contains valuable chemical compounds, it can be considered as a raw material for the generation of new materials. One method of utilisation is to produce recycled materials - street plates, decorative tiles, or floor tiles. Dense sintered glassceramics with a water uptake of 0.34-3.23 wt.%, a final density of 2.93-3.05 g/cm{sup 3}, and a bending strength of 80-96 MPa have been created from industrial waste. The mast chemically durable glassceramics contained clay additions. Thus, the material containing only waste had a durability (mass loss) of 3.02% in 0.1 N HCl, while the composition containing 30% clay addition had a durability of 0.2% in 0.1 N HCl.

  19. Processing of high-temperature simulated waste glass in a continuous ceramic melter

    International Nuclear Information System (INIS)

    Barnes, S.M.; Brouns, R.A.; Hanson, M.S.

    1980-01-01

    Recent operations have demonstrated that high-melting-point glasses and glass-ceramics can be successfully processed in joule-heated, ceramic-lined melters with minor modifications to the existing technology. Over 500 kg of simulated waste glasses have been processed at temperatures up to 1410 0 C. The processability of the two high-temperature waste forms tested is similar to existing borosilicate waste glasses. High-temperature waste glass formulations produced in the bench-scale melter exhibit quality comparing favorably to standard waste glass formulations

  20. Preparation of plutonium waste forms with ICPP calcined high-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Staples, B.A.; Knecht, D.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); O`Holleran, T.P. [Argonne National Lab.-West, Idaho Falls, ID (United States)] [and others

    1997-05-01

    Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce{sup +4}) as a surrogate for plutonium (Pu{sup +4}) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study.

  1. Preparation of plutonium waste forms with ICPP calcined high-level waste

    International Nuclear Information System (INIS)

    Staples, B.A.; Knecht, D.A.; O'Holleran, T.P.

    1997-05-01

    Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce +4 ) as a surrogate for plutonium (Pu +4 ) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study

  2. The Radiation Effect to Waste Glass that Resulting of Vitrification

    International Nuclear Information System (INIS)

    Herlan Martono; Aisyah

    2002-01-01

    The high level liquid waste (HLLW) is generated from the first step extraction of the nuclear fuel reprocessing. This waste was contain of few of actinide and many of fission product. The alpha radiation of actinide that contain on the HLLW cause the change the waste glass characteristic. The experiment was conducted by the doping, irradiation and heating of waste glass resulting from vitrification. The alpha radiation cause the change of composition that could be detected from change of waste glass density and mechanical strength. The increasing of alpha radiation dose cause the increasing change of density and mechanical strength, although the change of mechanical strength is not significant. Degree of change of waste glass density also depend on type of waste-glass and reach for saturated point at over of 5x10 24 alpha decay/m 3 . The gamma radiation of fission product that contain on the HLLW can increasing of waste glass temperature that cause the structure change, so devitrification was occur. The devitrification can the increasing of leaching rate. The cumulative of gamma dose rate was not cause the devitrification. (author)

  3. Glass Ceramic Formulation Data Package

    International Nuclear Information System (INIS)

    Crum, Jarrod V.; Rodriguez, Carmen P.; McCloy, John S.; Vienna, John D.; Chung, Chul-Woo

    2012-01-01

    A glass ceramic waste form is being developed for treatment of secondary waste streams generated by aqueous reprocessing of commercial used nuclear fuel (Crum et al. 2012b). The waste stream contains a mixture of transition metals, alkali, alkaline earths, and lanthanides, several of which exceed the solubility limits of a single phase borosilicate glass (Crum et al. 2009; Caurant et al. 2007). A multi-phase glass ceramic waste form allows incorporation of insoluble components of the waste by designed crystallization into durable heat tolerant phases. The glass ceramic formulation and processing targets the formation of the following three stable crystalline phases: (1) powellite (XMoO4) where X can be (Ca, Sr, Ba, and/or Ln), (2) oxyapatite Yx,Z(10-x)Si6O26 where Y is alkaline earth, Z is Ln, and (3) lanthanide borosilicate (Ln5BSi2O13). These three phases incorporate the waste components that are above the solubility limit of a single-phase borosilicate glass. The glass ceramic is designed to be a single phase melt, just like a borosilicate glass, and then crystallize upon slow cooling to form the targeted phases. The slow cooling schedule is based on the centerline cooling profile of a 2 foot diameter canister such as the Hanford High-Level Waste canister. Up to this point, crucible testing has been used for glass ceramic development, with cold crucible induction melter (CCIM) targeted as the ultimate processing technology for the waste form. Idaho National Laboratory (INL) will conduct a scaled CCIM test in FY2012 with a glass ceramic to demonstrate the processing behavior. This Data Package documents the laboratory studies of the glass ceramic composition to support the CCIM test. Pacific Northwest National Laboratory (PNNL) measured melt viscosity, electrical conductivity, and crystallization behavior upon cooling to identify a processing window (temperature range) for melter operation and cooling profiles necessary to crystallize the targeted phases in the

  4. Characterization of borosilicate glasses containing simulated high-level radioactive wastes from PNC

    International Nuclear Information System (INIS)

    Terai, R.; Eguchi, K.; Yamanaka, H.

    1979-01-01

    The characterization of borosilicate glasses containing simulated HLW from PNC has been carried out. Phase separation of molybdates, volatilization, viscosity, electrical resistivity, thermal conductivity, elastic modulus, chemical durability, and devitrification of these glasses have been measured, and the suitability of the glasses for the vitrified solidification processes is discussed from the viewpoint of safety

  5. Glass formulation for phase 1 high-level waste vitrification

    International Nuclear Information System (INIS)

    Vienna, J.D.; Hrma, P.R.

    1996-04-01

    The purpose of this study is to provide potential glass formulations for prospective Phase 1 High-Level Waste (HLW) vitrification at Hanford. The results reported here will be used to aid in developing a Phase 1 HLW vitrification request for proposal (RFP) and facilitate the evaluation of ensuing proposals. The following factors were considered in the glass formulation effort: impact on total glass volume of requiring the vendor to process each of the tank compositions independently versus as a blend; effects of imposing typical values of B 2 O 3 content and waste loading in HLW borosilicate glasses as restrictions on the vendors (according to WAPS 1995, the typical values are 5--10 wt% B 2 O 3 and 20--40 wt% waste oxide loading); impacts of restricting the processing temperature to 1,150 C on eventual glass volume; and effects of caustic washing on any of the selected tank wastes relative to glass volume

  6. Glass formulation for phase 1 high-level waste vitrification

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, J.D.; Hrma, P.R.

    1996-04-01

    The purpose of this study is to provide potential glass formulations for prospective Phase 1 High-Level Waste (HLW) vitrification at Hanford. The results reported here will be used to aid in developing a Phase 1 HLW vitrification request for proposal (RFP) and facilitate the evaluation of ensuing proposals. The following factors were considered in the glass formulation effort: impact on total glass volume of requiring the vendor to process each of the tank compositions independently versus as a blend; effects of imposing typical values of B{sub 2}O{sub 3} content and waste loading in HLW borosilicate glasses as restrictions on the vendors (according to WAPS 1995, the typical values are 5--10 wt% B{sub 2}O{sub 3} and 20--40 wt% waste oxide loading); impacts of restricting the processing temperature to 1,150 C on eventual glass volume; and effects of caustic washing on any of the selected tank wastes relative to glass volume.

  7. ANL technical support program for DOE Environmental Restoration and Waste Management

    International Nuclear Information System (INIS)

    Bates, J.K.; Bradley, C.R.; Buck, E.C.; Cunnane, J.C.; Dietz, N.L.; Ebert, W.L.; Emery, J.W.; Feng, X.; Gerding, T.J.; Gong, M.; Hoh, J.C.; Mazer, J.J.; Wronkiewicz, D.J.; Bourcier, W.L.; Morgan, L.E.; Newton, L.; Nielsen, J.K.; Phillips, B.L.; Ewing, R.C.; Wang, L.; Li, H.; Tomozawa, M.

    1993-05-01

    A program was established for DOE Environmental Restoration and Waste Management (EM) to evaluate factors that are anticipated to affect waste glass reaction during repository disposal, especially in an unsaturated environment typical of what may be expected for the proposed Yucca Mountain repository site. This report covers progress in FY 1992 on the following tasks: 1. A compendium of the characteristics of high-level nuclear waste borosilicate glass has been written. 2. A critical review of important parameters that affect the reactivity of glass in an unsaturated environment is being prepared. 3. A series of tests has been started to evaluate the reactivity of fully radioactive glasses in a high-level waste repository environment and compare it to the reactivity of synthetic, nonradioactive glasses of similar composition. 4. The effect of radiation upon the durability of waste glasses at a high glass surface area-to-liquid volume (SA/V) ratio and a high gas-to-liquid volume ratio will be assessed. These tests address both vapor and high SA/V liquid conditions. 5. A series of tests is being performed to compare the extent of reaction of nuclear waste glasses at various SAN ratios. Such differences in the SAN ratio may significantly affect glass durability. 6. A series of natural analogue tests is being analyzed to demonstrate a meaningful relationship between experimental and natural alteration conditions. 7. Analytical electron microscopy (AEM), infrared spectroscopys and nuclear resonant profiling are being used to assess the glass/water reaction pathway by identifying intermediate phases that appear on the reacting glass. Additionally, colloids from the leach solutions are being studied using AEM. 8. A technical review of AEM results is being provided. 9. A study of water diffusion involving nuclear waste glasses is being performed. 10. A mechanistically based model is being developed to predict the performance of glass over repository-relevant time periods

  8. Characterization study of industrial waste glass as starting material ...

    African Journals Online (AJOL)

    In present study, an industrial waste glass was characterized and the potential to assess as starting material in development of bioactive materials was investigated. A waste glass collected from the two different glass industry was grounded to fine powder. The samples were characterized using X-ray fluorescence (XRF), ...

  9. Borosilicate glass as a matrix for the immobilization of Savannah River Plant waste

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Wicks, G.G.; Bibler, N.E.

    1982-01-01

    The reference waste form for immobilization of Savannah River Plant (SRP) waste is borosilicate glass. In the reference process, waste is mixed with glass-forming chemicals and melted in a Joule-heated ceramic melter at 1150 0 C. Waste glass made with actual or simulated waste on a small scale and glass made with simulated waste on a large scale confirm that the current reference process and glass-former composition are able to accommodate all SRP waste compositions and can produce a glass with: high waste loading; low leach rates; good thermal stability; high resistance to radiation effects; and good impact resistance. Borosilicate glass has been studied as a matrix for the immobilization of SRP waste since 1974. This paper reviews the results of extensive characterization and performance testing of the glass product. These results show that borosilicate glass is a very suitable matrix for the immobilization of SRP waste. 18 references, 3 figures, 10 tables

  10. Vitrification of hazardous and radioactive wastes

    International Nuclear Information System (INIS)

    Bickford, D.F.; Schumacher, R.

    1995-01-01

    Vitrification offers many attractive waste stabilization options. Versatility of waste compositions, as well as the inherent durability of a glass waste form, have made vitrification the treatment of choice for high-level radioactive wastes. Adapting the technology to other hazardous and radioactive waste streams will provide an environmentally acceptable solution to many of the waste challenges that face the public today. This document reviews various types and technologies involved in vitrification

  11. Nuclear waste disposal: alternatives to solidification in glass proposed

    International Nuclear Information System (INIS)

    Kerr, R.A.

    1979-01-01

    More than a quarter-million cubic meters of liquid radioactive wastes are now being held at government installations awaiting final disposal. During the past 20 years, the disposal plan of choice has been to incorporate the 40 to 50 radioactive elements dissolved in liquid wastes into blocks of glass, seal the glass in metal canisters, and insert the canisters into deep, geologically stable salt beds. Over the last few years, some geologists and materials scientists have become concerned that perhaps not enough is known yet about the interaction of waste, container, and salt (or any rock) to have a reasonable assurance that the hazardous wastes will be contained successfully. The biggest advantage of glass at present is the demonstrated practicality of producing large, highly radioactive blocks of it. The frontrunner as a successor to glass is ceramics, which are nonmetallic crystalline materials formed at high temperature, such as chinaware or natural minerals. An apparent advantage of ceramics is that they already have an ordered atomic structure, whose properties can be tailored to a particular waste element and to conditions of a specific disposal site. A ceramic tailored for waste disposal called supercalcine-ceramic has been developed. It was emphasized that the best minerals for waste solidification may be those that have proved most stable under natural conditions over geologic time. Disadvantage to ceramics are radiation damage and transmutation. However, it is now obvious that some ceramics are more stable than glass under certain conditions. Metal-encapsulated ceramic, called cermet, is being developed as a waste form. Cermets are considerably more resistant at 100 0 C than a borosilicate waste glass. Researchers are now testing prospective waste forms under the most extreme conditions that might prevail in a waste disposal site

  12. Fixation of radioactive waste in glass

    International Nuclear Information System (INIS)

    Chapman, C.C.; Mendel, J.E.

    1976-08-01

    After a brief review of the source of high level wastes and the specific requirements and desirable characteristics of glass used as a storage vehicle, the development work done on two vitrification systems is outlined. One is an in-can melter system and the second is a ceramic melter. Primary emphasis has been placed on the in-can melter system for use in the near future. Both systems are capable of converting high level waste to a glass which possesses low release potential

  13. Formulation and Characterization of Waste Glasses with Varying Processing Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong-Sang; Schweiger, M. J.; Rodriguez, Carmen P.; Lepry, William C.; Lang, Jesse B.; Crum, Jarrod V.; Vienna, John D.; Johnson, Fabienne; Marra, James C.; Peeler, David K.

    2011-10-17

    This report documents the preliminary results of glass formulation and characterization accomplished within the finished scope of the EM-31 technology development tasks for WP-4 and WP-5, including WP-4.1.2: Glass Formulation for Next Generation Melter, WP-5.1.2.3: Systematic Glass Studies, and WP-5.1.2.4: Glass Formulation for Specific Wastes. This report also presents the suggested studies for eventual restart of these tasks. The initial glass formulation efforts for the cold crucible induction melter (CCIM), operating at {approx}1200 C, with selected HLW (AZ-101) and LAW (AN-105) successfully developed glasses with significant increase of waste loading compared to that is likely to be achieved based on expected reference WTP formulations. Three glasses formulated for AZ-101HLW and one glass for AN-105 LAW were selected for the initial CCIM demonstration melter tests. Melter tests were not performed within the finished scope of the WP-4.1.2 task. Glass formulations for CCIM were expanded to cover additional HLWs that have high potential to successfully demonstrate the unique advantages of the CCIM technologies based on projected composition of Hanford wastes. However, only the preliminary scoping tests were completed with selected wastes within the finished scope. Advanced glass formulations for the reference WTP melter, operating at {approx}1200 C, were initiated with selected specific wastes to determine the estimated maximum waste loading. The incomplete results from these initial formulation efforts are summarized. For systematic glass studies, a test matrix of 32 high-aluminum glasses was completed based on a new method developed in this study.

  14. Formulation and Characterization of Waste Glasses with Varying Processing Temperature

    International Nuclear Information System (INIS)

    Kim, Dong-Sang; Schweiger, M.J.; Rodriguez, Carmen P.; Lepry, William C.; Lang, Jesse B.; Crum, Jarrod V.; Vienna, John D.; Johnson, Fabienne; Marra, James C.; Peeler, David K.

    2011-01-01

    This report documents the preliminary results of glass formulation and characterization accomplished within the finished scope of the EM-31 technology development tasks for WP-4 and WP-5, including WP-4.1.2: Glass Formulation for Next Generation Melter, WP-5.1.2.3: Systematic Glass Studies, and WP-5.1.2.4: Glass Formulation for Specific Wastes. This report also presents the suggested studies for eventual restart of these tasks. The initial glass formulation efforts for the cold crucible induction melter (CCIM), operating at ∼1200 C, with selected HLW (AZ-101) and LAW (AN-105) successfully developed glasses with significant increase of waste loading compared to that is likely to be achieved based on expected reference WTP formulations. Three glasses formulated for AZ-101HLW and one glass for AN-105 LAW were selected for the initial CCIM demonstration melter tests. Melter tests were not performed within the finished scope of the WP-4.1.2 task. Glass formulations for CCIM were expanded to cover additional HLWs that have high potential to successfully demonstrate the unique advantages of the CCIM technologies based on projected composition of Hanford wastes. However, only the preliminary scoping tests were completed with selected wastes within the finished scope. Advanced glass formulations for the reference WTP melter, operating at ∼1200 C, were initiated with selected specific wastes to determine the estimated maximum waste loading. The incomplete results from these initial formulation efforts are summarized. For systematic glass studies, a test matrix of 32 high-aluminum glasses was completed based on a new method developed in this study.

  15. Hydration of high-silica glasses in the deep sea

    International Nuclear Information System (INIS)

    Federman, A.N.

    1986-01-01

    Natural analogs of nuclear waste glasses are important because they provide information of the one variable that is not controllable in the laboratory - long intervals of time in the actual environment of storage. Some natural glasses have persisted for millions of years in deep-sea sediments in the form of disseminated particles and distinct tephra layers, while other apparently similar specimens have been completely altered to clay assemblages relatively quickly. Geologists have reached no firm conclusions as to why these differences exist, and more research is certainly warranted. These glasses vary in age, composition, and in the in-situ conditions they have experienced. They may provide important information for two different aspects of nuclear waste glass research: First, the chemical composition and especially the water content of these glasses as a function of time may give an understanding of the mechanisms and rates of diffusion in glasses in the natural environment. Second, the apparent differing durability of these glasses in different environmental conditions may suggest the optimal characteristics of a nuclear waste glass depository

  16. Disposal costs for SRP high-level wastes in borosilicate glass and crystalline ceramic waste forms

    International Nuclear Information System (INIS)

    Rozsa, R.B.; Campbell, J.H.

    1982-01-01

    Purpose of this document is to compare and contrast the overall burial costs of the glass and ceramic waste forms, including processing, storage, transportation, packaging, and emplacement in a repository. Amount of waste will require approximately 10,300 standard (24 in. i.d. x 9-5/6 ft length) canisters of waste glass, each containing about 3260 lb of waste at 28% waste loading. The ceramic waste form requires about one-third the above number of standard canisters. Approximately $2.5 billion is required to process and dispose of this waste, and the total cost is independent of waste form (glass or ceramic). The major cost items (about 80% of the total cost) for all cases are capital and operating expenses. The capital and 20-year operating costs for the processing facility are the same order of magnitude, and their sum ranges from about one-half of the total for the reference glass case to two-thirds of the total for the ceramic cases

  17. Baseline Glass Development for Combined Fission Products Waste Streams

    International Nuclear Information System (INIS)

    Crum, Jarrod V.; Billings, Amanda Y.; Lang, Jesse B.; Marra, James C.; Rodriguez, Carmen P.; Ryan, Joseph V.; Vienna, John D.

    2009-01-01

    Borosilicate glass was selected as the baseline technology for immobilization of the Cs/Sr/Ba/Rb (Cs), lanthanide (Ln) and transition metal fission product (TM) waste steams as part of a cost benefit analysis study.(1) Vitrification of the combined waste streams have several advantages, minimization of the number of waste forms, a proven technology, and similarity to waste forms currently accepted for repository disposal. A joint study was undertaken by Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL) to develop acceptable glasses for the combined Cs + Ln + TM waste streams (Option 1) and Cs + Ln combined waste streams (Option 2) generated by the AFCI UREX+ set of processes. This study is aimed to develop baseline glasses for both combined waste stream options and identify key waste components and their impact on waste loading. The elemental compositions of the four-corners study were used along with the available separations data to determine the effect of burnup, decay, and separations variability on estimated waste stream compositions.(2-5) Two different components/scenarios were identified that could limit waste loading of the combined Cs + LN + TM waste streams, where as the combined Cs + LN waste stream has no single component that is perceived to limit waste loading. Combined Cs + LN waste stream in a glass waste form will most likely be limited by heat due to the high activity of Cs and Sr isotopes.

  18. Utilization of waste glass in ECO-cement: Strength properties and microstructural observations

    International Nuclear Information System (INIS)

    Sobolev, Konstantin; Tuerker, Pelin; Soboleva, Svetlana; Iscioglu, Gunsel

    2007-01-01

    Waste glass creates a serious environmental problem, mainly because of the inconsistency of the waste glass streams. The use of waste glass as a finely ground mineral additive (FGMA) in cement is a promising direction for recycling. Based on the method of mechano-chemical activation, a new group of ECO-cements was developed. In ECO-cement, relatively large amounts (up to 70%) of portland cement clinker can be replaced with waste glass. This report examines the effect of waste glass on the microstructure and strength of ECO-cement based materials. Scanning electron microscopy (SEM) investigations were used to observe the changes in the cement hydrates and interface between the cement matrix and waste glass particles. According to the research results, the developed ECO-cement with 50% of waste glass possessed compressive strength properties at a level similar to normal portland cement

  19. Glass-ceramics with multibarrier structure obtained from industrial waste

    Energy Technology Data Exchange (ETDEWEB)

    Berzina, L.; Cimdins, R.; Rozenstrauha, I. [Riga Tech. Univ. (Latvia). Fac. of Chem. Technol.; Bossert, J. [Technisches Inst.: Materialwissenschaft, Friedrich-Schiller-Univ., Jena (Germany); Kravtchenko, I. [Inst. for Problems of Material Science, Kiev (Ukraine)

    1997-12-31

    Recycling problem for various kind of waste is solved by processing the waste to ecological depositable products with multibarrier structure. In order to form a multibarrier structure the ecologically incompatible substances may be diluted and chemically bound until their recycling products gain a structure like natural mineral or glass (I. barrier). After that, remineralized materials are converted into a new product by melting or powder technology using an ecological compatible type of waste as a matrix phase (II. barrier). Waste which are treated this way could be applied to produce ceramic building materials and goods such as floor tiles, stone pavement and casting products. Industrial waste from the metallurgical factory in Latvia ``Liepajas metalurgs`` are metallurgical slag, filter dust, etching waste and sewage used in technologies. The main constituents of chemical compositions of these waste are: Fe, Ca, Si, Mg, Al, Mn etc. In some types of waste a small amount of ecologically risky elements such as Cr, Ni, Zr, Sn and Pb can occur. The combination of metallurgical waste with peat ashes from Riga thermal power station, oil shale ashes or glass waste under controlled sintering procedure gives bulk materials with surface or/and bulkcrystallization. The structure of glass-ceramics built this way may prevent the migration of ecologically risky elements into environment due to corrosion or friction. Physical-chemical properties and thermal behaviour (DTA, dilatometry, melting) of waste define the range of sintering for production of glass-ceramics (powder technology) and decorative glass-ceramic materials (melting and powder technology). (orig.) 5 refs.

  20. Defense Waste Processing Facility (DWPF) Viscosity Model: Revisions for Processing High TiO2 Containing Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-30

    Radioactive high-level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the Defense Waste Processing Facility (DWPF) since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it is poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to guarantee, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository. The DWPF SPC system is known as the Product Composition Control System (PCCS). The DWPF will soon be receiving wastes from the Salt Waste Processing Facility (SWPF) containing increased concentrations of TiO2, Na2O, and Cs2O . The SWPF is being built to pretreat the high-curie fraction of the salt waste to be removed from the HLW tanks in the F- and H-Area Tank Farms at the SRS. In order to process TiO2 concentrations >2.0 wt% in the DWPF, new viscosity data were developed over the range of 1.90 to 6.09 wt% TiO2 and evaluated against the 2005 viscosity model. An alternate viscosity model is also derived for potential future use, should the DWPF ever need to process other titanate-containing ion exchange materials. The ultimate limit on the amount of TiO2 that can be accommodated from SWPF will be determined by the three PCCS models, the waste composition of a given sludge

  1. Waste E-glass particles used in cementitious mixtures

    International Nuclear Information System (INIS)

    Chen, C.H.; Huang, R.; Wu, J.K.; Yang, C.C.

    2006-01-01

    The properties of concretes containing various waste E-glass particle contents were investigated in this study. Waste E-glass particles were obtained from electronic grade glass yarn scrap by grinding to small particle size. The size distribution of cylindrical glass particle was from 38 to 300 μm and about 40% of E-glass particle was less than 150 μm. The E-glass mainly consists of SiO 2 , Al 2 O 3 , Ca O and MgO, and is indicated as amorphous by X-ray diffraction (XRD) technique. Compressive strength and resistance of sulfate attack and chloride ion penetration were significantly improved by utilizing proper amount of waste E-glass in concrete. The compressive strength of specimen with 40 wt.% E-glass content was 17%, 27% and 43% higher than that of control specimen at age of 28, 91 and 365 days, respectively. E-glass can be used in concrete as cementitious material as well as inert filler, which depending upon the particle size, and the dividing size appears to be 75 μm. The workability decreased as the glass content increased due to reduction of fineness modulus, and the addition of high-range water reducers was needed to obtain a uniform mix. Little difference was observed in ASR testing results between control and E-glass specimens. Based on the properties of hardened concrete, optimum E-glass content was found to be 40-50 wt.%

  2. Hot isostatically-pressed aluminosilicate glass-ceramic with natural crystalline analogues for immobilizing the calcined high-level nuclear waste at the Idaho Chemical Processing Plant

    International Nuclear Information System (INIS)

    Raman, S.

    1993-12-01

    The additives Si, Al, MgO, P 2 O 5 were mechanically blended with fluorinelsodium calcine in varying proportions. The batches were vacuum sealed in stainless steel canisters and hot isostatically pressed at 20,000 PSI and 1000 C for 4 hours. The resulting suite of glass-ceramic waste forms parallels the natural rocks in microstructural and compositional heterogeneity. Several crystalline phases ar analogous in composition and structure to naturally occurring minerals. Additional crystalline phases are zirconia and Ca-Mg borate. The glasses are enriched in silica and alumina. Approximately 7% calcine elements occur dissolved in this glass and the total glass content in the waste forms averages 20 wt%. The remainder of the calcine elements are partitioned into crystalline phases at 75 wt% calcine waste loading. The waste forms were tested for chemical durability in accordance with the MCC1-test procedure. The leach rates are a function of the relative proportions of additives and calcine, which in turn influence the composition and abundances of the glass and crystalline phases. The DOE leach rate criterion of less than 1 g/m 2 -day is met by all the elements B, Cs and Na are increased by lowering the melt viscosity. This is related to increased crystallization or devitrification with increases in MgO addition. This exploratory work has shown that the increases in waste loading occur by preferred partitioning of the calcine components among crystalline and glass phases. The determination of optimum processing parameters in the form of additive concentration levels, homogeneous blending among the components, and pressure-temperature stabilities of phases must be continued to eliminate undesirable effects of chemical composition, microstructure and glass devitrification

  3. Characteristics of waste automotive glasses as silica resource in ferrosilicon synthesis.

    Science.gov (United States)

    Farzana, Rifat; Rajarao, Ravindra; Sahajwalla, Veena

    2016-02-01

    This fundamental research on end-of-life automotive glasses, which are difficult to recycle, is aimed at understanding the chemical and physical characteristics of waste glasses as a resource of silica to produce ferrosilicon. Laboratory experiments at 1550°C were carried out using different automotive glasses and the results compared with those obtained with pure silica. In situ images of slag-metal separation showed similar behaviour for waste glasses and silica-bearing pellets. Though X-ray diffraction (XRD) showed different slag compositions for glass and silica-bearing pellets, formation of ferrosilicon was confirmed. Synthesized ferrosilicon alloy from waste glasses and silica were compared by Raman, X-ray photoelectron spectroscopy and scanning electron microscopy (SEM) analysis. Silicon concentration in the synthesized alloys showed almost 92% silicon recovery from the silica-bearing pellet and 74-92% silicon recoveries from various waste glass pellets. The polyvinyl butyral (PVB) plastic layer in the windshield glass decomposed at low temperature and did not show any detrimental effect on ferrosilicon synthesis. This innovative approach of using waste automotive glasses as a silica source for ferrosilicon production has the potential to create sustainable pathways, which will reduce specialty glass waste in landfill. © The Author(s) 2015.

  4. Glass optimization for vitrification of Hanford Site low-level tank waste

    International Nuclear Information System (INIS)

    Feng, X.; Hrma, P.R.; Westsik, J.H. Jr.

    1996-03-01

    The radioactive defense wastes stored in 177 underground single-shell tanks (SST) and double-shell tanks (DST) at the Hanford Site will be separated into low-level and high-level fractions. One technology activity underway at PNNL is the development of glass formulations for the immobilization of the low-level tank wastes. A glass formulation strategy has been developed that describes development approaches to optimize glass compositions prior to the projected LLW vitrification facility start-up in 2005. Implementation of this strategy requires testing of glass formulations spanning a number of waste loadings, compositions, and additives over the range of expected waste compositions. The resulting glasses will then be characterized and compared to processing and performance specifications yet to be developed. This report documents the glass formulation work conducted at PNL in fiscal years 1994 and 1995 including glass formulation optimization, minor component impacts evaluation, Phase 1 and Phase 2 melter vendor glass development, liquidus temperature and crystallization kinetics determination. This report also summarizes relevant work at PNNL on high-iron glasses for Hanford tank wastes conducted through the Mixed Waste Integrated Program and work at Savannah River Technology Center to optimize glass formulations using a Plackett-Burnam experimental design

  5. Magnetic Glass Ceramics by Sintering of Borosilicate Glass and Inorganic Waste

    OpenAIRE

    Ponsot, In?s M. M. M.; Pontikes, Yiannis; Baldi, Giovanni; Chinnam, Rama K.; Detsch, Rainer; Boccaccini, Aldo R.; Bernardo, Enrico

    2014-01-01

    Ceramics and glass ceramics based on industrial waste have been widely recognized as competitive products for building applications; however, there is a great potential for such materials with novel functionalities. In this paper, we discuss the development of magnetic sintered glass ceramics based on two iron-rich slags, coming from non-ferrous metallurgy and recycled borosilicate glass. The substantial viscous flow of the glass led to dense products for rapid treatments at relatively low te...

  6. Glasses and ceramics for immobilisation of radioactive wastes for disposal

    International Nuclear Information System (INIS)

    Johnson, K.D.B.; Marples, J.A.C.

    1979-05-01

    The U.K. Research Programme on Radioactive Waste Management includes the development of processes for the conversion of high level liquid reprocessing wastes from thermal and fast reactors to borosilicate glasses. The properties of these glasses and their behaviour under storage and disposal conditions have been examined. Methods for immobilising activity from other wastes by conversion to glass or ceramic forms is described. The U.K. philosophy of final solutions to waste management and disposal is presented. (author)

  7. Investigation on Compressive Strength of Special Concrete made with Crushed Waste Glass

    Directory of Open Access Journals (Sweden)

    Mohd Sani Mohd Syahrul Hisyam

    2015-01-01

    Full Text Available Special concrete is the type of concrete that produced by using waste material or using unusual techniques/method of preparation. Special concrete made with waste material is becoming popular in a construction site. This is because the special concrete is selected due to quality, integrity, economic factor and environmental factor. The waste glass is selected as an additional material to provide a good in compressive strength value. The compressive strength is the importance of mechanical properties of concrete and typically the concrete is sustained and stiffed in compression load. The significant issue to utilize the waste glass from the automotive windscreen is to improve the strength of concrete. The waste glass is crushed to become 5 mm size and recognised as crushed waste glass that be used in concrete as additional material. The main objective of the study is to determine the appropriate percentage of crushed waste glass in concrete grade, 30 in order to enhance the compressive strength. There are four mixes of concrete that contained of crushed waste glass with percentage of 2 %, 4 %, 6 % and 8 % and one control mix with 0 % of crushed waste glass. As the result, crushed waste glass with an additional 4 % in concrete is reported having a higher value of compressive strength in early and mature stage. In addition, if the percentage of crushed glass wastes in concrete increases and it leads to a reduction in the workability of concrete.

  8. Recycling of waste glass as a partial replacement for fine aggregate in concrete.

    Science.gov (United States)

    Ismail, Zainab Z; Al-Hashmi, Enas A

    2009-02-01

    Waste glass creates serious environmental problems, mainly due to the inconsistency of waste glass streams. With increasing environmental pressure to reduce solid waste and to recycle as much as possible, the concrete industry has adopted a number of methods to achieve this goal. The properties of concretes containing waste glass as fine aggregate were investigated in this study. The strength properties and ASR expansion were analyzed in terms of waste glass content. An overall quantity of 80 kg of crushed waste glass was used as a partial replacement for sand at 10%, 15%, and 20% with 900 kg of concrete mixes. The results proved 80% pozzolanic strength activity given by waste glass after 28 days. The flexural strength and compressive strength of specimens with 20% waste glass content were 10.99% and 4.23%, respectively, higher than those of the control specimen at 28 days. The mortar bar tests demonstrated that the finely crushed waste glass helped reduce expansion by 66% as compared with the control mix.

  9. Polycarboxylic acids as network modifiers for water durability improvement of inorganic-organic hybrid tin-silico-phosphate low-melting glasses

    International Nuclear Information System (INIS)

    Menaa, Bouzid; Mizuno, Megumi; Takahashi, Masahide; Tokuda, Yomei; Yoko, Toshinobu

    2006-01-01

    We investigated the water durability of the inorganic-organic hybrid tin-silico-phosphate glasses Me 2 SiO-SnO-P 2 O 5 (Me designs the organic methyl group) doped with organic acids (salicylic acid (SA), tartaric acid (TA), citric acid (Canada) and butane tetracarboxylic acid (BTCA)) containing one or more of carboxylic groups per molecule. The structure, thermal properties and durability of the final glasses obtained via a non-aqueous acid-base reaction were discussed owing to the nature and the concentration of the acid added. 29 Si magic angle spinning (MAS) NMR and 31 P MAS NMR spectra, respectively, showed clearly a modification of the network in the host glass matrix of the Me 2 SiO-SnO-P 2 O 5 system. The polycondensation enhancement to form -P-O-Si-O-P- linkages (PSP) and the increase of the Q 2 unit (two bridging oxygens per phosphorus atom) over the Q 3 unit (three bridging oxygens per phosphorus atom) as a function of the acid in the order SA 2 SiO-SnO-P 2 O 5 matrix. In addition, this structural change is accompanied by a decrease of the coefficient of thermal expansion and an increase of the water durability of the glasses with the acids containing a large number of carboxylic groups per molecule. The presence of carboxylic groups of the acid acting as network modifier may retard the movement of water molecules through the glasses due to the steric hindrance strengthening the PSP connections in a chain-like structure

  10. Properties and characteristics of high-level waste glass

    International Nuclear Information System (INIS)

    Ross, W.A.

    1977-01-01

    This paper has briefly reviewed many of the characteristics and properties of high-level waste glasses. From this review, it can be noted that glass has many desirable properties for solidification of high-level wastes. The most important of these include: (1) its low leach rate; (2) the ability to tolerate large changes in waste composition; (3) the tolerance of anticipated storage temperatures; (4) its low surface area even after thermal shock or impact

  11. Conversion of radwaste-loaded zeolites into a borosilicate glass to improve their properties for final disposal (preliminary results)

    International Nuclear Information System (INIS)

    Kazemian, H.; Ghannadi Maraghe, M.; Mallah, M.H.; Firooz Zare, M.; Kooshkestani, R.; Naghavi, S.Sh.

    2002-01-01

    Research was undertaken to fix radioisotopes from simulated nuclear waste streams into durable and stable borosilicate glass matrices based on the conversion of different type of zeolites. It was found that the selectivity of Iranian clinoptilolite and the relevant synthetic zeolite toward cesium and strontium were quite good whereas the leach resistance of the loaded zeolite was relatively poor. To improve the leach resistance of the used zeolites, conversion of the spent exchangers into borosilicate glass was investigated. Results obtained in the non active bench scale tests phase were promising. It is concluded that spent zeolites loaded with radioactive materials can be converted into a durable, high leach resistant borosilicate glass, a proper matrix for final disposal of nuclear waste. (author)

  12. Simulation used to qualify nuclear waste glass for disposal

    International Nuclear Information System (INIS)

    Reimus, T.W.; Kuhn, W.L.

    1987-07-01

    A hypothetical vitrification system was simulated errors associated with controlling and predicting the composition of the nuclear waste glass produced in the system. The composition of the glass must fall within certain limits to qualify for permanent geologic disposal. The estimated error in predicting the concentrations of various constituents in the glass was 2% to 8%, depending on the strategy for sampling and analyzing the feed and on the assumed magnitudes of the process uncertainties. The estimated error in controlling the glass composition was 2% to 9%, depending on the strategy for sampling and analyzing the waste and on the assumed magnitudes of the uncertainties. This work demonstrates that simulation techniques can be used to assist in qualifying nuclear waste glass for disposal. 3 refs., 2 figs., 4 tabs

  13. The Production of Advanced Glass Ceramic HLW Forms using Cold Crucible Induction Melter

    Energy Technology Data Exchange (ETDEWEB)

    Veronica J Rutledge; Vince Maio

    2013-10-01

    Cold Crucible Induction Melters (CCIMs) will favorably change how High-Level radioactive Waste (from nuclear fuel recovery) is treated in the 21st century. Unlike the existing Joule-Heated Melters (JHMs) currently in operation for the glass-based immobilization of High-Level Waste (HLW), CCIMs offer unique material features that will increase melt temperatures, increase throughput, increase mixing, increase loading in the waste form, lower melter foot prints, eliminate melter corrosion and lower costs. These features not only enhance the technology for producing HLW forms, but also provide advantageous attributes to the waste form by allowing more durable alternatives to glass. This paper discusses advantageous features of the CCIM, with emphasis on features that overcome the historical issues with the JHMs presently utilized, as well as the benefits of glass ceramic waste forms over borosilicate glass waste forms. These advantages are then validated based on recent INL testing to demonstrate a first-of-a-kind formulation of a non-radioactive ceramic-based waste form utilizing a CCIM.

  14. Studies on the Potential of Waste Soda Lime Silica Glass in Glass Ionomer Cement Production

    Directory of Open Access Journals (Sweden)

    V. W. Francis Thoo

    2013-01-01

    Full Text Available Glass ionomer cements (GIC are produced through acid base reaction between calcium-fluoroaluminosilicate glass powder and polyacrylic acid (PAA. Soda lime silica glasses (SLS, mainly composed of silica (SiO2, have been utilized in this study as the source of SiO2 for synthesis of Ca-fluoroaluminosilicate glass. Therefore, the main objective of this study was to investigate the potential of SLS waste glass in producing GIC. Two glasses, GWX 1 (analytical grade SiO2 and GWX 2 (replacing SiO2 with waste SLS, were synthesized and then characterized using X-ray diffraction (XRD and energy dispersive X-ray (EDX. Synthesized glasses were then used to produce GIC, in which the properties were characterized using Fourier transform infrared spectroscopy (FT-IR and compressive test (from 1 to 28 days. XRD results showed that amorphous glass was produced by using SLS waste glass (GWX 2, which is similar to glass produced using analytical grade SiO2 (GWX 1. Results from FT-IR showed that the setting reaction of GWX 2 cements is slower compared to cement GWX 1. Compressive strengths for GWX 1 cements reached up to 76 MPa at 28 days, whereas GWX 2 cements showed a slightly higher value, which is 80 MPa.

  15. Tests with ceramic waste form materials made by pressureless consolidation

    International Nuclear Information System (INIS)

    Lewis, M. A.; Hash, M. C.; Hebden, A. S.; Ebert, W. L.

    2002-01-01

    A multiphase waste form referred to as the ceramic waste form (CWF) will be used to immobilize radioactively contaminated salt wastes recovered after the electrometallurgical treatment of spent sodium-bonded nuclear fuel. The CWF is made by first occluding salt in zeolite and then encapsulating the zeolite in a borosilicate binder glass. A variety of surrogate CWF materials were made using pressureless consolidation (PC) methods for comparison with CWF consolidated using a hot isostatic press (HIP) method and to study the effects of glass/zeolite batching ratio and processing conditions on the physical and chemical properties of the resulting materials. The data summarized in this report will also be used to support qualification of the PC CWF for disposal in the proposed federal high-level radioactive waste repository at Yucca Mountain. The phase composition and microstructure of HIP CWF and PC CWF are essentially identical: both are composed of about 70% sodalite, 25% binder glass, and a 5% total of inclusion phases (halite, nepheline, and various oxides and silicates). The primary difference is that PC CWF materials have higher porosities than HIP CWFs. The product consistency test (PCT) that was initially developed to monitor homogeneous glass waste forms was used to measure the chemical durabilities of the CWF materials. Series of replicate tests with several PC CWF materials indicate that the PCT can be conducted with the same precision with CWF materials as with borosilicate glasses. Short-term (7-day) PCTs were used to evaluate the repeatability of making the PC CWF and the effects of the glass/zeolite mass ratio, process temperature, and processing time on the chemical durability. Long-term (up to 1 year) PCTs were used to compare the durabilities of HIP and PC CWFs and to estimate the apparent solubility limit for the PC CWF that is needed for modeling. The PC and HIP CWF materials had similar disabilities, based on the release of silicon in long

  16. Glasses Containing Iron (II, III) Oxides For Immobilization Of Radioactive Technetium

    International Nuclear Information System (INIS)

    Kruger, A.A.; Heo, J.; Xu, K.; Choi, J.K.; Hrma, P.R.; Um, W.

    2011-01-01

    Technetium-99 (Tc-99) has posed serious environmental threats as US Department of Energy's high-level waste. This work reports the vitrification of Re, as surrogate for Tc-99, by iron-borosilicate and iron-phosphate glasses, respectively. Iron-phosphate glasses can dissolve Re as high as ∼ 1.2 wt. %, which can become candidate waste forms for Tc-99 disposal, while borosilicate glasses can retain less than 0.1 wt. % of Re due to high melting temperature and long melting duration. Vitrification of Re as Tc-99's mimic was investigated using iron-borosilicate and iron-phosphate glasses. The retention of Re in borosilicate glasses was less than 0.1 wt. % and more than 99 wt. % of Re were volatilized due to high melting temperature and long melting duration. Because the retention of Re in iron-phosphate glasses is as high as 1.2 wt. % and the volatilization is reduced down to ∼50 wt. %, iron-phosphate glasses can be one of the glass waste form candidates for Tc (or Re) disposal. The investigations of chemical durability and leaching test of iron-phosphate glasses containing Re are now underway to test the performance of the waste form.

  17. Fracturing of simulated high-level waste glass in canisters

    International Nuclear Information System (INIS)

    Peters, R.D.; Slate, S.C.

    1981-09-01

    Waste-glass castings generated from engineering-scale developmental processes at the Pacific Northwest Laboratory are generally found to have significant levels of cracks. The causes and extent of fracturing in full-scale canisters of waste glass as a result of cooling and accidental impact are discussed. Although the effects of cracking on waste-form performance in a repository are not well understood, cracks in waste forms can potentially increase leaching surface area. If cracks are minimized or absent in the waste-glass canisters, the potential for radionuclide release from the canister package can be reduced. Additional work on the effects of cracks on leaching of glass is needed. In addition to investigating the extent of fracturing of glass in waste-glass canisters, methods to reduce cracking by controlling cooling conditions were explored. Overall, the study shows that the extent of glass cracking in full-scale, passively-cooled, continuous melting-produced canisters is strongly dependent on the cooling rate. This observation agrees with results of previously reported Pacific Northwest Laboratory experiments on bench-scale annealed canisters. Thus, the cause of cracking is principally bulk thermal stresses. Fracture damage resulting from shearing at the glass/metal interface also contributes to cracking, more so in stainless steel canisters than in carbon steel canisters. This effect can be reduced or eliminated with a graphite coating applied to the inside of the canister. Thermal fracturing can be controlled by using a fixed amount of insulation for filling and cooling of canisters. In order to maintain production rates, a small amount of additional facility space is needed to accomodate slow-cooling canisters. Alternatively, faster cooling can be achieved using the multi-staged approach. Additional development is needed before this approach can be used on full-scale (60-cm) canisters

  18. Surface layer effects on waste glass corrosion

    International Nuclear Information System (INIS)

    Feng, X.

    1993-01-01

    Water contact subjects waste glass to chemical attack that results in the formation of surface alteration layers. Two principal hypotheses have been advanced concerning the effect of surface alteration layers on continued glass corrosion: (1) they act as a mass transport barrier and (2) they influence the chemical affinity of the glass reaction. In general, transport barrier effects have been found to be less important than affinity effects in the corrosion of most high-level nuclear waste glasses. However, they can be important under some circumstances, for example, in a very alkaline solution, in leachants containing Mg ions, or under conditions where the matrix dissolution rate is very low. The latter suggests that physical barrier effect may affect the long-term glass dissolution rate. Surface layers influence glass reaction affinity through the effects of the altered glass and secondary phases on the solution chemistry. The reaction affinity may be controlled by various precipitates and crystalline phases, amorphous silica phases, gel layer, or all the components of the glass. The surface alteration layers influence radionuclide release mainly through colloid formation, crystalline phase incorporation, and gel layer retention. This paper reviews current understanding and uncertainties

  19. Radiation effects in glass waste forms for high-level waste and plutonium disposal

    International Nuclear Information System (INIS)

    Weber, W.J.; Ewing, R.C.

    1997-01-01

    A key challenge in the permanent disposal of high-level waste (HLW), plutonium residues/scraps, and excess weapons plutonium in glass waste forms is the development of predictive models of long-term performance that are based on a sound scientific understanding of relevant phenomena. Radiation effects from β-decay and α-decay can impact the performance of glasses for HLW and Pu disposition through the interactions of the α-particles, β-particles, recoil nuclei, and γ-rays with the atoms in the glass. Recently, a scientific panel convened under the auspices of the DOE Council on Materials Science to assess the current state of understanding, identify important scientific issues, and recommend directions for research in the area of radiation effects in glasses for HLW and Pu disposition. The overall finding of the panel was that there is a critical lack of systematic understanding on radiation effects in glasses at the atomic, microscopic, and macroscopic levels. The current state of understanding on radiation effects in glass waste forms and critical scientific issues are presented

  20. Characterization of Incorporation the Glass Waste in Adhesive Mortar

    Science.gov (United States)

    Santos, D. P.; Azevedo, A. R. G.; Hespanhol, R. L.; Alexandre, J.

    Ehe search for reuse generated waste in urban centers, intending to preserve natural resources, has remained fairly constant, both in context of preventing exploitation of resources as the emplacement of waste on the environment. Glass waste glass created a serious environmental problem, mainly because of inconsistency of its flows. Ehe use of this product as a mineral additive, finely ground, cement replacement and aggregate is a promising direction for recycling. This work aims to study the influence of glass waste from cutting process in adhesive mortar, replacing part of cement. Ehe glass powder is used replacing Portland cement at 10, 15 and 20% by mass. Ehe produced mortars will be evaluated its performance in fresh and hardened states through tests performed in laboratory. Ehe selected feature is indicated by producers of additive and researchers to present good results when used as adhesive mortar.

  1. Preliminary assessment of the controlled release of radionuclides from waste packages containing borosilicate waste glass

    International Nuclear Information System (INIS)

    Strachan, D.M.; McGrail, B.P.; Apted, M.J.; Engle, D.W.; Eslinger, P.W.

    1990-06-01

    The purpose of this report is to provide a preliminary assessment of the release-rate for an engineered barriers subsystem (EBS) containing waste packages of defense high-level waste borosilicate glass at geochemical and hydrological conditions similar to the those at Yucca Mountain. The relationship between the proposed Waste Acceptance Preliminary Specifications (WAPS) test of glass- dissolution rate and compliance with the NRC's release-rate criterion is also evaluated. Calculations are reported for three hierarchical levels: EBS analysis, waste-package analysis, and waste-glass analysis. The following conclusions identify those factors that most acutely affect the magnitude of, or uncertainty in, release-rate performance

  2. Vitrification of liquid waste from nuclear power plants

    International Nuclear Information System (INIS)

    Sheng Jiawei; Choi, Kwansik; Song, Myung-Jae

    2001-01-01

    Glass is an acceptable waste form to solidify the low-level waste from nuclear power plants (NPPs) because of the simplicity of processing and its unique ability to accept a wide variety of waste streams. Vitrification is being considered to solidify the high-boron-containing liquid waste generated from Korean NPPs. This study dealt with the development of a glass formulation to solidify the liquid waste. Studies were conducted in a borosilicate glass system. Crucible studies have been performed with surrogate waste. Several developed glass frits were evaluated to determine their suitability for vitrifying the liquid waste. The results indicated that the 20 wt% waste oxides loading required could not be obtained using these glass frits. Flyash produced from coal-burning electric power stations, whose major components are SiO 2 and Al 2 O 3 , is a desirable glass network former. Detailed product evaluations including waste loading, homogeneity, chemical durability and viscosity, etc., were carried out on selected formulations using flyash. Up to 30 wt% of the waste oxides was successfully solidified into the flyash after the addition of 5-10 wt% Na 2 O at 1200 deg. C

  3. The production of advanced glass ceramic HLW forms using cold crucible induction melter

    International Nuclear Information System (INIS)

    Rutledge, V.J.; Maio, V.

    2013-01-01

    Cold Crucible Induction Melters (CCIM) will favorably change how High-Level radioactive Waste (from nuclear fuel recovery) is treated in a near future. Unlike the existing Joule-Heated Melters (JHM) currently in operation for the glass-based immobilization of High-Level Waste (HLW), CCIM offers unique material features that will increase melt temperatures, increase throughput, increase mixing, increase loading in the waste form, lower melter foot prints, eliminate melter corrosion and lower costs. These features not only enhance the technology for producing HLW forms, but also provide advantageous attributes to the waste form by allowing more durable alternatives to glass. It is concluded that glass ceramic waste forms that are tailored to immobilize fission products of HLW can be can be made from the HLW processed with the CCIM. The advantageous higher temperatures reached with the CCIM and unachievable with JHM allows the lanthanides, alkali, alkaline earths, and molybdenum to dissolve into a molten glass. Upon controlled cooling they go into targeted crystalline phases to form a glass ceramic waste form with higher waste loadings than achievable with borosilicate glass waste forms. Natural cooling proves to be too fast for the formation of all targeted crystalline phases

  4. Glass Formulation Development for INEEL Sodium-Bearing Waste

    International Nuclear Information System (INIS)

    Vienna, J.D.; Schweiger, M.J.; Smith, D.E.; Smith, H.D.; Crum, J.V.; Peeler, D.K.; Reamer, I.A.; Musick, C.A.; Tillotson, R.D.

    1999-01-01

    For about four decades, radioactive wastes have been collected and calcined from nuclear fuels reprocessing at the Idaho Nuclear Technology and Engineering Center (INTEC), formerly Idaho Chemical Processing Plant (ICPP). Over this time span, secondary radioactive wastes have also been collected and stored as liquid from decontamination, laboratory activities, and fuel-storage activities. These liquid wastes are collectively called sodium-bearing wastes (SBW). About 5.7 million liters of these wastes are temporarily stored in stainless steel tanks at the Idaho National Engineering and Environmental Laboratory (INEEL). Vitrification is being considered as an immobilization step for SBW with a number of treatment and disposal options. A systematic study was undertaken to develop a glass composition to demonstrate direct vitrification of INEEL's SBW. The objectives of this study were to show the feasibility of SBW vitrification, not a development of an optimum formulation. The waste composition is relatively high in sodium, aluminum, and sulfur. A specific composition and glass property restrictions, discussed in Section 2, were used as a basis for the development. Calculations based on first-order expansions of selected glass properties in composition and some general tenets of glass chemistry led to an additive (fit) composition (68.69 mass % SiO 2 , 14.26 mass% B 2 O 3 , 11.31 mass% Fe 2 O 3 , 3.08 mass% TiO 2 , and 2.67 mass % Li 2 O) that meets all property restrictions when melted with 35 mass % of SBW on an oxide basis, The glass was prepared using oxides, carbonates, and boric acid and tested to confirm the acceptability of its properties. Glass was then made using waste simulant at three facilities, and limited testing was performed to test and optimize processing-related properties and confirm results of glass property testing. The measured glass properties are given in Section 4. The viscosity at 1150 C, 5 Pa·s, is nearly ideal for waste-glass processing in

  5. The solidification of radioactive waste

    International Nuclear Information System (INIS)

    Nagaya, Kiichi; Fujimoto, Yoshio; Hashimoto, Yasuo; Nomura, Ichiro

    1985-01-01

    A previous paper covered the decomposition and vitrification of Na 2 SO 4 (the primary component of the liquid waste from BWR) with silica. Now, in order to establish an integrated treatment system for the radioactive waste from BWR, this paper examines the effects of combining incinerator ash and other incinerator wastes with radioactive waste on the durability of the final vitrified products. A bench scale test plat consisting of a waiped file evaporator/dryer, a Joule-heated glass melter and SO 2 absorber was therefore put into operation and run safety for a period of 3000 hours. The combination of the radioactive waste with incinerator ash and the secondary waste of the incinerator was found to make no difference on the durability of the final vitrified products effecting no increase or decrease. Durability similar to that displayed in the beaker tests was proven, with the final vitrified products exhibiting a leaching rate less than 3 x 10 -4 g/cm 2 /day at 95 deg C. (author)

  6. Physical and chemical characterization of borosilicate glasses containing Hanford high-level wastes

    International Nuclear Information System (INIS)

    Kupfer, M.J.; Palmer, R.A.

    1980-10-01

    Scouting studies are being performed to develop and evaluate silicate glass forms for immobilization of Hanford high-level wastes. Detailed knowledge of the physical and chemical properties of these glasses is required to assess their suitability for long-term storage or disposal. Some key properties to be considered in selecting a glass waste form include leach resistance, resistance to radiation, microstructure (includes devitrification behavior or crystallinity), homogeneity, viscosity, electrical resistivity, mechanical ruggedness, thermal expansion, thermal conductivity, density, softening point, annealing point, strain point, glass transformation temperature, and refractive index. Other properties that are important during processing of the glass include volatilization of glass and waste components, and corrosivity of the glass on melter components. Experimental procedures used to characterize silicate waste glass forms and typical properties of selected glass compositions containing simulated Hanford sludge and residual liquid wastes are presented. A discussion of the significance and use of each measured property is also presented

  7. Comparison of SRP high-level waste disposal costs for borosilicate glass and crystalline ceramic waste forms

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1982-04-01

    An evaluation of costs for the immobilization and repository disposal of SRP high-level wastes indicates that the borosilicate glass waste form is less costly than the crystalline ceramic waste form. The wastes were assumed immobilized as glass with 28% waste loading in 10,300 reference 24-in.-diameter canisters or as crystalline ceramic with 65% waste loading in either 3400 24-in.-diameter canisters or 5900 18-in.-diameter canisters. After an interim period of onsite storage, the canisters would be transported to the federal repository for burial. Total costs in undiscounted 1981 dollars of the waste disposal operations, excluding salt processing for which costs are not yet well defined, were about $2500 million for the borosilicate glass form in reference 24-in.-diameter canisters, compared to about $2900 million for the crystalline ceramic form in 24-in.-diameter canisters and about $3100 million for the crystalline ceramic form in 18-in.-diameter canisters. No large differences in salt processing costs for the borosilicate glass and crystalline ceramic forms are expected. Discounting to present values, because of a projected 2-year delay in startup of the DWPF for the crystalline ceramic form, preserved the overall cost advantage of the borosilicate glass form. The waste immobilization operations for the glass form were much less costly than for the crystalline ceramic form. The waste disposal operations, in contrast, were less costly for the crystalline ceramic form, due to fewer canisters requiring disposal; however, this advantage was not sufficient to offset the higher development and processing costs of the crystalline ceramic form. Changes in proposed Nuclear Regulatory Commission regulations to permit lower cost repository packages for defense high-level wastes would decrease the waste disposal costs of the more numerous borosilicate glass forms relative to the crystalline ceramic forms

  8. High level waste forms: glass marbles and thermal spray coatings

    International Nuclear Information System (INIS)

    Treat, R.L.; Oma, K.H.; Slate, S.C.

    1982-01-01

    A process that converts high-level waste to glass marbles and then coats the marbles has been developed at Pacific Northwest Laboratory (PNL) under sponsorship of the US Department of Energy. The process consists of a joule-heated glass melter, a marble-making device based on a patent issued to Corning Glass Works, and a coating system that includes a plasma spray coater and a marble tumbler. The process was developed under the Alternative Waste Forms Program which strived to improve upon monolithic glass for immobilizing high-level wastes. Coated glass marbles were found to be more leach-resistant, and the marbles, before coating were found to be very homogeneous, highly impact resistant, and conductive to encapsulation in a metal matric for improved heat transfer and containment. Marbles are also ideally suited for quality assurance and recycling. However, the marble process is more complex, and marbles require a larger number of canisters for waste containment and have a higher surface area than do glass monoliths

  9. Computer Modeling Of High-Level Waste Glass Temperatures Within DWPF Canisters During Pouring And Cool Down

    International Nuclear Information System (INIS)

    Amoroso, J.

    2011-01-01

    This report describes the results of a computer simulation study to predict the temperature of the glass at any location inside a DWPF canister during pouring and subsequent cooling. These simulations are an integral part of a larger research focus aimed at developing methods to predict, evaluate, and ultimately suppress nepheline formation in HLW glasses. That larger research focus is centered on holistically understanding nepheline formation in HLW glass by exploring the fundamental thermal and chemical driving forces for nepheline crystallization with respect to realistic processing conditions. Through experimental work, the goal is to integrate nepheline crystallization potential in HLW glass with processing capability to ultimately optimize waste loading and throughput while maintaining an acceptable product with respect to durability. The results of this study indicated severe temperature gradients and prolonged temperature dwell times exist throughout different locations in the canister and that the time and temperatures that HLW glass is subjected to during processing is a function of pour rate. The simulations indicate that crystallization driving forces are not uniform throughout the glass volume in a DWPF (or DWPF-like) canister and illustrate the importance of considering overall kinetics (chemical and thermal driving forces) of nepheline formation when developing methods to predict and suppress its formation in HLW glasses. The intended path forward is to use the simulation data both as a driver for future experimental work and, as an investigative tool for evaluating the impact of experimental results. Simulation data will be used to develop laboratory experiments to more acutely evaluate nepheline formation in HLW glass by incorporating the simulated temperatures throughout the canister into the laboratory experiments. Concurrently, laboratory experiments will be performed to identify nepheline crystallization potential in HLW glass as a function of

  10. Ceramics: Durability and radiation effects

    Energy Technology Data Exchange (ETDEWEB)

    Ewing, R.C.; Lutze, W. [Univ. of New Mexico, Albuquerque, NM (United States); Weber, W.J. [Pacific Northwest National Lab., Richland, WA (United States)

    1996-05-01

    At present, there are three seriously considered options for the disposition of excess weapons plutonium: (1) incorporation, partial burn-up and direct disposal of MOX-fuel; (2) vitrification with defense waste and disposal as glass {open_quotes}logs{close_quotes}; (3) deep borehole disposal. The first two options provide a safeguard due to the high activity of fission products in the irradiated fuel and the defense waste. The latter option has only been examined in a preliminary manner, and the exact form of the plutonium has not been identified. In this paper, we review the potential for the immobilization of plutonium in highly durable crystalline ceramics apatite, pyrochlore, zirconolite, monazite and zircon. Based on available data, we propose zircon as the preferred crystalline ceramic for the permanent disposition of excess weapons plutonium.

  11. X-ray tomography of feed-to-glass transition of simulated borosilicate waste glasses

    Czech Academy of Sciences Publication Activity Database

    Harris, W.H.; Guillen, D.P.; Kloužek, Jaroslav; Pokorný, P.; Yano, T.; Lee, S.; Schweiger, M. J.; Hrma, P.

    2017-01-01

    Roč. 100, č. 9 (2017), s. 3883-3894 ISSN 0002-7820 Institutional support: RVO:67985891 Keywords : borosilicate glass * computed tomography * glass melting * morphology * nuclear waste * X-ray Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass OBOR OECD: Ceramics Impact factor: 2.841, year: 2016

  12. Immobilization of {sup 99}Tc (Re) using Iron-Phosphate Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Jong; Xu, Kai; Um, Woo Yong; Hrma, Pavel [Pohang Univ. of Science and Technology, Pohang (Korea, Republic of)

    2012-05-15

    Technetium-99 ({sup 99}Tc) is a fission product artificially generated during the irradiation of {sup 235}U for commercial power production or {sup 239}Pu for nuclear weapons. Under oxidizing conditions, the dominant species of Tc, the pertechnetate anion (TcO{sub 4} {sup -}), is highly soluble in ground water and thus easily transports through the geologic systems. In addition, because of its high fission yield ({approx}6 %) and long half-life (2.1x10{sup 5} yr), immobilization of {sup 99}Tc has been investigated for decades. Several waste forms such as metallic alloys, sintered titanate ceramics and chemically bonded phosphate ceramics have been proposed to encapsulate {sup 99}Tc. They have not yet been realized in the industrial-scale, mostly either due to the high volatilization of {sup 99}Tc during high temperature process (>1300 .deg. C), or the low {sup 99}Tc loading. Iron-phosphate (FeP) glasses have been developed as alternative waste forms because of their chemical durability equivalent to borosilicate glasses. Additionally, vitrification of radioactive waste by FeP glasses can be done at a relatively low temperature ({approx}1000 .deg. C) and the low-temperature process can reduce the volatilization of {sup 99}Tc significantly. Thus, this work reports the immobilization of {sup 99}Tc by FeP glasses using rhenium (Re) as a surrogate. We also examine the chemical durability of Re-containing FeP glasses using product consistency test (PCT). Experimental results reveal that FeP glass can become a promising candidate for immobilizing {sup 99}Tc

  13. Devitrification of defense nuclear waste glasses: role of melt insolubles

    International Nuclear Information System (INIS)

    Bickford, D.F.; Jantzen, C.M.

    1985-01-01

    Time-temperature-transformation (TTT) curves have been determined for simulated nuclear waste glasses bounding the compositional range in the Defense Waste Processing Facility (DWPF). Formulations include all of the minor chemical elements such as ruthenium and chromium which have limited solubility in borosilicate glasses. Heterogeneous nucleation of spinel on ruthenium dioxide, and subsequent nucleation of acmite on spinel is the major devitrification path. Heterogeneous nucleation on melt insolubles causes more rapid growth of crystalline devitrification phases, than in glass free of melt insolubles. These studies point out the importance of simulating waste glass composition and processing as accurately as possible to obtain reliable estimates of glass performance. 11 refs., 8 figs., 1 tab

  14. Wastes based glasses and glass-ceramics

    Directory of Open Access Journals (Sweden)

    Barbieri, L.

    2001-12-01

    Full Text Available Actually, the inertization, recovery and valorisation of the wastes coming from municipal and industrial processes are the most important goals from the environmental and economical point of view. An alternative technology capable to overcome the problem of the dishomogeneity of the raw material chemical composition is the vitrification process that is able to increase the homogeneity and the constancy of the chemical composition of the system and to modulate the properties in order to address the reutilization of the waste. Moreover, the glasses obtained subjected to different controlled thermal treatments, can be transformed in semy-cristalline material (named glass-ceramics with improved properties with respect to the parent amorphous materials. In this review the tailoring, preparation and characterization of glasses and glass-ceramics obtained starting from municipal incinerator grate ash, coal and steel fly ashes and glass cullet are described.

    Realmente la inertización, recuperación y valorización de residuos que proceden de los procesos de incineración de residuos municipales y de residuos industriales son metas importantes desde el punto de vista ambiental y económico. Una tecnología alternativa capaz de superar el problema de la heterogeneidad de la composición química de los materiales de partida es el proceso de la vitrificación que es capaz de aumentar la homogeneidad y la constancia de la composición química del sistema y modular las propiedades a fin de la reutilización del residuo. En este artículo se presentan los resultados de vitrificación en que los vidrios fueron sometidos a tratamientos térmicos controlados diferentes, de manera que se transforman en materiales semicristalinos (también denominados vitrocerámicos con mejores propiedades respecto a los materiales amorfos originales. En esta revisión se muestra el diseño, preparación y caracterización de vidrios y vitrocerámicos partiendo de

  15. Geochemical and petrographic studies and the relationships to durability and leach resistance of vitrified products from the in situ vitrification process

    International Nuclear Information System (INIS)

    Timmons, D.M.; Thompson, L.E.

    1996-01-01

    Soil and sludge contaminated with hazardous and radioactive materials from sites in the United States and Australia were vitrified using in situ vitrification. Some of the resulting products were subjected to detailed geochemical, leach and durability testing using a variety of analytical techniques. The leach resistance and durability performance was compared to that of vitrified high level waste with borosilicate composition. Particular attention was given to crystallization behavior, the effects of crystallization on residual melt chemistry and how crystallization influences the behavior of contaminant ions. The results of this work show that the vitrified material studied has superior chemical durability and leach resistance relative to typical borosilicate waste glasses. Crystallization behavior was variable depending upon melt chemistry and cooling history. Crystallization was not observed to adversely affect chemical durability or leach resistance

  16. TIME-TEMPERATURE-TRANSFORMATION (TTT) DIAGRAMS FOR FUTURE WASTE COMPOSITIONS

    International Nuclear Information System (INIS)

    Billings, A.; Edwards, T.

    2010-01-01

    As a part of the Waste Acceptance Product Specifications (WAPS) for Vitrified High-Level Waste Forms defined by the Department of Energy - Office of Environmental Management, the waste form stability must be determined for each of the projected high-level waste (HLW) types at the Savannah River Site (SRS). Specifically, WAPS 1.4.1 requires the glass transition temperature (T g ) to be defined and time-temperature-transformation (TTT) diagrams to be developed. The T g of a glass is an indicator of the approximate temperature where the supercooled liquid converts to a solid on cooling or conversely, where the solid begins to behave as a viscoelastic solid on heating. A TTT diagram identifies the crystalline phases that can form as a function of time and temperature for a given waste type or more specifically, the borosilicate glass waste form. In order to assess durability, the Product Consistency Test (PCT) was used and the durability results compared to the Environmental Assessment (EA) glass. The measurement of glass transition temperature and the development of TTT diagrams have already been performed for the seven Defense Waste Processing Facility (DWPF) projected compositions as defined in the Waste Form Compliance Plan (WCP) and in SRNL-STI-2009-00025. Additional phase transformation information exists for other projected compositions, but overall these compositions did not cover composition regions estimated for future waste processing. To develop TTT diagrams for future waste types, the Savannah River National Laboratory (SRNL) fabricated two caches of glass from reagent grade oxides to simulate glass compositions which would be likely processed with and without Al dissolution. These were used for glass transition temperature measurement and TTT diagram development. The glass transition temperatures of both glasses were measured using differential scanning calorimetry (DSC) and were recorded to be 448 C and 452 C. Using the previous TTT diagrams as guidance

  17. The chemistry of copper chalcogenides in waste glasses

    International Nuclear Information System (INIS)

    Schreiber, H.D.; Lambert, H.W.

    1994-01-01

    The solubilities of copper chalcogenides (CuS, CuSe, CuTe) were measured in a glass melt which is representative of those proposed for nuclear waste immobilization and circuit board vitrification. CuTe is more soluble than CuS and CuSe in the glass melt under relatively oxidizing conditions. However, the solubilities of all the copper chalcogenides in the glass melt are virtually identical at reducing conditions, probably a result of the redox-controlled solubility of copper metal in all cases. The redox chemistry of a glass melt coexisting with an immiscible copper chalcogenide depends primarily on the prevailing oxygen fugacity, not on the identity of the chalcogenide. The target concentration of less than 0.3 to 0.5 wt% copper in the waste glass should eliminate the precipitation of copper chalcogenides during processing

  18. Characterization of projected DWPF glasses heat treated to simulate canister centerline cooling

    International Nuclear Information System (INIS)

    Marra, S.L.; Jantzen, C.M.

    1992-05-01

    Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). Eventually these canistered waste forms will be sent to a geologic repository for final disposal. In order to assure acceptability by the repository, the Department of Energy has defined requirements which DWPF canistered waste forms must meet. These requirements are the Waste Acceptance Preliminary Specifications (WAPS). The WAPS require DWPF to identify the crystalline phases expected to be present in the final glass product. Knowledge of the thermal history of the borosilicate glass during filling and cooldown of the canister is necessary to determine the amount and type of crystalline phases present in the final glass product. Glass samples of seven projected DWPF compositions were cooled following the same temperature profile as that of glass at the centerline of the full-scale DWPF canister. The glasses were characterized by x-ray diffraction and scanning electron microscopy to identify the crystalline phases present The volume percents of each crystalline phase present were determined by quantitative x-ray diffraction. The Product Consistency Test (PCI) was used to determine the durability of the heat-treated glasses

  19. Leaching of glass

    International Nuclear Information System (INIS)

    Hench, L.L.

    1977-01-01

    Understanding surface compositional profiles of glasses over a range of 0-2000 A with a variety of analytical instruments shows that five general types of glass surfaces exist. The surface character of a glass article depends upon bulk composition and environmental history during which surface dealkalization, film formation, and network dissolution can occur. Environmental-surface interactions generally result in complex compositional profiles of all the constituents in a glass. Durable glasses almost always develop a stable surface film which has a higher concentration of network formers than the bulk composition. Compositional effects that are used to improve glass durability usually improve the stability of the surface films. Durability tests or service conditions that lead to film destruction are especially severe for the most silicate glasses. 43 references

  20. MIIT: International in-situ testing of nuclear-waste glasses: Performance of SRS simulated waste glass after five years of burial at the Waste Isolation Pilot Plant (WIPP)

    International Nuclear Information System (INIS)

    Wicks, G.G.; Lodding, A.R.; Macedo, P.B.; Clark, D.E.

    1991-01-01

    In July of 1986, the first in-situ test involving burial of simulated high-level waste (HLW) forms conducted in the United States was started. This program, called the Materials Interface Interactions Test or MIIT, comprises the largest, most cooperative field-testing venture in the international waste management community. In July of 1991, the experimental portion of the 5-year MIIT study was completed on schedule. During this time interval, many in-situ measurements were performed, thousands of brine analyses conducted, and hundreds of waste glass and package components exhumed and evaluated after 6 mo., 1 yr., 2 yr. and 5 yr. burial periods. Although analyses are still in progress, the performance of SRS waste glass based on all data currently available has been seen to be excellent thus far. Initial analyses and assessment of Savannah River (SR) waste glass after burial in WIPP at 90 degrees C for 5 years are presented in this document

  1. Property/composition relationships for Hanford high-level waste glasses melting at 115 degrees C volume 1: Chapters 1-11

    International Nuclear Information System (INIS)

    Hrma, P.R.; Piepel, G.F.

    1994-12-01

    A Composition Variation study (CVS) is being performed within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) project in support of a future high-level nuclear waste vitrification plant at the Hanford site in Washington. From 1989 to 1994, over 120 nonradioactive glasses were melted and properties measured in five statistically-designed experimental phases. Glass composition is represented by the 10 components SiO 2 , B 2 O 3 , Al 2 O 3 , Fe 2 O 3 , ZrO 2 , Na 2 O, Li 2 O, CaO, MgO, and Others (all remaining components). The properties measured include viscosity (η), electrical conductivity (ε), glass transition temperature (T g ), thermal expansion of solid glass (α s ) and molten glass (α m ), crystallinity (quenched and canister centerline cooled glasses), liquidus temperature (T L ), durability based on normalized elemental releases from the Materials Characterization Center-1 28-day dissolution test (MCC-1, r mi ) and the 7-day Product Consistency Test (PCT, r pi ), and solution pHs from MCC-1 and PCT. Amorphous phase separation was also evaluated. Empirical first- and second-order mixture models were fit using the CVS data to relate the various properties to glass composition. Equations for calculating the uncertainty associated with property values predicted by the models were also developed. The models were validated using both internal and external data. Other modeling approaches (e.g., non-bridging oxygen, free energy of hydration, phase-equilibria T L ) were investigated for specific properties. A preliminary Qualified Composition Region was developed to identify glass compositions with high confidence of being processable in a melter and meeting waste form acceptance criteria

  2. Immobilization of radioactive wastes in glasses and ceramics

    International Nuclear Information System (INIS)

    Zanotto, E.D.

    1983-01-01

    A large amount of radioactive liquid wastes arises from the reprocessing of spent nuclear fuels to recover uranium and plutonium. Immobilization of such wastes in solid form and disposal of the solidified wastes in safe places, to prevent contamination of the human environment, are topics of considerable interest for both the scientific community and the public in general. The great majority of materials candidate for the encapsulation of radioactive wastes are inorganic non-metalic, such as glasses, glass-ceramics, special cements, calcined ceramics and few more. Among these materials, certain glasses have received special attention, and are being studied for over twenty years. It is estimated that about US$2 billion have already been spent in these studies. The disposal (long term storage) of these solid wastes may be possible in deep geological formations, salt mines, the ocean bed, by evacuation to the outer space, etc. A brief review on the several options avaiable for encapsulation and disposal of high level radioactive liquid wastes is presented, along with the relative merits and disadvantages of the candidate materials for encapsulation. A few suggestions for the solution of the Brazilian problem are advanced. (Author) [pt

  3. Chemical states of molybdenum in radioactive waste glass

    International Nuclear Information System (INIS)

    Ishiguro, Katsuhiko; Kawanishi, Nobuo; Nagaki, Hiroshi; Naito, Aritsune

    1982-01-01

    In order to confirm an expectation that the chemical state of molybdenum in glass reflects the phase separation tendency of the yellow solid from the melt of borosilicate glass, simulated waste glasses were prepared, and ESCA analysis was performed using a commercially available electron spectrometer (PHI550 E) with an excitation source consisting of Mg Kα-ray. The effects of the concentration of Mo and FE 2 O 3 and the melting atmosphere (oxidizing or reducing) in which the samples were prepared on the chemical state of Mo and the solubility of MoO 3 were examined. From the observation of Mo spectra, it was shown that Mo in waste glass had several valencies, e.g., Mo(3), Mo(4), Mo(5) and Mo(6), while Mo in the yellow solid separated from the melts exhibited hexa-valent state, the peak intensity of higher valencies increased relatively with the increase of MoO 3 concentration, but the chemical state of Mo did not change remarkably around the solubility limit of MoO 3 , the melting atmosphere influenced on the Mo state in the waste glass, the peak intensity of Mo(6) increased relatively with the increasing Fe 2 O 3 concentration, and Mo in devitrified glass exhibited hexa-valent state. (Yoshitake, I.)

  4. Experimental Investigation on the Durability of Glass Fiber-Reinforced Polymer Composites Containing Nanocomposite

    Directory of Open Access Journals (Sweden)

    Weiwen Li

    2013-01-01

    Full Text Available Nanoclay layers incorporated into polymer/clay nanocomposites can inhibit the harmful penetration of water and chemicals into the material, and thus the durability of glass fiber-reinforced polymer (GFRP composites should be enhanced by using polymer/clay nanocomposite as the matrix material. In this study, 1.5 wt% vinyl ester (VE/organoclay and 2 wt% epoxy (EP/organoclay nanocomposites were prepared by an in situ polymerization method. The dispersion states of clay in the nanocomposites were studied by performing XRD analysis. GFRP composites were then fabricated with the prepared 1.5 wt% VE/clay and 2.0 wt% EP/clay nanocomposites to investigate the effects of a nanocomposite matrix on the durability of GFRP composites. The durability of the two kinds of GFRP composites was characterized by monitoring tensile properties following degradation of GFRP specimens aged in water and alkaline solution at 60°C, and SEM was employed to study fracture behaviors of aged GFRP composites under tension. The results show that tensile properties of the two types of GFRP composites with and without clay degrade significantly with aging time. However, the GFRP composites with nanoclay show a lower degradation rate compared with those without nanoclay, supporting the aforementioned hypothesis. And the modification of EP/GFRP enhanced the durability more effectively.

  5. Waste glass as eco-friendly replacement material in construction products

    Science.gov (United States)

    Sharma, Gayatri; Sharma, Anu

    2018-05-01

    Atpresent time the biggest issue is increasing urban population, industrialization and development all over the world. The quantity of the raw materials of construction products like cement, concrete etc is gradually depleting. This is important because if we don't find the alternative material to accomplish need of this industry, with every year it will put pressure on natural resources which are limited in quantity. This major issue can be solved by partial replacing with waste glass of different construction products. This paper gives an overview of the current growth and recycling situation of waste glass and point out the direction for the proper use of waste glass as replacement of construction material. These will not only help in the reuse of waste glass but also create eco-friendly environment.

  6. Predictive modeling of crystal accumulation in high-level waste glass melters processing radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Matyáš, Josef; Gervasio, Vivianaluxa; Sannoh, Sulaiman E.; Kruger, Albert A.

    2017-11-01

    The effectiveness of HLW vitrification is limited by precipitation/accumulation of spinel crystals [(Fe, Ni, Mn, Zn)(Fe, Cr)2O4] in the glass discharge riser of Joule-heated ceramic melters during idling. These crystals do not affect glass durability; however, if accumulated in thick layer, they can clog the melter and prevent discharge of molten glass into canisters. To address this problem, an empirical model was developed that can predict thicknesses of accumulated layers as a function of glass composition. This model predicts well the accumulation of single crystals and/or small-scale agglomerates, but, excessive agglomeration observed in high-Ni-Fe glass resulted in an under-prediction of accumulated layers, which gradually worsen over time as an increased number of agglomerates formed. Accumulation rate of ~53.8 ± 3.7 µm/h determined for this glass will result in ~26 mm thick layer in 20 days of melter idling.

  7. Durable concrete for a waste repository - Measurement of ionic ingress

    International Nuclear Information System (INIS)

    Feldman, R.F.; Beaudoin, J.J.; Philipose, K.E.

    1990-01-01

    A waste repository for the below ground disposal of low level radioactive waste is planned at Chalk River Nuclear Laboratories. It relies greatly on the durability of concrete for the required 500 year service life. A research program to design durable concrete and predict its service life is in progress. The degradation of the concrete depends to a large extent on the rate of ingress of corrosive agents. Penetration of chloride and sulfate ions are particularly relevant. Twenty mix formulations were developed to create various types and qualities of concrete, and to study their behavior in different site environmental conditions. A total of 1,000 concrete specimens are being exposed at 20C and 45C to 25 different combinations of the corrosive agents including CO 2 . Procedures to measure the ionic profiles and to determine the factors controlling diffusion of the ions in the various concretes have been developed. Results of selected concrete systems exposed to chloride and sulfate solutions for 1 year are presented and discussed in terms of pore structure and permeability parameters of the concrete

  8. Method of immobilizing weapons plutonium to provide a durable, disposable waste product

    Science.gov (United States)

    Ewing, Rodney C.; Lutze, Werner; Weber, William J.

    1996-01-01

    A method of atomic scale fixation and immobilization of plutonium to provide a durable waste product. Plutonium is provided in the form of either PuO.sub.2 or Pu(NO.sub.3).sub.4 and is mixed with and SiO.sub.2. The resulting mixture is cold pressed and then heated under pressure to form (Zr,Pu)SiO.sub.4 as the waste product.

  9. Glass corrosion in natural environments

    Science.gov (United States)

    Thorpe, Arthur N.; Barkatt, Aaron

    1992-01-01

    Experiments carried out during the progress period are summarized. Experiments carried out involving glass samples exposed to solutions of Tris have shown the appearance of 'spikes' upon monitoring glass dissolution as a function of time. The periodic 'spikes' observed in Tris-based media were interpreted in terms of cracking due to excessive stress in the surface region of the glass. Studies of the interactions of silicate glasses with metal ions in buffered media were extended to systems containing Al. Caps buffer was used to establish the pH. The procedures used are described and the results are given. Preliminary studies were initiated as to the feasibility of adding a slowly dissolving solid compound of the additive to the glass-water system to maintain a supply of dissolved additive. It appears that several magnesium compounds have a suitable combination of solubility and affinity towards silicate glass surfaces to have a pronounced retarding effect on the extraction of uranium from the glass. These preliminary findings raise the possibility that introducing a magnesium source into geologic repositories for nuclear waste glass in the form of a sparingly soluble Mg-based backfill material may cause a substantial reduction in the extent of long-term glass corrosion. The studies described also provide mechanistic understanding of the roles of various metal solutes in the leachant. Such understanding forms the basis for developing long-term predictions of nuclear waste glass durability under repository conditions. From what is known about natural highly reduced glasses such as tektites, it is clear that iron is dissolved as ferrous iron with little or no ferric iron. The reducing conditions were high enough to cause metallic iron to exsolve out of the glass in the form of submicroscopic spherules. As the nuclear waste glass is much less reduced, a study was initiated on other natural glasses in addition to the nuclear waste glass. Extensive measurements were

  10. Comparative assessment of TRU waste forms and processes. Volume I. Waste form and process evaluations

    International Nuclear Information System (INIS)

    Ross, W.A.; Lokken, R.O.; May, R.P.; Roberts, F.P.; Timmerman, C.L.; Treat, R.L.; Westsik, J.H. Jr.

    1982-09-01

    This study provides an assesses seven waste forms and eight processes for immobilizing transuranic (TRU) wastes. The waste forms considered are cast cement, cold-pressed cement, FUETAP (formed under elevated temperature and pressure) cement, borosilicate glass, aluminosilicate glass, basalt glass-ceramic, and cold-pressed and sintered silicate ceramic. The waste-immobilization processes considered are in-can glass melting, joule-heated glass melting, glass marble forming, cement casting, cement cold-pressing, FUETAP cement processing, ceramic cold-pressing and sintering, basalt glass-ceramic processing. Properties considered included gas generation, chemical durability, mechanical strength, thermal stability, and radiation stability. The ceramic products demonstrated the best properties, except for plutonium release during leaching. The glass and ceramic products had similar properties. The cement products generally had poorer properties than the other forms, except for plutonium release during leaching. Calculations of the Pu release indicated that the waste forms met the proposed NRC release rate limit of 1 part in 10 5 per year in most test conditions. The cast-cement process had the lowest processing cost, followed closely by the cold-pressed and FUETAP cement processes. Joule-heated glass melting had the lower cost of the glass processes. In-can melting in a high-quality canister had the highest cost, and cold-pressed and sintered ceramic the second highest. Labor and canister costs for in-can melting were identified. The major contributor to costs of disposing of TRU wastes in a defense waste repository is waste processing costs. Repository costs could become the dominant cost for disposing of TRU wastes in a commercial repository. It is recommended that cast and FUETAP cement and borosilicate glass waste-form systems be considered. 13 figures, 16 tables

  11. Performance Modeling Applied to the Treatment and Disposal of a Mixed Waste at the SRS

    International Nuclear Information System (INIS)

    Pickett, J.B.; Jantzen, C.M.; Cook, J.R.; Whited, A.R.; Field, R.A.

    1997-05-01

    Performance modeling for Low Level Mixed Waste disposal was conducted using the measured leach rates from a number of vitrified waste formulations. The objective of the study was to determine if the improved durability of a vitrified mixed waste would allow trench disposal at the Savannah River Site (SRS). Leaching data were compiled from twenty-nine diverse reference glasses, encompassing a wide range of exposed glass surface area to leachant volume ratios (SA/V), and various leachant solutions; all of which had been leached at 90 degrees Celsius, using the MCC-1 or PCT procedures (ASTM Procedures C1220-92 and C1285-94, respectively). The normalized leach rates were scaled to the ambient disposal temperature of 25 degrees Celsius, and compared to the allowable leach rate of uranium - which would meet the performance assessment requirements. The results indicated that a glass of above average durability (vs. the reference glasses) would meet the uranium leaching concentration for direct SRS trench disposal

  12. Molecular environmental science using synchrotron radiation: Chemistry and physics of waste form materials

    International Nuclear Information System (INIS)

    Lindle, Dennis W.; Shuh, David K.

    2005-01-01

    Production of defense-related nuclear materials has generated large volumes of complex chemical wastes containing a mixture of radionuclides. The disposition of these wastes requires conversion of the liquid and solid-phase components into durable, solid forms suitable for long-term immobilization [1]. Specially formulated glass compositions, many of which have been derived from glass developed for commercial purposes, and ceramics such as pyrochlores and apatites, will be the main recipients for these wastes. The performance characteristics of waste-form glasses and ceramics are largely determined by the loading capacity for the waste constituents (radioactive and non-radioactive) and the resultant chemical and radiation resistance of the waste-form package to leaching (durability). There are unique opportunities for the use of near-edge soft-x-ray absorption fine structure (NEXAFS) spectroscopy to investigate speciation of low-Z elements forming the backbone of waste-form glasses and ceramics. Although nuclear magnetic resonance (NMR) is the primary technique employed to obtain speciation information from low-Z elements in waste forms, NMR is incompatible with the metallic impurities contained in real waste and is thus limited to studies of idealized model systems. In contrast, NEXAFS can yield element-specific speciation information from glass constituents without sensitivity to paramagnetic species. Development and use of NEXAFS for eventual studies of real waste glasses has significant implications, especially for the low-Z elements comprising glass matrices [5-7]. The NEXAFS measurements were performed at Beamline 6.3.1, an entrance-slitless bend-magnet beamline operating from 200 eV to 2000 eV with a Hettrick-Underwood varied-line-space (VLS) grating monochromator, of the Advanced Light Source (ALS) at LBNL. Complete characterization and optimization of this beamline was conducted to enable high-performance measurements

  13. Molecular environmental science using synchrotron radiation:Chemistry and physics of waste form materials

    Energy Technology Data Exchange (ETDEWEB)

    Lindle, Dennis W.; Shuh, David K.

    2005-02-28

    Production of defense-related nuclear materials has generated large volumes of complex chemical wastes containing a mixture of radionuclides. The disposition of these wastes requires conversion of the liquid and solid-phase components into durable, solid forms suitable for long-term immobilization [1]. Specially formulated glass compositions, many of which have been derived from glass developed for commercial purposes, and ceramics such as pyrochlores and apatites, will be the main recipients for these wastes. The performance characteristics of waste-form glasses and ceramics are largely determined by the loading capacity for the waste constituents (radioactive and non-radioactive) and the resultant chemical and radiation resistance of the waste-form package to leaching (durability). There are unique opportunities for the use of near-edge soft-x-ray absorption fine structure (NEXAFS) spectroscopy to investigate speciation of low-Z elements forming the backbone of waste-form glasses and ceramics. Although nuclear magnetic resonance (NMR) is the primary technique employed to obtain speciation information from low-Z elements in waste forms, NMR is incompatible with the metallic impurities contained in real waste and is thus limited to studies of idealized model systems. In contrast, NEXAFS can yield element-specific speciation information from glass constituents without sensitivity to paramagnetic species. Development and use of NEXAFS for eventual studies of real waste glasses has significant implications, especially for the low-Z elements comprising glass matrices [5-7]. The NEXAFS measurements were performed at Beamline 6.3.1, an entrance-slitless bend-magnet beamline operating from 200 eV to 2000 eV with a Hettrick-Underwood varied-line-space (VLS) grating monochromator, of the Advanced Light Source (ALS) at LBNL. Complete characterization and optimization of this beamline was conducted to enable high-performance measurements.

  14. Immobilization of Technetium Waste from Pyro-processing Using Tellurite Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Jong; Pyo, Jae-Young; Lee, Cheong-Won [POSTECH, Pohang (Korea, Republic of); Yang, Jae-Hwan; Park, Hwan-Seo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Vitrification of Tc wastes has been challenging because of the low solubility in the silicate glass and high volatility in the melting process. In previous studies, the measured solubility of Tc and Re was ⁓ 3000 ppm at 1000 .deg. C in low activity waste (LAW) glass. And retention of Tc has been reported within 12 - 77% during the borosilicate vitrification process. Tellurite glasses have been studied for halide waste immobilization due to low melting temperatures (Tm= 600-800 .deg. C) and flexibility of network with foreign ions. Tellurite glasses offered higher halide retention than borosilicate glasses. The structure of pure tellurite (TeO{sub 2}) consists of TeO{sub 4} trigonal bipyramids (tbp), but TeO{sub 4} units are converted to TeO{sub 3} trigonal pyramids (tp) having non-bridging oxygen (NBO) as the modifiers added. Objectives of this study are to investigate the tellurite glasses for Tc immobilization using Re as a surrogate. Retention and waste loading of Re were analyzed during the vitrification process of tellurite glass. We investigated local structures of Re ions in glasses by Raman and X-ray absorption spectroscopies. The tellurite glass was investigated to immobilize the Ca(TcO{sub 4}){sub 2}, surrogated by Ca(ReO{sub 4}){sub 2}. The average of Re retention in tellurite glass was 86%. The 7-day PCT results were satisfied with U.S requirement up to 9 mass% of Ca(ReO{sub 4}){sub 2} content. Re in the tellurite glass exists +7 oxidation state and was coordinated with 4 oxygen.

  15. Corrosion of inconel in high-temperature borosilicate glass melts containing simulant nuclear waste

    Science.gov (United States)

    Mao, Xianhe; Yuan, Xiaoning; Brigden, Clive T.; Tao, Jun; Hyatt, Neil C.; Miekina, Michal

    2017-10-01

    The corrosion behaviors of Inconel 601 in the borosilicate glass (MW glass) containing 25 wt.% of simulant Magnox waste, and in ZnO, Mn2O3 and Fe2O3 modified Mg/Ca borosilicate glasses (MZMF and CZMF glasses) containing 15 wt.% of simulant POCO waste, were evaluated by dimensional changes, the formation of internal defects and changes in alloy composition near corrosion surfaces. In all three kinds of glass melts, Cr at the inconel surface forms a protective Cr2O3 scale between the metal surface and the glass, and alumina precipitates penetrate from the metal surface or formed in-situ. The corrosion depths of inconel 601 in MW waste glass melt are greater than those in the other two glass melts. In MW glass, the Cr2O3 layer between inconel and glass is fragmented because of the reaction between MgO and Cr2O3, which forms the crystal phase MgCr2O4. In MZMF and CZMF waste glasses the layers are continuous and a thin (Zn, Fe, Ni, B)-containing layer forms on the surface of the chromium oxide layer and prevents Cr2O3 from reacting with MgO or other constituents. MgCr2O4 was observed in the XRD analysis of the bulk MW waste glass after the corrosion test, and ZrSiO4 in the MZMF waste glass, and ZrSiO4 and CaMoO4 in the CZMF waste glass.

  16. Glass compositions suitable for PFR wastes

    International Nuclear Information System (INIS)

    Boult, K.A.; Dalton, J.T.; Eccles, E.W.; Hough, A.; Marples, J.A.C.; Paige, E.L.; Sutcliffe, P.W.

    1988-03-01

    Previous work had identified glass compositions that were suitable for vitrifying current and future high level wastes from the Prototype Fast Reactor (PFR) fuel reprocessing plant. Further work on these glasses has shown that: a) Foaming and crystallisation can occur under certain conditions, both probably associated with the presence of iron in the waste. Either of these could lead to greater difficulties in processing. b) Inconel 690, the preferred JCM (Joule-heated Ceramic Melter) electrode material has an acceptable corrosion rate at 1200 0 C: ca 0.6mm.y -1 . c) The leach rates are unaffected by radiation damage. The density of the glass decreases slightly with α-dose, with a dependency that extrapolates, at infinite time, to an 0.13% linear expansion. d) The concentrations of the radiologically important elements Tc, Np, Pu and Am, observed in a 'repository simulation' leach test, were satisfactorily low. (author)

  17. Methods of vitrifying waste with low melting high lithia glass compositions

    Science.gov (United States)

    Jantzen, Carol M.; Pickett, John B.; Cicero-Herman, Connie A.; Marra, James C.

    2001-01-01

    The invention relates to methods of vitrifying waste and for lowering the melting point of glass forming systems by including lithia formers in the glass forming composition in significant amounts, typically from about 0.16 wt % to about 11 wt %, based on the total glass forming oxides. The lithia is typically included as a replacement for alkali oxide glass formers that would normally be present in a particular glass forming system. Replacement can occur on a mole percent or weight percent basis, and typically results in a composition wherein lithia forms about 10 wt % to about 100 wt % of the alkali oxide glass formers present in the composition. The present invention also relates to the high lithia glass compositions formed by these methods. The invention is useful for stabilization of numerous types of waste materials, including aqueous waste streams, sludge solids, mixtures of aqueous supernate and sludge solids, combinations of spent filter aids from waste water treatment and waste sludges, supernate alone, incinerator ash, incinerator offgas blowdown, or combinations thereof, geological mine tailings and sludges, asbestos, inorganic filter media, cement waste forms in need of remediation, spent or partially spent ion exchange resins or zeolites, contaminated soils, lead paint, etc. The decrease in melting point achieved by the present invention desirably prevents volatilization of hazardous or radioactive species during vitrification.

  18. Superconducting Open-Gradient Magnetic Separation for the Pretreatment of Radioactive or Mixed Waste Vitrification Feeds

    International Nuclear Information System (INIS)

    Nunez', L.; Kaminsky', M.D.; Crawford, C.; Ritter, J.A.

    1999-01-01

    An open-gradient magnetic separation (OGMS) process is being considered to separate deleterious elements from radioactive and mixed waste streams prior to vitrification or stabilization. By physically segregating solid wastes and slurries based on the magnetic properties of the solid constituents, this potentially low-cost process may serve the U.S. Department of Energy (DOE) by reducing the large quantities of glass produced from defense-related high-level waste (HLW). Furthermore, the separation of deleterious elements from low-level waste (LLW) also can reduce the total quantity of waste produced in LLW immobilization activities. Many HLW 'and LLW waste' streams at both Hanford and the Savannah River Site (SRS) include constituents deleterious to the durability of borosilicate glass and the melter many of the constituents also possess paramagnetism. For example, Fe, Cr, Ni, and other transition metals may limit the waste loading and affect the durability of the glass by forming spine1 phases at the high operating temperature used in vitrification. Some magnetic spine1 phases observed in glass formation are magnetite (Fe,O,), chromite (FeCrO,), and others [(Fe, Ni, Mg, Zn, Mn)(Al, Fe, Ti, Cr)O,] as described elsewhere [Bates-1994, Wronkiewicz-1994] Stable spine1 phases can cause segregation between the glass and the crystalline phases. As a consequence of the difference in density, the spine1 phases tend to accumulate at the bottom of the glass melter, which decreases the conductivity and melter lifetime [Sproull-1993]. Crystallization also can affect glass durability [Jantzen-1985, Turcotte- 1979, Buechele-1990] by changing the chemical composition of the matrix glass surrounding the crystals or causing stress at the glass/crystal interface. These are some of the effects that can increase leaching [Jantzen-1985]. A SRS glass that was partially crystallized to contain 10% vol. crystals composed of spinels, nepheline, and acmite phases showed minimal changes in

  19. Glass-solidification method for high level radioactive liquid waste

    International Nuclear Information System (INIS)

    Kawamura, Kazuhiro; Kometani, Masayuki; Sasage, Ken-ichi.

    1996-01-01

    High level liquid wastes are removed with precipitates mainly comprising Mo and Zr, thereafter, the high level liquid wastes are mixed with a glass raw material comprising a composition having a B 2 O 3 /SiO 2 ratio of not less than 0.41, a ZnO/Li 2 O ratio of not less than 1.00, and an Al 2 O 3 /Li 2 O ratio of not less than 2.58, and they are melted and solidified into glass-solidification products. The liquid waste content in the glass-solidification products can be increased up to about 45% by using the glass raw material having such a predetermined composition. In addition, deposition of a yellow phase does not occur, and a leaching rate identical with that in a conventional case can be maintained. (T.M.)

  20. High-level waste borosilicate glass a compendium of corrosion characteristics. Volume 1

    International Nuclear Information System (INIS)

    Cunnane, J.C.

    1994-03-01

    Current plans call for the United States Department of Energy (DOE) to start up facilities for vitrification of high-level radioactive waste (HLW) stored in tanks at the Savannah River Site, Aiken, South Carolina, in 1995; West Valley Demonstration Project, West Valley, New York, in 1996; and at the Hanford Site, Richland, Washington, after the year 2000. The product from these facilities will be canistered HLW borosilicate glass, which will be stored, transported, and eventually disposed of in a geologic repository. The behavior of this glass waste product, under the range of likely service conditions, is the subject of considerable scientific and public interest. Over the past few decades, a large body of scientific information on borosilicate waste glass has been generated worldwide. The intent of this document is to consolidate information pertaining to our current understanding of waste glass corrosion behavior and radionuclide release. The objective, scope, and organization of the document are discussed in Section 1.1, and an overview of borosilicate glass corrosion is provided in Section 1.2. The history of glass as a waste form and the international experience with waste glass are summarized in Sections 1.3 and 1.4, respectively

  1. Incorporation of tv tube glass waste in aluminous porcelain

    Energy Technology Data Exchange (ETDEWEB)

    Holanda, J.N.F.; Santos, T.F.; Paes Junior, H.R. [Universidade Estadual do Norte Fluminense (UENF), Campos dos Goytacazes, RJ (Brazil)

    2016-07-01

    Full test: This work analyzes the reuse of TV tube glass waste as a method to provide alternative raw material for aluminous porcelain, through of replacement of natural sodic feldspar by up to 30 wt.%. Aluminous porcelain formulations containing TV tube glass waste were pressed and fired in air at 1300 deg C using a fast-firing cycle. Ceramic pieces were characterized by X-ray diffraction, scanning electron microscopy, linear shrinkage, apparent density, apparent porosity, water absorption, and electrical resistivity. XRD and SEM results indicated that all aluminous porcelain pieces are composed essentially of mullite, quartz, and ?-alumina embedded in a vitreous matrix. The results also showed that the aluminous porcelain pieces containing TV tube glass waste presented low water absorption values between 0.42 and 0.45 %, apparent density between 2.44 and 2.46 g/cm3, and volume electrical resistivity between 1.91 and 2.93 x 1011 ?.cm. Thus, the TV tube glass waste could be used into aluminous porcelain formulations, in the range up to 30 wt.%, as a replacement for traditional flux material (sodic feldspar). (author)

  2. High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 2

    International Nuclear Information System (INIS)

    Cunnane, J.C.

    1994-03-01

    The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion.This document is organized into three volumes. Volumes I and II represent a tiered set of information intended for somewhat different audiences. Volume I is intended to provide an overview of waste glass corrosion, and Volume 11 is intended to provide additional experimental details on experimental factors that influence waste glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II. Volume I is intended for managers, decision makers, and modelers, the combined set of Volumes I, II, and III is intended for scientists and engineers working in the field of high-level waste

  3. High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    Cunnane, J.C. [comp.; Bates, J.K.; Bradley, C.R. [Argonne National Lab., IL (United States)] [and others

    1994-03-01

    The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion.This document is organized into three volumes. Volumes I and II represent a tiered set of information intended for somewhat different audiences. Volume I is intended to provide an overview of waste glass corrosion, and Volume 11 is intended to provide additional experimental details on experimental factors that influence waste glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II. Volume I is intended for managers, decision makers, and modelers, the combined set of Volumes I, II, and III is intended for scientists and engineers working in the field of high-level waste.

  4. ANL technical support program for DOE Environmental Restoration and Waste Management. Annual report, October 1991--September 1992

    Energy Technology Data Exchange (ETDEWEB)

    Bates, J.K.; Bradley, C.R.; Buck, E.C.; Cunnane, J.C.; Dietz, N.L.; Ebert, W.L.; Emery, J.W.; Feng, X.; Gerding, T.J.; Gong, M.; Hoh, J.C.; Mazer, J.J.; Wronkiewicz, D.J. [Argonne National Lab., IL (United States); Bourcier, W.L.; Morgan, L.E.; Newton, L.; Nielsen, J.K.; Phillips, B.L. [Lawrence Livermore National Lab., CA (United States); Ewing, R.C.; Wang, L. [Univ. of New Mexico, Albuquerque, NM (United States); Li, H.; Tomozawa, M. [Rensselaer Polytechnic Inst., Troy, NY (United States)

    1993-05-01

    A program was established for DOE Environmental Restoration and Waste Management (EM) to evaluate factors that are anticipated to affect waste glass reaction during repository disposal, especially in an unsaturated environment typical of what may be expected for the proposed Yucca Mountain repository site. This report covers progress in FY 1992 on the following tasks: 1. A compendium of the characteristics of high-level nuclear waste borosilicate glass has been written. 2. A critical review of important parameters that affect the reactivity of glass in an unsaturated environment is being prepared. 3. A series of tests has been started to evaluate the reactivity of fully radioactive glasses in a high-level waste repository environment and compare it to the reactivity of synthetic, nonradioactive glasses of similar composition. 4. The effect of radiation upon the durability of waste glasses at a high glass surface area-to-liquid volume (SA/V) ratio and a high gas-to-liquid volume ratio will be assessed. These tests address both vapor and high SA/V liquid conditions. 5. A series of tests is being performed to compare the extent of reaction of nuclear waste glasses at various SAN ratios. Such differences in the SAN ratio may significantly affect glass durability. 6. A series of natural analogue tests is being analyzed to demonstrate a meaningful relationship between experimental and natural alteration conditions. 7. Analytical electron microscopy (AEM), infrared spectroscopys and nuclear resonant profiling are being used to assess the glass/water reaction pathway by identifying intermediate phases that appear on the reacting glass. Additionally, colloids from the leach solutions are being studied using AEM. 8. A technical review of AEM results is being provided. 9. A study of water diffusion involving nuclear waste glasses is being performed. 10. A mechanistically based model is being developed to predict the performance of glass over repository-relevant time periods.

  5. Chemical durability and degradation mechanisms of HT9 based alloy waste forms with variable Zr content

    Energy Technology Data Exchange (ETDEWEB)

    Olson, L. N. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-10-30

    In Corrosion studies were undertaken on alloy waste forms that can result from advanced electrometallurgical processing techniques to better classify their durability and degradation mechanisms. The waste forms were based on the RAW3-(URe) composition, consisting primarily of HT9 steel and other elemental additions to simulate nuclear fuel reprocessing byproducts. The solution conditions of the corrosion studies were taken from an electrochemical testing protocol, and meant to simulate conditions in a repository. The alloys durability was examined in alkaline and acidic brines.

  6. Method of volume-reducing processing for radioactive wastes

    International Nuclear Information System (INIS)

    Sato, Koei; Yamauchi, Noriyuki; Hirayama, Toshihiko.

    1985-01-01

    Purpose: To process the processing products of radioactive liquid wastes and burnable solid wastes produced from nuclear facilities into stable solidification products by heat melting. Method: At first, glass fiber wastes of contaminated air filters are charged in a melting furnace. Then, waste products obtained through drying, sintering, incineration, etc. are mixed with a proper amount of glass fibers and charged into the melting furnace. Both of the charged components are heated to a temperature at which the glass fibers are melted. The burnable materials are burnt out to provide a highly volume-reduced products. When the products are further heated to a temperature at which metals or metal oxides of a higher melting point than the glass fiber, the glass fibers and the metals or metal oxides are fused to each other to be combined in a molecular structure into more stabilized products. The products are excellent in strength, stability, durability and leaching resistance at ambient temperature. (Kamimura, M.)

  7. UK program: glasses and ceramics for immobilization of radioactive wastes for disposal

    International Nuclear Information System (INIS)

    Johnson, K.D.B.

    1979-01-01

    The UK Research Program on Radioactive Waste Management includes the development of processes for the conversion of high-level-liquid-reprocessing wastes from thermal and fast reactors to borosilicate glasses. The properties of these glasses and their behavior under storage and disposal conditions have been examined. Methods for immobilizing activity from other wastes by conversion to glass or ceramic forms are described. The UK philosophy of final solutions to waste management and disposal is presented

  8. Effects of waste glass additions on quality of textile sludge-based bricks.

    Science.gov (United States)

    Rahman, Ari; Urabe, Takeo; Kishimoto, Naoyuki; Mizuhara, Shinji

    2015-01-01

    This research investigated the utilization of textile sludge as a substitute for clay in brick production. The addition of textile sludge to a brick specimen enhanced its pores, thus reducing the quality of the product. However, the addition of waste glass to brick production materials improved the quality of the brick in terms of both compressive strength and water absorption. Maximum compressive strength was observed with the following composition of waste materials: 30% textile sludge, 60% clay and 10% waste glass. The melting of waste glass clogged up pores on the brick, which improved water absorption performance and compressive strength. Moreover, a leaching test on a sludge-based brick to which 10% waste glass did not detect significant heavy metal compounds in leachates, with the product being in conformance with standard regulations. The recycling of textile sludge for brick production, when combined with waste glass additions, may thus be promising in terms of both product quality and environmental aspects.

  9. Radiation and Thermal Ageing of Nuclear Waste Glass

    Energy Technology Data Exchange (ETDEWEB)

    Weber, William J [ORNL

    2014-01-01

    The radioactive decay of fission products and actinides incorporated into nuclear waste glass leads to self-heating and self-radiation effects that may affect the stability, structure and performance of the glass in a closed system. Short-lived fission products cause significant self-heating for the first 600 years. Alpha decay of the actinides leads to self-radiation damage that can be significant after a few hundred years, and over the long time periods of geologic disposal, the accumulation of helium and radiation damage from alpha decay may lead to swelling, microstructural evolution and changes in mechanical properties. Four decades of research on the behavior of nuclear waste glass are reviewed.

  10. Demonstration of sulfur solubility determinations in high waste loading, low-activity waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-04-25

    A method recommended by Pacific Northwest National Laboratory (PNNL) for sulfate solubility determinations in simulated low-activity waste glasses was demonstrated using three compositions from a recent Hanford high waste loading glass study. Sodium and sulfate concentrations in the glasses increased after each re-melting step. Visual observations of the glasses during the re-melting process reflected the changes in composition. The measured compositions showed that the glasses met the targeted values. The amount of SO3 retained in the glasses after washing was relatively high, ranging from 1.6 to 2.6 weight percent (wt %). Measured SnO2 concentrations were notably low in all of the study glasses. The composition of the wash solutions should be measured in future work to determine whether SnO2 is present with the excess sulfate washed from the glass. Increases in batch size and the amount of sodium sulfate added did not have a measureable impact on the amount of sulfate retained in the glass, although this was tested for only a single glass composition. A batch size of 250 g and a sodium sulfate addition targeting 7 wt %, as recommended by PNNL, will be used in future experiments.

  11. Development and characterization of basalt-glass ceramics for the immobilization of transuranic wastes

    International Nuclear Information System (INIS)

    Lokken, R.O.; Chick, L.A.; Thomas, L.E.

    1982-09-01

    Basalt-based waste forms were developed for the immobilization of transuranic (TRU) contaminated wastes. The specific waste studied is a 3:1 blend of process sludge and incinerator ash. Various amounts of TRU blended waste were melted with Pomona basalt powder. The vitreous products were subjected to a variety of heat treatment conditions to form glass ceramics. The total crystallinity of the glass ceramic, ranging from 20 to 45 wt %, was moderately dependent on composition and heat treatment conditions. Three parent glasses and four glass ceramics with varied composition and heat treatment were produced for detailed phase characterization and leaching. Both parent glasses and glass ceramics were mainly composed of a continuous, glassy matrix phase. This glass matrix entered into solution during leaching in both types of materials. The Fe-Ti rich dispersed glass phase was not significantly degraded by leaching. The glass ceramics, however, exhibited four to ten times less elemental releases during leaching than the parent glasses. The glass ceramic matrix probably contains higher Fe and Na and lower Ca and Mg relative to the parent glass matrix. The crystallization of augite in the glass ceramics is believed to contribute to the improved leach rates. Leach rates of the basalt glass ceramic are compared to those of other TRU nuclear waste forms containing 239 Pu

  12. DWPF waste glass Product Composition Control System

    International Nuclear Information System (INIS)

    Brown, K.G.; Postles, R.L.

    1992-01-01

    The Defense Waste Processing Facility (DWPF) will be used to blend aqueous radwaste (PHA) with solid radwaste (Sludge) in a waste receipt vessel (the SRAT). The resulting SRAT material is transferred to the SME an there blended with ground glass (Frit) to produce a batch of melter feed slurry. The SME material is passed to a hold tank (the MFT) which is used to continuously feed the DWPF melter. The melter. The melter produces a molten glass wasteform which is poured into stainless steel canisters for cooling and, ultimately, shipment to and storage in a geologic repository. The Product Composition Control System (PCCS) is the system intended to ensure that the melt will be processible and that the glass wasteform will be acceptable. This document provides a description of this system

  13. Performance Characteristics of Waste Glass Powder Substituting Portland Cement in Mortar Mixtures

    Science.gov (United States)

    Kara, P.; Csetényi, L. J.; Borosnyói, A.

    2016-04-01

    In the present work, soda-lime glass cullet (flint, amber, green) and special glass cullet (soda-alkaline earth-silicate glass coming from low pressure mercury-discharge lamp cullet and incandescent light bulb borosilicate glass waste cullet) were ground into fine powders in a laboratory planetary ball mill for 30 minutes. CEM I 42.5N Portland cement was applied in mortar mixtures, substituted with waste glass powder at levels of 20% and 30%. Characterisation and testing of waste glass powders included fineness by laser diffraction particle size analysis, specific surface area by nitrogen adsorption technique, particle density by pycnometry and chemical analysis by X-ray fluorescence spectrophotometry. Compressive strength, early age shrinkage cracking and drying shrinkage tests, heat of hydration of mortars, temperature of hydration, X-ray diffraction analysis and volume stability tests were performed to observe the influence of waste glass powder substitution for Portland cement on physical and engineering properties of mortar mixtures.

  14. High-Strength / High Alkaline Resistant Fe-Phosphate Glass Fibers as Concrete Reinforcement

    Energy Technology Data Exchange (ETDEWEB)

    Mariano Velez

    2008-03-31

    Calcium-iron-phosphate glasses were developed whose chemical durabilities in alkaline solutions (pH 13) were comparable or superior to those of commercial alkaline-resistant (AR) silica-based glasses. However, the tensile strength of Ca-Fe-phosphate fibers, after being exposed to alkaline environments, including wet Portland cement pastes, is lower than that of current AR silicate fibers. Another series of Ca-Fe-phosphate glasses were developed with excellent chemical durability in strong acidic solutions (H2SO4, HF), indicating potential applications where silica-based fibers degrade very quickly, including E-glass. The new Ca-Fe-phosphate glasses can be melted and processed 300 to 500°C lower than silica-based glasses. This offers the possibility of manufacturing glass fibers with lower energy costs by 40-60% and the potential to reduce manufacturing waste and lower gas emissions. It was found that Ca-Fe-phosphate melts can be continuously pulled into fibers depending on the slope of the viscosity-temperature curve and with viscosity ~100 poise, using multi-hole Pt/Rh bushings.

  15. Waste glass corrosion modeling: Comparison with experimental results

    International Nuclear Information System (INIS)

    Bourcier, W.L.

    1993-11-01

    A chemical model of glass corrosion will be used to predict the rates of release of radionuclides from borosilicate glass waste forms in high-level waste repositories. The model will be used both to calculate the rate of degradation of the glass, and also to predict the effects of chemical interactions between the glass and repository materials such as spent fuel, canister and container materials, backfill, cements, grouts, and others. Coupling between the degradation processes affecting all these materials is expected. Models for borosilicate glass dissolution must account for the processes of (1) kinetically-controlled network dissolution, (2) precipitation of secondary phases, (3) ion exchange, (4) rate-limiting diffusive transport of silica through a hydrous surface reaction layer, and (5) specific glass surface interactions with dissolved cations and anions. Current long-term corrosion models for borosilicate glass employ a rate equation consistent with transition state theory embodied in a geochemical reaction-path modeling program that calculates aqueous phase speciation and mineral precipitation/dissolution. These models are currently under development. Future experimental and modeling work to better quantify the rate-controlling processes and validate these models are necessary before the models can be used in repository performance assessment calculations

  16. Letter report: Minor component study for low-level radioactive waste glasses

    International Nuclear Information System (INIS)

    Li, H.

    1996-03-01

    During the waste vitrification process, troublesome minor components in low-level radioactive waste streams could adversely affect either waste vitrification rate or melter life-time. Knowing the solubility limits for these minor components is important to determine pretreatment options for waste streams and glass formulation to prevent or to minimize these problems during the waste vitrification. A joint study between Pacific Northwest Laboratory and Rensselaer Polytechnic Institute has been conducted to determine minor component impacts in low-level nuclear waste glass

  17. The quality study of recycled glass phosphor waste for LED

    Science.gov (United States)

    Tsai, Chun-Chin; Chen, Guan-Hao; Yue, Cheng-Feng; Chen, Cin-Fu; Cheng, Wood-Hi

    2017-02-01

    To study the feasibility and quality of recycled glass phosphor waste for LED packaging, the experiments were conducted to compare optical characteristics between fresh color conversion layer and that made of recycled waste. The fresh color conversion layer was fabricated through sintering pristine mixture of Y.A.G. powder [yellow phosphor (Y3AlO12 : Ce3+). Those recycled waste glass phosphor re-melted to form Secondary Molten Glass Phosphor (S.M.G.P.). The experiments on such low melting temperature glass results showed that transmission rates of S.M.G.P. are 9% higher than those of first-sintered glass phosphor, corresponding to 1.25% greater average bubble size and 36% more bubble coverage area in S.M.G.P. In the recent years, high power LED modules and laser projectors have been requiring higher thermal stability by using glass phosphor materials for light mixing. Nevertheless, phosphor and related materials are too expensive to expand their markets. It seems a right trend and research goal that recycling such waste of high thermal stability and quality materials could be preferably one of feasible cost-down solutions. This technical approach could bring out brighter future for solid lighting and light source module industries.

  18. Fostering a durable relationship between a waste management facility and its host community

    International Nuclear Information System (INIS)

    2007-01-01

    Any long-term radioactive waste management project is likely to last decades to centuries. It requires a physical site and will impact in a variety of ways on the surrounding community over that whole period. The societal durability of an agreed solution is essential to success. This report identifies a number of design elements (including functional, cultural and physical features) that favour a durable relationship between the facility and its host community by improving prospects for quality of life across generations

  19. Lead-iron phosphate glass: a stable storage medium for high-level nuclear waste

    International Nuclear Information System (INIS)

    Sales, B.C.; Boatner, L.A.

    1984-01-01

    Results are presented which show that lead-iron phosphate glasses are a promising new waste form for the safe immobilization of both high-level defense and high-level commercial radioactive waste. Relative to the borosilicate nuclear waste glasses that are currently the ''reference'' waste form for the long-term disposal of nuclear waste, lead-iron phosphate glasses have several distinct advantages: (1) an aqueous corrosion rate that is about 1000 times lower, (2) a processing temperature that is 100 0 to 250 0 C lower and, (3) a much lower melt viscosity in the temperature range from 800 0 to 1000 0 C. Most significantly, the lead-iron phosphate waste form can be processed using a technology similar to that developed for borosilicate nuclear waste glasses

  20. Studies of glass waste form performance at Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    Banba, Tsunetaka; Kamizono, Hiroshi; Nakayama, Shinichi; Tashiro, Shingo

    1989-08-01

    The recent studies of glass waste form performance at Japan Atomic Energy Research Institute can be classified into the following three categories; (1) Study on the volatilization of radionuclides from the waste glass, which is necessary to estimate the safety in relation to operation of a storage facility. (2) Study on the radiation (alpha-radiation) effects which have relation to the long-term stability of the waste glass. (3) Study on the leaching behavior of actinides under the repository conditions, which is necessary to predict the long-term release rate of radionuclides from the waste glass. In the present report, the recent results corresponding to the above categories are described. (author)

  1. Characterization of damage created by alpha disintegrations in radionuclear waste glass

    International Nuclear Information System (INIS)

    Jacquet-Francillon, N.; Mueller, P.

    1990-01-01

    Study of thermostimulated luminescence of an alpha irradiated glass used as radionuclear waste glass has revealed the formation of a structural defect induced by alpha irradiation. To detect this structural modification the thermostimulated signal of an alpha irradiated sample is recorded under certain conditions. The nature of generated defects has been established using synthetic glasses of more simple composition such as silica or boro-silicate glasses. Results obtained with these simple glasses are transposed to alpha irradiated radionuclear waste glass. The problem is to see how autoirradiated glass could evolve in time. For this purpose actinide-doped glasses are now being fabricated and specific thermostimulated luminescence equipment has been developed for this purpose

  2. Solubility effects in waste-glass/demineralized-water systems

    International Nuclear Information System (INIS)

    Fullam, H.T.

    1981-06-01

    Aqueous systems involving demineralized water and four glass compositions (including standins for actinides and fission products) at temperatures of up to 150 0 C were studied. Two methods were used to measure the solubility of glass components in demineralized water. One method involved approaching equilibrium from subsaturation, while the second method involved approaching equilibrium from supersaturation. The aqueous solutions were analyzed by induction-coupled plasma spectrometry (ICP). Uranium was determined using a Scintrex U-A3 uranium analyzer and zinc and cesium were determined by atomic absorption. The system that results when a waste glass is contacted with demineralized water is a complex one. The two methods used to determine the solubility limits gave very different results, with the supersaturation method yielding much higher solution concentrations than the subsaturation method for most of the elements present in the waste glasses. The results show that it is impossible to assign solubility limits to the various glass components without thoroughly describing the glass-water systems. This includes not only defining the glass type and solution temperature, but also the glass surface area-to-water volume ratio (S/V) of the system and the complete thermal history of the system. 21 figures, 22 tables

  3. Application of waste glass in translucent and photocatalytic concrete

    NARCIS (Netherlands)

    Lieshout, van B.; Spiesz, P.R.; Brouwers, H.J.H.

    2012-01-01

    Container glass aggregates and glass powder are waste products of the glass recycling industry. In this research, these products are incorporated in self-compacting concrete (SCC) mixtures, replacing conventional aggregates and fine powders. The SCC mixtures were designed using a particle packing

  4. Materials interactions relating to long-term geologic disposal of nuclear waste glass

    International Nuclear Information System (INIS)

    Bibler, N.E.; Jantzen, C.M.

    1987-01-01

    In the geologic disposal of nuclear waste glass, the glass will eventually interact with groundwater in the repository system. Interactions can also occur between the glass and other waste package materials that are present. These include the steel canister that holds the glass, the metal overpack over the canister, backfill materials that may be used, and the repository host rock. This review paper systematizes the additional interactions that materials in the waste package will impose on the borosilicate glass waste form-groundwater interactions. The repository geologies reviewed are tuff, salt, basalt, and granite. The interactions emphasized are those appropriate to conditions expected after repository closure, e.g. oxic vs anoxic conditions. Whenever possible, the effect of radiation from the waste form on the interactions is examined. The interactions are evaluated based on their effect on the release and speciation of various elements including radionuclides from the glass. It is noted when further tests of repository interactions are needed before long-term predictions can be made. 63 references, 1 table

  5. Economic comparison of crystalline ceramic and glass waste forms for HLW disposal

    International Nuclear Information System (INIS)

    McKee, R.W.; Daling, P.M.; Wiles, L.E.

    1983-05-01

    A titanate-based, crystalline ceramic produced by hot isostatic pressing has been proposed as a potentially more stable and improved waste form for high-level nuclear waste disposal compared to the currently favored borosilicate glass waste form. This paper describes the results of a study to evaluate the relative costs for disposal of high-level waste from a 70,000 metric ton equivalent (MTE) system. The entire waste management system, including waste processing and encapsulation, transportation, and final repository disposal, was included in this analysis. The repository concept is based on the current basalt waste isolation project (BWIP) reference design. A range of design basis alternatives is considered to determine if this would influence the relative economics of the two waste forms. A thermal analysis procedure was utilized to define optimum canister sizes to assure that each waste form was compared under favorable conditions. Repository costs are found to favor the borosilicate glass waste form while transportation costs greatly favor the crystalline ceramic waste form. The determining component in the cost comparison is the waste processing cost, which strongly favors the borosilicate glass process because of its relative simplicity. A net cost advantage on the order of 12% to 15% on a waste management system basis is indicated for the glass waste form

  6. Glass and Glass-Ceramic Materials from Simulated Composition of Lunar and Martian Soils: Selected Properties and Potential Applications

    Science.gov (United States)

    Ray, C. S.; Sen, S.; Reis, S. T.; Kim, C. W.

    2005-01-01

    In-situ resource processing and utilization on planetary bodies is an important and integral part of NASA's space exploration program. Within this scope and context, our general effort is primarily aimed at developing glass and glass-ceramic type materials using lunar and martian soils, and exploring various applications of these materials for planetary surface operations. Our preliminary work to date have demonstrated that glasses can be successfully prepared from melts of the simulated composition of both lunar and martian soils, and the melts have a viscosity-temperature window appropriate for drawing continuous glass fibers. The glasses are shown to have the potential for immobilizing certain types of nuclear wastes without deteriorating their chemical durability and thermal stability. This has a direct impact on successfully and economically disposing nuclear waste generated from a nuclear power plant on a planetary surface. In addition, these materials display characteristics that can be manipulated using appropriate processing protocols to develop glassy or glass-ceramic magnets. Also discussed in this presentation are other potential applications along with a few selected thermal, chemical, and structural properties as evaluated up to this time for these materials.

  7. ANL technical support program for DOE environmental restoration and waste management. Annual report, October 1993--September 1994

    International Nuclear Information System (INIS)

    Bates, J.K.; Brown, N.R.; Buck, E.C.

    1995-06-01

    A program was established for DOE Environmental Restoration and Waste Management (EM) to evaluate factors that are anticipated to affect waste glass reaction during repository disposal, especially in an unsaturated environment typical of what may be expected for the proposed Yucca Mountain repository site. This report covers progress in FY 1994 on the following tasks: (1) Critical Reviews of important parameters that affect the reactivity of glass in an unsaturated environment are being prepared. (2) A series of tests is ongoing to evaluate the reactivity of fully radioactive glasses in a high-level waste repository environment and compare it to the reactivity of synthetic, nonradioactive glasses of similar composition. (3) The effect of radiation upon the durability of waste glasses at a high SA/V ratio and a high gas-to-liquid volume ratio has been assessed. (4) A series of tests is being performed to compare the extent of reaction of nuclear waste glasses at various SA/V ratios. Such differences in the SA/V ratio may significantly affect glass durability. At long-term periods and high SA/V ratios, acceleration in glass reaction has been observed. (5) Tests were initiated on West Valley Reference 6 (WV6) glass and on the Environmental Assessment (EA) glass. (6) Tests with the actinide-doped West Valley glass ATM-10 have been in progress for over seven years as a part of work for the Yucca Mountain Site Characterization Project (YMP). (7) Analytical electron microscopy (AEM) is being used to assess the glass/water reaction pathway by identifying intermediate phases that appear on the reacting glass. Also, colloids from the leach solutions are being studied using AEM

  8. A new viscosity model for waste glass formulations

    International Nuclear Information System (INIS)

    Sadler, A.L.K.

    1996-01-01

    Waste glass formulation requires prediction, with reasonable accuracy, of properties over much wider ranges of composition than are typically encountered in any single industrial application. Melt viscosity is one such property whose behavior must be predicted in formulating new waste glasses. A model was developed for silicate glasses which relates the Arrhenius activation energy for flow to an open-quotes effectiveclose quotes measure of non-bridging oxygen content in the melt, NBO eff . The NBO eff parameter incorporates the differing effects of modifying cations on the depolymerization of the silicate network. The activation energy-composition relationship implied by the model is in accordance with experimental behavior. The model was validated against two different databases, with satisfactory results

  9. Control of high level radioactive waste-glass melters

    International Nuclear Information System (INIS)

    Bickford, D.F.; Choi, A.S.

    1991-01-01

    Slurry Fed Melters (SFM) are being developed in the United States, Europe and Japan for the conversion of high-level radioactive waste to borosilicate glass for permanent disposal. The high transition metal, noble metal, nitrate, organic, and sulfate contents of these wastes lead to unique melter redox control requirements. Pilot waste-glass melter operations have indicated the possibility of nickel sulfide or noble-metal fission-product accumulation on melter floors, which can lead to distortion of electric heating patterns, and decrease melter life. Sulfide formation is prevented by control of the redox chemistry of the melter feed. The redox state of waste-glass melters is determined by balance between the reducing potential of organic compounds in the feed, and the oxidizing potential of gases above the melt, and nitrates and polyvalent elements in the waste. Semiquantitative models predicting limitations of organic content have been developed based on crucible testing. Computerized thermodynamic computations are being developed to predict the sequence and products of redox reactions and is assessing process variations. Continuous melter test results have been compared to improved computer staged-thermodynamic-models of redox behavior. Feed chemistry control to prevent sulfide and moderate noble metal accumulations are discussed. 17 refs., 3 figs

  10. A Glass-Ceramic Waste Forms for the Immobilization of Rare Earth Oxides from the Pyroprocessing Waste salt

    International Nuclear Information System (INIS)

    Ahn, Byung-Gil; Park, Hwan-Seo; Kim, Hwan-Young; Kim, In-Tae

    2008-01-01

    The fission product of rare earth (RE) oxide wastes are generates during the pyroprocess . Borosilicate glass or some ceramic materials such as monazite, apatite or sodium zirconium phosphate (NZP) have been a prospective host matrix through lots of experimental results. Silicate glasses have long been the preferred waste form for the immobilization of HLW. In immobilization of the RE oxides, the developed process on an industrial scale involves their incorporation into a glass matrix, by melting under 1200 ∼ 1300 .deg. C. Instead of the melting process, glass powder sintering is lower temperature (∼ 900 .deg. C) required for the process which implies less demanding conditions for the equipment and a less evaporation of volatile radionuclides. This study reports the behaviors, direct vitrification of RE oxides with glass frit, glass powder sintering of REceramic with glass frit, formation of RE-apatite (or REmonazite) ceramic according to reaction temperature, and the leach resistance of the solidified waste forms

  11. Zero photoelastic and water durable ZnO-SnO-P2O5-B2O3 glasses

    Science.gov (United States)

    Saitoh, Akira; Nakata, Kohei; Tricot, Grégory; Chen, Yuanyuan; Yamamoto, Naoki; Takebe, Hiromichi

    2015-04-01

    We report properties of zero birefringent xZnO-(67-x)SnO-(33-y)P2O5-y B2O3 glasses, within 18.5 ≤ x ≤ 22 and y = 0, 3, and 10 mol. %. These compositions of boro-phosphate glasses provide both zero photoelastic constant (PEC) and improved water durability. x = 19 and y = 3 compositions show minimum PEC of -0.002 × 10-12 Pa-1, which can contribute to candidate material for fiber current sensor devise without lead. The structures of zero photoelastic glasses were investigated by Raman scattering and nuclear magnetic resonance spectroscopies. Compositions of zero PEC glasses are explained by the empirical model proposed by Zwanziger et al. [Chem. Mater. 19, 286-290 (2007)].

  12. Durability aspects of high-performance concretes for a waste repository. Appendix 3: Canada

    International Nuclear Information System (INIS)

    Philipose, K.E.

    2001-01-01

    The IRUS facility for the disposal of low level radioactive waste at the Chalk River Nuclear Laboratories in Ontario, Canada relies on the durability of concrete for the required 500 years of service life. A research programme based on laboratory testing to design a durable concrete and assess its long-term behaviour was initiated in 1988. This appendix discusses the methodology to assess the long-term behaviour of concrete, and some initial observations. Longevity predictions for concrete formulations based on diffusion testing are also presented

  13. Nuclear waste under glass, further discussion

    Science.gov (United States)

    O'Keefe, J. A.; Barkatt, A.; Glass, B. P.; Alterescu, S.

    J. J. Crovisier and J. Honnorez [1988] discuss an article by W. W. Maggs, “Mg May Protect Waste Under Glass” [Maggs, 1988] summarizing work by A. Barkatt (Catholic University, Washington, D.C.), B. P. Glass (University of Delaware, Newark), and S. Alterescu and J. A. O'Keefe (NASA/GSFC, Greenbelt, Md.). We found that seawater is orders of magnitude less corrosive t h an fresh water in attacking tektite glass; traced the protective effect to the presence of magnesium, at a level of about 1.3 g/L in seawater; and suggested that the effect might be useful in protecting nuclear waste glasses from corrosion.Crovisier and Honnorez first make the point that the rate of corrosion of glass is, in principle, a function of the ratio of surface area 5 to the effective volume V. This concept, which is usually discussed in American literature under the name of S/V effects, is discussed by Crovisier and Honnorez in terms of the “permeability of the environment.” These effects have been carefully considered throughout our work (see, for example, Barkatt et al. [19867rsqb;). It turns out that in the sea the effective S/V is so small that the effects referred to by Crovisier and Honnorez can be ignored.

  14. Durability of cement and geopolimer composites

    Science.gov (United States)

    Błaszczyński, T.; Król, M.

    2017-10-01

    Concrete structures are constantly moving in the direction of improving the durability. This main feature depends on many factors, which are the composition of concrete mix, the usage of additives and admixtures and the place, where material will work and carry the load. The introduction of new geopolymer binders for geopolymer structures adds a new aspect that is type of used activator. This substance with strongly alkaline reaction is divided because of the physical state, the alkaline degree and above all the chemical composition. Taking into account, that at present the geopolymer binders are made essentially from waste materials or by products from the combustion of coal or iron ore smelting, unambiguous determination of the effect of the activator on the properties of the geopolymer material requires a number of trials, researches and observation. This paper shows the influence of the most alkaline activators on the basic parameters of the durability of geopolymer binders. In this study there were used a highly alkaline hydroxides, water glasses and granules, which are waste materials in a variety of processes taking place in a chemical plants. As the substrate of geopolymer binders there were used fly ash which came from coal and high calcium ash from the burning of lignite.

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

    Science.gov (United States)

    Piovesan, V.; Bardez-Giboire, I.; Perret, D.; Montouillout, V.; Pellerin, N.

    2017-01-01

    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.

  16. Characterization of Mechanical and Bactericidal Properties of Cement Mortars Containing Waste Glass Aggregate and Nanomaterials

    Science.gov (United States)

    Sikora, Pawel; Augustyniak, Adrian; Cendrowski, Krzysztof; Horszczaruk, Elzbieta; Rucinska, Teresa; Nawrotek, Pawel; Mijowska, Ewa

    2016-01-01

    The recycling of waste glass is a major problem for municipalities worldwide. The problem concerns especially colored waste glass which, due to its low recycling rate as result of high level of impurity, has mostly been dumped into landfills. In recent years, a new use was found for it: instead of creating waste, it can be recycled as an additive in building materials. The aim of the study was to evaluate the possibility of manufacturing sustainable and self-cleaning cement mortars with use of commercially available nanomaterials and brown soda-lime waste glass. Mechanical and bactericidal properties of cement mortars containing brown soda-lime waste glass and commercially available nanomaterials (amorphous nanosilica and cement containing nanocrystalline titanium dioxide) were analyzed in terms of waste glass content and the effectiveness of nanomaterials. Quartz sand is replaced with brown waste glass at ratios of 25%, 50%, 75% and 100% by weight. Study has shown that waste glass can act as a successful replacement for sand (up to 100%) to produce cement mortars while nanosilica is incorporated. Additionally, a positive effect of waste glass aggregate for bactericidal properties of cement mortars was observed. PMID:28773823

  17. Characterization of Mechanical and Bactericidal Properties of Cement Mortars Containing Waste Glass Aggregate and Nanomaterials

    Directory of Open Access Journals (Sweden)

    Pawel Sikora

    2016-08-01

    Full Text Available The recycling of waste glass is a major problem for municipalities worldwide. The problem concerns especially colored waste glass which, due to its low recycling rate as result of high level of impurity, has mostly been dumped into landfills. In recent years, a new use was found for it: instead of creating waste, it can be recycled as an additive in building materials. The aim of the study was to evaluate the possibility of manufacturing sustainable and self-cleaning cement mortars with use of commercially available nanomaterials and brown soda-lime waste glass. Mechanical and bactericidal properties of cement mortars containing brown soda-lime waste glass and commercially available nanomaterials (amorphous nanosilica and cement containing nanocrystalline titanium dioxide were analyzed in terms of waste glass content and the effectiveness of nanomaterials. Quartz sand is replaced with brown waste glass at ratios of 25%, 50%, 75% and 100% by weight. Study has shown that waste glass can act as a successful replacement for sand (up to 100% to produce cement mortars while nanosilica is incorporated. Additionally, a positive effect of waste glass aggregate for bactericidal properties of cement mortars was observed.

  18. Composition - structure - properties relationships of peraluminous glasses for nuclear waste containment

    International Nuclear Information System (INIS)

    Piovesan, Victor

    2016-01-01

    Part of the Research and Development program concerning high level nuclear waste conditioning aims to assess new glass formulations able to incorporate a high waste content with enhanced properties in terms of homogeneity, thermal stability, long term behavior and process ability. This study focuses on peraluminous glasses, defined by an excess of aluminum ions Al"3"+ in comparison with modifier elements such as Na"+, Li"+ or Ca"2"+. A Design of Experiment approach has been employed to determine relationships between composition of simplified peraluminous glasses (SiO_2 - B_2O_3 - Al_2O_3 - Na_2O - Li_2O - CaO - La_2O_3) and their physical properties such as viscosity, glass transition temperature and glass homogeneity. Moreover, some structural investigation (NMR) was performed in order to better understand the structural role of Na"+, Li"+ and Ca"2"+ and the structural organization of peraluminous glasses. Then, physical and chemical properties of fully simulated peraluminous glasses were characterized to evaluate transposition between simplified and fully simulated glasses and also to put forward the potential of peraluminous glasses for nuclear waste containment. (author) [fr

  19. Property/composition relationships for Hanford high-level waste glasses melting at 1150 degrees C volume 2: Chapters 12-16 and appendices A-K

    International Nuclear Information System (INIS)

    Hrma, P.R.; Piepel, G.F.

    1994-12-01

    A Composition Variation Study (CVS) is being performed within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) project in support of a future high-level nuclear waste vitrification plant at the Hanford site in Washington. From 1989 to 1994, over 120 nonradioactive glasses were melted and properties measured in five statistically-designed experimental phases. Glass composition is represented by the 10 components SiO 2 , B 2 O 3 , ZrO 2 , Na 2 O, Li 2 O, CaO, MgO, and Others (all remaining components). The properties measured include viscosity (η), electrical conductivity (ε), glass transition temperature (T g ), thermal expansion of solid glass (α s ) and molten glass (α m ), crystallinity (quenched and canister centerline cooled glasses), liquidus temperature (T L ), durability based on normalized elemental releases from the Materials Characterization Center-1 28-day dissolution test (MCC-1, r mi ) and the 7-day Product Consistency Test (PCT, r pi ), and solution pHs from MCC-1 and PCT. Amorphous phase separation was also evaluated. Empirical first- and second-order mixture models were fit using the CVS data to relate the various properties to glass composition. Equations for calculating the uncertainty associated with property values predicted by the models were also developed. The models were validated using both internal and external data. Other modeling approaches (e.g., non-bridging oxygen, free energy of hydration, phase-equilibria T L ) were investigated for specific properties. A preliminary Qualified Composition Region was developed to identify glass compositions with high confidence of being processable in a melter and meeting waste form acceptance criteria

  20. Time-Dependent Deformation Modelling for a Chopped-Glass Fiber Composite for Automotive Durability Design Criteria

    Energy Technology Data Exchange (ETDEWEB)

    Ren, W

    2001-08-24

    Time-dependent deformation behavior of a polymeric composite with chopped-glass-fiber reinforcement was investigated for automotive applications, The material under stress was exposed to representative automobile service environments. Results show that environment has substantial effects on time-dependent deformation behavior of the material. The data were analyzed and experimentally-based models developed for the time-dependent deformation behavior as a basis for automotive structural durability design criteria.

  1. High-Level Waste Glass Formulation Model Sensitivity Study 2009 Glass Formulation Model Versus 1996 Glass Formulation Model

    International Nuclear Information System (INIS)

    Belsher, J.D.; Meinert, F.L.

    2009-01-01

    This document presents the differences between two HLW glass formulation models (GFM): The 1996 GFM and 2009 GFM. A glass formulation model is a collection of glass property correlations and associated limits, as well as model validity and solubility constraints; it uses the pretreated HLW feed composition to predict the amount and composition of glass forming additives necessary to produce acceptable HLW glass. The 2009 GFM presented in this report was constructed as a nonlinear optimization calculation based on updated glass property data and solubility limits described in PNNL-18501 (2009). Key mission drivers such as the total mass of HLW glass and waste oxide loading are compared between the two glass formulation models. In addition, a sensitivity study was performed within the 2009 GFM to determine the effect of relaxing various constraints on the predicted mass of the HLW glass.

  2. Radioactive Demonstration Of Mineralized Waste Forms Made From Hanford Low Activity Waste (Tank SX-105 And AN-103) By Fluidized Bed Steam Reformation

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, Carol; Herman, Connie; Crawford, Charles; Bannochie, Christopher; Burket, Paul; Daniel, Gene; Cozzi, Alex; Nash, Charles; Miller, Donald; Missimer, David

    2014-01-10

    One of the immobilization technologies under consideration as a Supplemental Treatment for Hanford’s Low Activity Waste (LAW) is Fluidized Bed Steam Reforming (FBSR). The FBSR technology forms a mineral waste form at moderate processing temperatures thus retaining and atomically bonding the halides, sulfates, and technetium in the mineral phases (nepheline, sodalite, nosean, carnegieite). Additions of kaolin clay are used instead of glass formers and the minerals formed by the FBSR technology offers (1) atomic bonding of the radionuclides and constituents of concern (COC) comparable to glass, (2) short and long term durability comparable to glass, (3) disposal volumes comparable to glass, and (4) higher Na2O and SO{sub 4} waste loadings than glass. The higher FBSR Na{sub 2}O and SO{sub 4} waste loadings contribute to the low disposal volumes but also provide for more rapid processing of the LAW. Recent FBSR processing and testing of Hanford radioactive LAW (Tank SX-105 and AN-103) waste is reported and compared to previous radioactive and non-radioactive LAW processing and testing.

  3. Direct conversion of plutonium metal, scrap, residue, and transuranic waste to glass

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Beahm, E.C.; Parker, G.W.; Malling, J.F.; Rudolph, J.

    1995-01-01

    A method for the direct conversion of metals, ceramics, organics, and amorphous solids to borosilicate glass has been invented. The process is called the Glass Material Oxidation and Dissolution System (GMODS). Traditional glass-making processes can convert only oxide materials to glass. However, many wastes contain complex mixtures of metals, ceramics, organics, and amorphous solids. Conversion of such mixtures to oxides followed by their conversion to glass is often impractical. GMODS may create a practical method to convert such mixtures to glass. Plutonium-containing materials (PCMS) exist in many forms, including metals, ceramics, organics, amorphous solids, and mixtures thereof. These PCMs vary from plutonium metal to filters made of metal, organic binders, and glass fibers. For storage and/or disposal of PCMS, it is desirable to convert PCMs to borosilicate glass. Borosilicate glass is the preferred repository waste form for high-level waste (HLW) because of its properties. PCMs converted to a transuranic borosilicate homogeneous glass would easily pass all waste acceptance and storage criteria. Conversion of PCMs to a glass would also simplify safeguards by conversion of heterogeneous PCMs to homogeneous glass. Thermodynamic calculations and proof-of-principle experiments on the GMODS process with cerium (plutonium surrogate), uranium, stainless steel, aluminum, Zircaloy-2, and carbon were successfully conducted. Initial analysis has identified potential flowsheets and equipment. Major unknowns remain, but the preliminary data suggests that GMODS may be a major new treatment option for PCMs

  4. Borosilicate glass as a matrix for immobilization of SRP high-level waste

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1980-01-01

    Approximately 22 million gallons of high-level radioactive defense waste are currently being stored in large underground tanks located on the Savannah River Plant (SRP) site in Aiken, South Carolina. One option now being considered for long-term management of this waste involves removing the waste from the tanks, chemically processing the waste, and immobilizing the potentially harmful radionuclides in the waste into a borosilicate glass matrix. The technology for producing waste glass forms is well developed and has been demonstrated on various scales using simulated as well as radioactive SRP waste. Recently, full-scale prototypical equipment has been made operational at SRP. This includes both a joule-heated ceramic melter and an in-can melter. These melters are a part of an integrated vitrification system which is under evaluation and includes a spray calciner, direct liquid feed apparatus, and various elements of an off-gas system. Two of the most important properties of the waste glass are mechanical integrity and leachability. Programs are in progress at SRL aimed at minimizing thermally induced cracking by carefully controlling cooling cycles and using ceramic liners or coatings. The leachability of SRP waste glass has been studied under many different conditions and consistently found to be low. For example, the leachability of actual SRP waste glass was found to be 10 -6 to 10 -5 g/(cm 2 )(day) initially and decreasing to 10 -9 to 10 -8 g/(cm 2 )(day) after 100 days. Waste glass is also being studied under anticipated storage conditions. In brine at 90 0 C, the leachability is about 5 x 10 -8 g/(cm 2 )(day) after 60 days. The effects of other geological media including granite, basalt, shale, and tuff are also being studied as part of the multibarrier isolation system

  5. Impacts of Process and Prediction Uncertainties on Projected Hanford Waste Glass Amount

    Energy Technology Data Exchange (ETDEWEB)

    Gervasio, V.; Kim, D. S.; Vienna, J. D.; Kruger, A. A.

    2018-03-08

    Analyses were performed to evaluate the impacts of using the advanced glass models, constraints (Vienna et al. 2016), and uncertainty descriptions on projected Hanford glass mass. The maximum allowable waste oxide loading (WOL) was estimated for waste compositions while simultaneously satisfying all applicable glass property and composition constraints with sufficient confidence. Different components of prediction and composition/process uncertainties were systematically included in the calculations to evaluate their impacts on glass mass. The analyses estimated the production of 23,360 MT of immobilized high-level waste (IHLW) glass when no uncertainties were taken into account. Accounting for prediction and composition/process uncertainties resulted in 5.01 relative percent increase in estimated glass mass of 24,531 MT. Roughly equal impacts were found for prediction uncertainties (2.58 RPD) and composition/process uncertainties (2.43 RPD). The immobilized low-activity waste (ILAW) mass was predicted to be 282,350 MT without uncertainty and with waste loading “line” rules in place. Accounting for prediction and composition/process uncertainties resulted in only 0.08 relative percent increase in estimated glass mass of 282,562 MT. Without application of line rules the glass mass decreases by 10.6 relative percent (252,490 MT) for the case with no uncertainties. Addition of prediction uncertainties increases glass mass by 1.32 relative percent and the addition of composition/process uncertainties increase glass mass by an additional 7.73 relative percent (9.06 relative percent increase combined). The glass mass estimate without line rules (275,359 MT) was 2.55 relative percent lower than that with the line rules (282,562 MT), after accounting for all applicable uncertainties.

  6. Consolidated waste forms: glass marbles and ceramic pellets

    International Nuclear Information System (INIS)

    Treat, R.L.; Rusin, J.M.

    1982-05-01

    Glass marbles and ceramic pellets have been developed at Pacific Northwest Laboratory as part of the multibarrier concept for immobilizing high-level radioactive waste. These consolidated waste forms served as substrates for the application of various inert coatings and as ideal-sized particles for encapsulation in protective matrices. Marble and pellet formulations were based on existing defense wastes at Savannah River Plant and proposed commercial wastes. To produce marbles, glass is poured from a melter in a continuous stream into a marble-making device. Marbles were produced at PNL on a vibratory marble machine at rates as high as 60 kg/h. Other marble-making concepts were also investigated. The marble process, including a lead-encapsulation step, was judged as one of the more feasible processes for immobilizing high-level wastes. To produce ceramic pellets, a series of processing steps are required, which include: spray calcining - to dry liquid wastes to a powder; disc pelletizing - to convert waste powders to spherical pellets; sintering - to densify pellets and cause desired crystal formation. These processing steps are quite complex, and thereby render the ceramic pellet process as one of the least feasible processes for immobilizing high-level wastes

  7. Hydrogen speciation in hydrated layers on nuclear waste glass

    International Nuclear Information System (INIS)

    Aines, R.D.; Weed, H.C.; Bates, J.K.

    1987-01-01

    The hydration of an outer layer on nuclear waste glasses is known to occur during leaching, but the actual speciation of hydrogen (as water or hydroxyl groups) in these layers has not been determined. As part of the Nevada Nuclear Waste Storage Investigations Project, we have used infrared spectroscopy to determine hydrogen speciations in three nuclear waste glass compositions (SRL-131 and 165, and PNL 76-68), which were leached at 90 0 C (all glasses) or hydrated in a vapor-saturated atmosphere at 202 0 C (SRL-131 only). Hydroxyl groups were found in the surface layers of all the glasses. Molecular water was found in the surface of SRL-131 and PNL 76-68 glasses that had been leached for several months in deionized water, and in the vapor-hydrated sample. The water/hydroxyl ratio increases with increasing reaction time; molecular water makes up most of the hydrogen in the thick reaction layers on vapor-phase hydrated glass while only hydroxyl occurs in the least reacted samples. Using the known molar absorptivities of water and hydroxyl in silica-rich glass the vapor-phase layer contained 4.8 moles/liter of molecular water, and 0.6 moles water in the form hydroxyl. A 15 μm layer on SRL-131 glass formed by leaching at 90 0 C contained a total of 4.9 moles/liter of water, 2/3 of which was as hydroxyl. The unreacted bulk glass contains about 0.018 moles/liter water, all as hydroxyl. The amount of hydrogen added to the SRL-131 glass was about 70% of the original Na + Li content, not the 300% that would result from alkali=hydronium ion interdiffusion. If all the hydrogen is then assumed to be added as the result of alkali-H + interdiffusion, the molecular water observed may have formed from condensation of the original hydroxyl groups

  8. Feasibility Study for Preparation and Use of Glass Grains as an Alternative to Glass Nodules for Vitrification of Nuclear Waste

    Energy Technology Data Exchange (ETDEWEB)

    Sonavane, M S; Mishra, P.K., E-mail: maheshss@barc.gov.in [Nuclear Recycle Board, Bhabha Atomic Research Centre, Mumbai (India); Mandal, S; Barik, S; Roy Chowdhury, A; Sen, R [Central Glass and Ceramic Institute, Kolkata (India)

    2012-10-15

    High level nuclear liquid waste (HLW) is immobilized using borosilicate glass matrix. Presently joule heated ceramic melter is being employed for vitrification of HLW in India. Preformed nodules of base glass are fed to melter along with liquid waste in predetermined ratio. In order to reduce the cost incurred for production of glass nodules of base glass, an alternative option of using glass grains was evaluated for its preparation and its suitability for the melter operation. (author)

  9. Feasibility Study for Preparation and Use of Glass Grains as an Alternative to Glass Nodules for Vitrification of Nuclear Waste

    International Nuclear Information System (INIS)

    Sonavane, M.S.; Mishra, P.K.; Mandal, S.; Barik, S.; Roy Chowdhury, A.; Sen, R.

    2012-01-01

    High level nuclear liquid waste (HLW) is immobilized using borosilicate glass matrix. Presently joule heated ceramic melter is being employed for vitrification of HLW in India. Preformed nodules of base glass are fed to melter along with liquid waste in predetermined ratio. In order to reduce the cost incurred for production of glass nodules of base glass, an alternative option of using glass grains was evaluated for its preparation and its suitability for the melter operation. (author)

  10. Molecular Environmental Science Using Synchrotron Radiation: Chemistry and Physics of Waste Form Materials

    Energy Technology Data Exchange (ETDEWEB)

    Lindle, Dennis W.

    2011-04-21

    Production of defense-related nuclear materials has generated large volumes of complex chemical wastes containing a mixture of radionuclides. The disposition of these wastes requires conversion of the liquid and solid-phase components into durable, solid forms suitable for long-term immobilization. Specially formulated glass compositions and ceramics such as pyrochlores and apatites are the main candidates for these wastes. An important consideration linked to the durability of waste-form materials is the local structure around the waste components. Equally important is the local structure of constituents of the glass and ceramic host matrix. Knowledge of the structure in the waste-form host matrices is essential, prior to and subsequent to waste incorporation, to evaluate and develop improved waste-form compositions based on scientific considerations. This project used the soft-x-ray synchrotron-radiation-based technique of near-edge x-ray-absorption fine structure (NEXAFS) as a unique method for investigating oxidation states and structures of low-Z elemental constituents forming the backbones of glass and ceramic host matrices for waste-form materials. In addition, light metal ions in ceramic hosts, such as titanium, are also ideal for investigation by NEXAFS in the soft-x-ray region. Thus, one of the main objectives was to understand outstanding issues in waste-form science via NEXAFS investigations and to translate this understanding into better waste-form materials, followed by eventual capability to investigate “real” waste-form materials by the same methodology. We conducted several detailed structural investigations of both pyrochlore ceramic and borosilicate-glass materials during the project and developed improved capabilities at Beamline 6.3.1 of the Advanced Light Source (ALS) to perform the studies.

  11. Using of borosilicate glass waste as a cement additive

    International Nuclear Information System (INIS)

    Han, Weiwei; Sun, Tao; Li, Xinping; Sun, Mian; Lu, Yani

    2016-01-01

    Highlights: • Borosilicate glass waste used as cement additive can improves its radiation shielding. • When content is 14.8%, the linear attenuation coefficient is 0.2457 cm"−"1 after 28 d. • From 0 to 22.2%, linear attenuation coefficient firstly increase and then decrease. - Abstract: Borosilicate glass waste is investigated as a cement additive in this paper to improve the properties of cement and concrete, such as setting time, compressive strength and radiation shielding. The results demonstrate that borosilicate glass is an effective additive, which not only improves the radiation shielding properties of cement paste, but also shows the irradiation effect on the mechanical and optical properties: borosilicate glass can increase the compressive strength and at the same time it makes a minor impact on the setting time and main mineralogical compositions of hydrated cement mixtures; and when the natural river sand in the mortar is replaced by borosilicate glass sand (in amounts from 0% to 22.2%), the compressive strength and the linear attenuation coefficient firstly increase and then decrease. When the glass waste content is 14.8%, the compressive strength is 43.2 MPa after 28 d and the linear attenuation coefficient is 0.2457 cm"−"1 after 28 d, which is beneficial for the preparation of radiation shielding concrete with high performances.

  12. Using of borosilicate glass waste as a cement additive

    Energy Technology Data Exchange (ETDEWEB)

    Han, Weiwei [State Key Laboratory of Silicate Materials for Architectures, Wuhan University of Technology, Wuhan, Hubei 430070 (China); School of Civil Engineering and Architecture, Wuhan University of Technology, Wuhan, Hubei 430070 (China); Sun, Tao, E-mail: sunt@whut.edu.cn [State Key Laboratory of Silicate Materials for Architectures, Wuhan University of Technology, Wuhan, Hubei 430070 (China); Key Laboratory of Roadway Bridge & Structure Engineering, Wuhan University of Technology, Wuhan, Hubei 430070 (China); Li, Xinping [Key Laboratory of Roadway Bridge & Structure Engineering, Wuhan University of Technology, Wuhan, Hubei 430070 (China); Sun, Mian [School of Materials Science and Engineering, Wuhan University of Technology, Wuhan, Hubei 430070 (China); Lu, Yani [Urban Construction Institute, Hubei Engineering University, Xiaogan, Hubei 432000 (China)

    2016-08-15

    Highlights: • Borosilicate glass waste used as cement additive can improves its radiation shielding. • When content is 14.8%, the linear attenuation coefficient is 0.2457 cm{sup −1} after 28 d. • From 0 to 22.2%, linear attenuation coefficient firstly increase and then decrease. - Abstract: Borosilicate glass waste is investigated as a cement additive in this paper to improve the properties of cement and concrete, such as setting time, compressive strength and radiation shielding. The results demonstrate that borosilicate glass is an effective additive, which not only improves the radiation shielding properties of cement paste, but also shows the irradiation effect on the mechanical and optical properties: borosilicate glass can increase the compressive strength and at the same time it makes a minor impact on the setting time and main mineralogical compositions of hydrated cement mixtures; and when the natural river sand in the mortar is replaced by borosilicate glass sand (in amounts from 0% to 22.2%), the compressive strength and the linear attenuation coefficient firstly increase and then decrease. When the glass waste content is 14.8%, the compressive strength is 43.2 MPa after 28 d and the linear attenuation coefficient is 0.2457 cm{sup −1} after 28 d, which is beneficial for the preparation of radiation shielding concrete with high performances.

  13. Effects of waste glass and waste foundry sand additions on reclaimed tiles containing sewage sludge ash.

    Science.gov (United States)

    Lin, Deng-Fong; Luo, Huan-Lin; Lin, Kuo-Liang; Liu, Zhe-Kun

    2017-07-01

    Applying sewage sludge ash (SSA) to produce reclaimed tiles is a promising recycling technology in resolving the increasing sludge wastes from wastewater treatment. However, performance of such reclaimed tiles is inferior to that of original ceramic tiles. Many researchers have therefore tried adding various industrial by-products to improve reclaimed tile properties. In this study, multiple materials including waste glass and waste foundry sand (WFS) were added in an attempt to improve physical and mechanical properties of reclaimed tiles with SSA. Samples with various combinations of clay, WFS, waste glass and SSA were made with three kiln temperatures of 1000°C, 1050°C, and 1100°C. A series of tests on the samples were next conducted. Test results showed that waste glass had positive effects on bending strength, water absorption and weight loss on ignition, while WFS contributed the most in reducing shrinkage, but could decrease the tile bending strength when large amount was added at a high kiln temperature. This study suggested that a combination of WFS from 10% to 15%, waste glass from 15% to 20%, SSA at 10% at a kiln temperature between 1000°C and 1050°C could result in quality reclaimed tiles with a balanced performance.

  14. Task plan: Temperatures in DWPF Glass Waste Storage Building

    International Nuclear Information System (INIS)

    Hardy, B.J.

    1993-01-01

    The Bechtel National, Inc. Detailed Design Instructions for Structural Design (DDI-02) requires that concrete components of the GWSB not exceed 150 degrees F for structural elements and 200 degrees F locally over a 24 hour period. In addition, the Waste Acceptance Product Specifications (WAPS) sets the maximum post cooldown temperature of the glass waste-form at 400 degrees C. Various scenarios can be postulated which result in elevated glass and concrete temperatures in the GWSB. Therefore, it is important to determine the concrete and glass temperatures during both normal and off-normal conditions. This document details specific tasks required to develop a technically defensible and verifiable methodology for determining maximum temperatures for the waste-forms and the GWSB concrete structures. All models used in this analysis will satisfy Quality Assurance requirements and be defensible to review and oversight committees

  15. Development of glass compositions with 9% waste content for the vitrification of high-level waste from LWR nuclear reactors

    International Nuclear Information System (INIS)

    Lakatos, T.

    1979-10-01

    Reduction of the contents of waste in glass from 20-25% to 9% causes a decrease of the leaching resistance of the glass. The addition of Zn0 reduces the leaching values by a factor of approximately 10. The crystallized glass ceramics have a lower coefficient of thermal expansion than glassy waste bodies. The separation of the phase which contains Mo occurs during heat treatment. The amount of separated Mo is lower for low alkali sac type (Si0 2 - A1 2 0 3 -Ca0 system) of glasses by a factor of approximately 50. All the glasses were prepared with simulated waste composition. (GBn.)

  16. Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

    International Nuclear Information System (INIS)

    Boatner, L.A.; Sales, B.C.

    1989-01-01

    This patent describes lead-iron phosphate glasses containing a high level of Fe 2 O 3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90 0 C, with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe 2 O 3 in forming the lead-iron phosphate glass is critical. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms

  17. Localized chemistry of 99Tc in simulated low activity waste glass

    Science.gov (United States)

    Weaver, Jamie L.

    A priority of the United States Department of Energy (DOE) is to dispose of the nuclear waste accumulated in the underground tanks at the Hanford Nuclear Reservation in Richland, WA. Incorporation and stabilization of technetium (99Tc) from these tanks into vitrified waste forms is a concern to the waste glass community and DOE due to 99Tc's long half-life ( 2.13˙105 y), and its high mobility in the subsurface environment under oxidizing conditions. Working in collaboration with researchers at Pacific Northwest National Laboratory (PNNL) and other national laboratories, plans were formulated to obtain first-of-a-kind chemical structure determination of poorly understood and environmentally relevant technetium compounds that relate to the chemistry of the Tc in nuclear waste glasses. Knowledge of the structure and spectral signature of these compounds aid in refining the understanding of 99Tc incorporation into and release from oxide based waste glass. In this research a first-of-its kind mechanism for the behavior of 99Tc during vitrification is presented, and the structural role of Tc(VII) and (IV) in borosilicate waste glasses is readdressed.

  18. Long-term durability experiments with concrete-based waste packages in simulated repository conditions

    International Nuclear Information System (INIS)

    Ipatti, A.

    1993-03-01

    Two extensive experiments on long-term durability of waste packages in simulated repository conditions are described. The first one is a 'half-scale experiment' comprising radioactive waste product and half-scale concrete containers in site specific groundwater conditions. The second one is 'full-scale experiment' including simulated inactive waste product and full-scale concrete container stored in slowly flowing fresh water. The scope of the experiments is to demonstrate long-term behaviour of the designed waste packages in contact with moderately concrete aggressive groundwater, and to evaluate the possible interactions between the waste product, concrete container and ground water. As the waste packages are made of high-quality concrete, provisions have been made to continue the experiments for several years

  19. Glass formulation development and offgas analysis of microwave melter powder samples

    International Nuclear Information System (INIS)

    Semones, G.B.; Hoffman, C.R.; Phillips, J.A.

    1994-04-01

    Production of nuclear materials for defense applications has resulted in the accumulation of vast amounts of nuclear waste. This contaminated waste is in a variety of forms that require subsequent reprocessing to isolate and encapsulate the nuclear (e.g., uranium, plutonium, strontium, cesium, and americium) and toxic (e.g., lead, chromium, and cadmium) constituents. The encapsulating material must possess good chemical and mechanical durability to resist leaching of the nuclear and toxic constituents into the environment during permanent storage at a waste repository. Glass is an ideal encapsulating material because its open structure allows the introduction of different waste forms and the final vitreous product possesses a high degree of chemical stability. Microwave heating and melting is a relatively new advancement in glass processing which uses microwave radiation to heat the glass formers to adequate temperatures for sintering or melting. An advantage to this technique is that it enables more rapid heating than traditional heating mechanisms. This decrease in cycle time may help to limit exposure to workers encapsulating radioactive and/or toxic waste

  20. Aqueous corrosion of silicate glasses. Analogy between volcanic glasses and the French nuclear waste glass R7T7

    International Nuclear Information System (INIS)

    Goldschmidt, F.

    1991-01-01

    The behaviour of borosilicate glasses upon aqueous corrosion is controlled for long periods of time (>10,000 years) by processes which are not directly accessible by means of laboratory experiments. The analogical approach consists here to compare leaching performances between the french nuclear waste glass R7T7 and natural volcanic glasses, basaltic and rhyolitic ones. The three glasses were leached in the same conditions; open system, 90 deg C, initial pH of 9.7. Basaltic and R7T7 glasses having the same kinetic of dissolution, the basaltic glass was chosen as the best analogue. (author). refs., figs., tabs

  1. In-situ thermoelectric stabilization of radioactive wastes

    International Nuclear Information System (INIS)

    Brouns, R.A.; Timmerman, C.L.

    1982-02-01

    A new process for stabilizing buried radioactive wastes without exhumation is being developed by Pacific Northwest Laboratory (PNL). The process, known as in situ vitrification, converts waste and contaminated soil to a durable glass and crystalline material by passing an electric current between electrodes placed in the ground. Joule heating created by the flowing current has generated temperatures over 1700 0 C which cause the soil to melt and dissolve or encapsulate the wastes. Engineering-scale tests conducted in the laboratory have melted approximately 45 kgs (30 liters) of soil at a time by this technique. Encouraging results from these engineering-scale tests led to the design and construction of a pilot-scale field test unit which has solidified approximately 9000 kg of simulated contaminated soil per test. Test results and evaluations to date have been very promising. No detectable migration of hazardous species into uncontaminated soil has been found, and volatilization during melting has been very low. Leach studies have found the vitrified soil to be a highly durable waste form similar to pyrex glass. Electrical power costs to solidify a disposal site have been calculated at less than $70 per cubic meter ($2/ft 3 ) of waste. Future activities include both radioactive and nonradioactive pilot and large-scale tests

  2. Composition and redox control of waste glasses: Recommendation for process control limit

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Plodinec, M.J.

    1986-01-01

    An electrochemical series of redox couples, originally developed for Savannah River Laboratory glass frit 131 (SRL-131) as a reference composition, has been extended to two other alkali borosilicate compositions that are candidate glasses for nuclear waste immobilization. Since no dramatic differences were ascertained in the redox chemistry of selected multivalent elements in SRL-131 versus that in Savannah River Laboratory glass frit 165 (SRL-165) and in West Valley glass number-sign 205 (WV-205), the comprehensive electrochemical series can readily be applied to a range of nuclear waste glass compositions. In order to alleviate potential problems with foaming and precipitation of insolubles during the processing of the nuclear waste in these glass melts, the [Fe 2+ ]/[Fe 3+ ] ratio of the melt should be between 0.1 and 0.5. 27 refs., 4 figs., 2 tabs

  3. Experimental design of a waste glass study

    International Nuclear Information System (INIS)

    Piepel, G.F.; Redgate, P.E.; Hrma, P.

    1995-04-01

    A Composition Variation Study (CVS) is being performed to support a future high-level waste glass plant at Hanford. A total of 147 glasses, covering a broad region of compositions melting at approximately 1150 degrees C, were tested in five statistically designed experimental phases. This paper focuses on the goals, strategies, and techniques used in designing the five phases. The overall strategy was to investigate glass compositions on the boundary and interior of an experimental region defined by single- component, multiple-component, and property constraints. Statistical optimal experimental design techniques were used to cover various subregions of the experimental region in each phase. Empirical mixture models for glass properties (as functions of glass composition) from previous phases wee used in designing subsequent CVS phases

  4. Vitrification of low level and mixed (radioactive and hazardous) wastes: Lessons learned from high level waste vitrification

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1994-01-01

    Borosilicate glasses will be used in the USA and in Europe immobilize radioactive high level liquid wastes (HLLW) for ultimate geologic disposal. Simultaneously, tehnologies are being developed by the US Department of Energy's (DOE) Nuclear Facility sites to immobilize low-level and mixed (radioactive and hazardous) wastes (LLMW) in durable glass formulations for permanent disposal or long-term storage. Vitrification of LLMW achieves large volume reductions (86--97 %) which minimize the associated long-term storage costs. Vitrification of LLMW also ensures that mixed wastes are stabilized to the highest level reasonably possible, e.g. equivalent to HLLW, in order to meet both current and future regulatory waste disposal specifications The tehnologies being developed for vitrification of LLMW rely heavily on the technologies developed for HLLW and the lessons learned about process and product control

  5. Applicability of slags as waste forms for hazardous waste

    International Nuclear Information System (INIS)

    Bates, J.K.; Buck, E.C.; Dietz, N.L.; Wronkiewicz, D.J.; Feng, X.; Whitworth, C.; Filius, K.; Battleson, D.

    1994-01-01

    Slags, which are a combination of glassy and ceramic phases, were produced by the Component Development and Integration Facility, using a combination of soil and metal feeds. The slags were tested for durability using accelerated test methods in both water vapor and liquid water for time periods up to 179 days. The results indicated that under both conditions there was little reaction of the slag, in terms of material released to solution, or the reaction of the slag to form secondary mineral phases. The durability of the slags tested exceeded that of current high-level nuclear glass formulations and are viable materials, for waste disposal

  6. Determination of the Structure of Vitrified Hydroceramic/CBC Waste Form Glasses Manufactured from DOE Reprocessing Waste

    International Nuclear Information System (INIS)

    Scheetz, B.E.; White, W. B.; Chesleigh, M.; Portanova, A.; Olanrewaju, J.

    2005-01-01

    The selection of a glass-making option for the solidification of nuclear waste has dominated DOE waste form programs since the early 1980's. Both West Valley and Savannah River are routinely manufacturing glass logs from the high level waste inventory in tank sludges. However, for some wastes, direct conversion to glass is clearly not the optimum strategy for immobilization. INEEL, for example, has approximately 4400 m 3 of calcined high level waste with an activity that produces approximately 45 watts/m 3 , a rather low concentration of radioactive constituents. For these wastes, there is value in seeking alternatives to glass. An alternative approach has been developed and the efficacy of the process demonstrated that offers a significant savings in both human health and safety exposures and also a lower cost relative to the vitrification option. The alternative approach utilizes the intrinsic chemical reactivity of the highly alkaline waste with the addition of aluminosilicate admixtures in the appropriate proportions to form zeolites. The process is one in which a chemically bonded ceramic is produced. The driving force for reaction is derived from the chemical system itself at very modest temperatures and yet forms predominantly crystalline phases. Because the chemically bonded ceramic requires an aqueous medium to serve as a vehicle for the chemical reaction, the proposed zeolite-containing waste form can more adequately be described as a hydroceramic. The hydrated crystalline materials are then subject to hot isostatic pressing (HIP) which partially melts the material to form a glass ceramic. The scientific advantages of the hydroceramic/CBC approach are: (1) Low temperature processing; (2) High waste loading and thus only modest volumetric bulking from the addition of admixtures; (3) Ability to immobilize sodium; (4) Ability to handle low levels of nitrate (2-3% NO 3 - ); (5) The flexibility of a vitrifiable waste; and (6) A process that is based on an

  7. First use of in situ vitrification on radioactive wastes

    International Nuclear Information System (INIS)

    Bowlds, L.

    1992-01-01

    A high-temperature method for containing hazardous wastes, which was first developed in the 1980s, is being adapted for the in situ treatment of buried radioactive wastes by the US DOE's Idaho National Engineering Laboratory (INEL), following its recent report on successful preliminary tests. The method, called in situ vitrification (ISV), is an electrically induced thermal process that melts and fuses soil and wastes into a glass-like material at least as strong as natural obsidian or granite. Gases released during the process are captured and treated by an off-gas treatment system. After the wastes are vitrified, they could be left in place, or the mass could be broken up and transported to a disposal site. The glass-like substance would be chemically and physically similar to obsidian and from 4 to 10 times more durable than typical borosilicate glasses used to immobolize high-level nuclear wastes

  8. Chemical analysis of simulated high level waste glasses to support stage III sulfate solubility modeling

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-17

    The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms within the DOE complex. These wastes can contain relatively high concentrations of sulfate, which has low solubility in borosilicate glass. This is a significant issue for low-activity waste (LAW) glass and is projected to have a major impact on the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Sulfate solubility has also been a limiting factor for recent high level waste (HLW) sludge processed at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). The low solubility of sulfate in glass, along with melter and off-gas corrosion constraints, dictate that the waste be blended with lower sulfate concentration waste sources or washed to remove sulfate prior to vitrification. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerate mission completion.The objective of the current scope being pursued by SHU is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DWPF and WTP, allowing for enhanced waste loadings and waste throughput at these facilities. A series of targeted glass compositions was selected to resolve data gaps in the model and is identified as Stage III. SHU fabricated these glasses and sent samples to SRNL for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for the Stage III, simulated HLW glasses fabricated by SHU in support of the sulfate solubility model development.

  9. Fusibility of medical glass in hospital waste incineration: Effect of glass components

    International Nuclear Information System (INIS)

    Jiang, X.G.; An, C.G.; Li, C.Y.; Fei, Z.W.; Jin, Y.Q.; Yan, J.H.

    2009-01-01

    Medical glass, which is the principal incombustible component in hospital wastes, has a bad influence on combustion. In a rotary kiln incinerator, medical glass melts and turns into slag, possibly adhering to the inner wall. Prediction of the melting characteristics of medical glass hence is important for preventing slagging. The effect of various glass components on fusibility has been investigated experimentally; that of Na 2 O is the most marked. The softening temperature and flow temperature decrease 19.8 o C and 34.0 o C, respectively, with a rise of Na 2 O content in the Basic Content (standard composition of medical glass) of 1%. Correlations between fusion temperatures and glass components have been investigated; predictive functions of four characteristic melting temperatures have been obtained by simplifying the multi-variant series and were verified by testing glass samples. Relative errors of fusion temperatures (computed vs. measured) are mostly less than 5%.

  10. The role of natural glasses as analogues in projecting the long-term alteration of high-level nuclear waste glasses: Part 1

    International Nuclear Information System (INIS)

    Mazer, J.J.

    1993-01-01

    The common observation of glasses persisting in natural environments for long periods of time (up to tens of millions of years) provides compelling evidence that these materials can be kinetically stable in a variety of subsurface environments. This paper reviews how natural and historical synthesized glasses can be employed as natural analogues for understanding and projecting the long-term alteration of high-level nuclear waste glasses. The corrosion of basaltic glass results in many of the same alteration features found in laboratory testing of the corrosion of high-level radioactive waste glasses. Evidence has also been found indicating similarities in the rate controlling processes, such as the effects of silica concentration on corrosion in groundwater and in laboratory leachates. Naturally altered rhyolitic glasses and tektites provide additional evidence that can be used to constrain estimates of long-term waste glass alteration. When reacted under conditions where water is plentiful, the corrosion for these glasses is dominated by network hydrolysis, while the corrosion is dominated by molecular water diffusion and secondary mineral formation under conditions where water contact is intermittent or where water is relatively scarce. Synthesized glasses that have been naturally altered result in alkali-depleted alteration features that are similar to those found for natural glasses and for nuclear waste glasses. The characteristics of these alteration features appear to be dependent on the alteration conditions which affect the dominant reaction processes during weathering. In all cases, care must be taken to ensure that the information being provided by natural analogues is related to nuclear waste glass corrosion in a clear and meaningful way

  11. Redox reaction and foaming in nuclear waste glass melting

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, J.L.

    1995-08-01

    This document was prepared by Pacific Northwest Laboratory (PNL) and is an attempt to analyze and estimate the effects of feed composition variables and reducing agent variables on the expected chemistry of reactions occurring in the cold cap and in the glass melt in the nuclear waste glass Slurry-fed, joule-heated melters as they might affect foaming during the glass-making process. Numerous redox reactions of waste glass components and potential feed additives, and the effects of other feed variables on these reactions are reviewed with regard to their potential effect on glass foaming. A major emphasis of this report is to examine the potential positive or negative aspects of adjusting feed with formic acid as opposed to other feed modification techniques including but not limited to use of other reducing agents. Feed modification techniques other than the use of reductants that should influence foaming behavior include control of glass melter feed pH through use of nitric acid. They also include partial replacement of sodium salts by lithium salts. This latter action (b) apparently lowers glass viscosity and raises surface tension. This replacement should decrease foaming by decreasing foam stability.

  12. Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

    Science.gov (United States)

    Boatner, Lynn A.; Sales, Brian C.

    1989-01-01

    Lead-iron phosphate glasses containing a high level of Fe.sub.2 O.sub.3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90.degree. C., with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10.sup.2 to 10.sup.3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe.sub.2 O.sub.3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800.degree. C., since they exhibit very low melt viscosities in the 800.degree. to 1050.degree. C. temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550.degree. C. and are not adversely affected by large doses of gamma radiation in H.sub.2 O at 135.degree. C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms.

  13. On confirmation of abandonment of imported waste (glass solidified bodies) outside business places

    International Nuclear Information System (INIS)

    1996-01-01

    Electric power companies entrust the reprocessing of spent fuel generated from nuclear power stations to COGEMA in France, and in April, 1995, 28 high level radioactive wastes (glass solidified bodies) generated by the reprocessing were returned. When these glass solidified wastes are abandoned in the waste management facility of Japan Nuclear Fuel Service Co., it was decided to receive the confirmation of the prime minister on the measures based on the relevant law. Four electric power companies submitted the application and the explanation paper. As to the contents of the glass solidified wastes, the technical inspection was carried out by Bureau Veritas. Considering that this import of glass solidified wastes is the first in Japan, Science and Technology Agency carried out the measurement of all 28 wastes. The results are reported. It was confirmed that the measures for the abandonment taken by four electric power companies conform to the stipulation. The contents of the confirmation are reported in the order of the stipulation. These wastes were solidified with borosilicate glass in 5 mm thick stainless steel vessels, and the welding was done properly. (K.I.)

  14. ENHANCED DOE HIGH LEVEL WASTE MELTER THROUGHPUT STUDIES: SRNL GLASS SELECTION STRATEGY

    Energy Technology Data Exchange (ETDEWEB)

    Raszewski, F; Tommy Edwards, T; David Peeler, D

    2008-01-23

    The Department of Energy has authorized a team of glass formulation and processing experts at the Savannah River National Laboratory (SRNL), the Pacific Northwest National Laboratory (PNNL), and the Vitreous State Laboratory (VSL) at Catholic University of America to develop a systematic approach to increase high level waste melter throughput (by increasing waste loading with minimal or positive impacts on melt rate). This task is aimed at proof-of-principle testing and the development of tools to improve waste loading and melt rate, which will lead to higher waste throughput. Four specific tasks have been proposed to meet these objectives (for details, see WSRC-STI-2007-00483): (1) Integration and Oversight, (2) Crystal Accumulation Modeling (led by PNNL)/Higher Waste Loading Glasses (led by SRNL), (3) Melt Rate Evaluation and Modeling, and (4) Melter Scale Demonstrations. Task 2, Crystal Accumulation Modeling/Higher Waste Loading Glasses is the focus of this report. The objective of this study is to provide supplemental data to support the possible use of alternative melter technologies and/or implementation of alternative process control models or strategies to target higher waste loadings (WLs) for the Defense Waste Processing Facility (DWPF)--ultimately leading to higher waste throughputs and a reduced mission life. The glass selection strategy discussed in this report was developed to gain insight into specific technical issues that could limit or compromise the ability of glass formulation efforts to target higher WLs for future sludge batches at the Savannah River Site (SRS). These technical issues include Al-dissolution, higher TiO{sub 2} limits and homogeneity issues for coupled-operations, Al{sub 2}O{sub 3} solubility, and nepheline formation. To address these technical issues, a test matrix of 28 glass compositions has been developed based on 5 different sludge projections for future processing. The glasses will be fabricated and characterized based on

  15. High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 3

    International Nuclear Information System (INIS)

    Cunnane, J.C.

    1994-03-01

    The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II

  16. High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    Cunnane, J.C. [comp.; Bates, J.K.; Bradley, C.R. [Argonne National Lab., IL (United States)] [and others

    1994-03-01

    The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II.

  17. Fabrication, characterization, and evaluation of a fully radioactive glass

    International Nuclear Information System (INIS)

    Olson, K.M.; Elliott, M.L.; Shade, J.W.; Smith, H.D.

    1991-01-01

    West Valley Sludge Glass-1 (WVSG-1) was fabricated using high-level waste from the West Valley Demonstration Project. Melt foaming was a problem during fabrication and a unique two-step funnel system was designed to feed the melting calcine into a crucible. The resultant glass was essentially bubble free. Crushed WVSG-1 was tested for durability using the Product Consistency Test developed at WSRC. Except for Al, the release of nonradioactive elements from WVSG-1 was lower than the release of the same elements from ATM-10, the West Valley reference glass and the release of the radionuclides Th, U, and Tc was about the same as for ATM-10

  18. Recycling of inorganic waste in monolithic and cellular glass-based materials for structural and functional applications.

    Science.gov (United States)

    Rincón, Acacio; Marangoni, Mauro; Cetin, Suna; Bernardo, Enrico

    2016-07-01

    The stabilization of inorganic waste of various nature and origin, in glasses, has been a key strategy for environmental protection for the last decades. When properly formulated, glasses may retain many inorganic contaminants permanently, but it must be acknowledged that some criticism remains, mainly concerning costs and energy use. As a consequence, the sustainability of vitrification largely relies on the conversion of waste glasses into new, usable and marketable glass-based materials, in the form of monolithic and cellular glass-ceramics. The effective conversion in turn depends on the simultaneous control of both starting materials and manufacturing processes. While silica-rich waste favours the obtainment of glass, iron-rich wastes affect the functionalities, influencing the porosity in cellular glass-based materials as well as catalytic, magnetic, optical and electrical properties. Engineered formulations may lead to important reductions of processing times and temperatures, in the transformation of waste-derived glasses into glass-ceramics, or even bring interesting shortcuts. Direct sintering of wastes, combined with recycled glasses, as an example, has been proven as a valid low-cost alternative for glass-ceramic manufacturing, for wastes with limited hazardousness. The present paper is aimed at providing an up-to-date overview of the correlation between formulations, manufacturing technologies and properties of most recent waste-derived, glass-based materials. © 2016 The Authors. Journal of Chemical Technology & Biotechnology published by John Wiley & Sons Ltd on behalf of Society of Chemical Industry.

  19. Electrical resistivities of glass melts containing simulated SRP waste sludges

    International Nuclear Information System (INIS)

    Wiley, J.R.

    1978-08-01

    One option for the long-term management of radioactive waste at the Savannah River Plant is to solidify the waste in borosilicate glass by using a continuous, joule-heated, ceramic melter. Electrical resistivities that are needed for melter design were measured for melts of two borosilicate, glass-forming mixtures, each of which was combined with various amounts of several simulated-waste sludges. The simulated sludge spanned the composition range of actual sludges sampled from SRP waste tanks. Resistivities ranged from 6 to 10 ohm-cm at 500 0 C. Melt composition and temperature were correlated with resistivity. Resistivity was not a simple function of viscosity. 15 figures, 4 tables

  20. Antagonist effects of calcium on borosilicate glass alteration

    Energy Technology Data Exchange (ETDEWEB)

    Mercado-Depierre, S. [CEA Marcoule, DTCD SPDE LCLT, 30207 Bagnols sur Cèze (France); Angeli, F., E-mail: frederic.angeli@cea.fr [CEA Marcoule, DTCD SPDE LCLT, 30207 Bagnols sur Cèze (France); Frizon, F. [CEA Marcoule, DTCD SECM LP2C, 30207 Bagnols sur Cèze (France); Gin, S. [CEA Marcoule, DTCD SPDE LCLT, 30207 Bagnols sur Cèze (France)

    2013-10-15

    Graphical abstract: Display Omitted -- Highlights: •Kinetic study of glass alteration is investigated in calcium-enriched solutions. •New insights into silicon–calcium interactions in glass/cement systems are proposed. •Glass alteration is controlled by pH, Ca concentration and reaction progress. •Evidence of antagonist effects according to the importance of these parameters. -- Abstract: Numerous studies have been conducted on glass and cement durability in contact with water, but very little work to date has focused directly on interactions between the two materials. These interactions are mostly controlled by silicon–calcium reactivity. However, the physical and chemical processes involved remain insufficiently understood to predict the evolution of coupled glass–cement systems used in several industrial applications. Results are reported from borosilicate glass alteration in calcium-rich solutions. Our data show that four distinct behaviors can be expected according to the relative importance of three key parameters: the pH, the reaction progress (short- or long-term alteration) and the calcium concentration. Glass alteration is thus controlled by specific mechanisms depending on the solution chemistry: calcium complexation at the glass surface, precipitation of calcium silicate hydrates (C–S–H) or calcium incorporation in the altered layer. These findings highlight the impact of silicon–calcium interactions on glass durability and open the way for a better understanding of glass–cement mixing in civil engineering applications as well as in nuclear waste storage.

  1. Antagonist effects of calcium on borosilicate glass alteration

    International Nuclear Information System (INIS)

    Mercado-Depierre, S.; Angeli, F.; Frizon, F.; Gin, S.

    2013-01-01

    Graphical abstract: Display Omitted -- Highlights: •Kinetic study of glass alteration is investigated in calcium-enriched solutions. •New insights into silicon–calcium interactions in glass/cement systems are proposed. •Glass alteration is controlled by pH, Ca concentration and reaction progress. •Evidence of antagonist effects according to the importance of these parameters. -- Abstract: Numerous studies have been conducted on glass and cement durability in contact with water, but very little work to date has focused directly on interactions between the two materials. These interactions are mostly controlled by silicon–calcium reactivity. However, the physical and chemical processes involved remain insufficiently understood to predict the evolution of coupled glass–cement systems used in several industrial applications. Results are reported from borosilicate glass alteration in calcium-rich solutions. Our data show that four distinct behaviors can be expected according to the relative importance of three key parameters: the pH, the reaction progress (short- or long-term alteration) and the calcium concentration. Glass alteration is thus controlled by specific mechanisms depending on the solution chemistry: calcium complexation at the glass surface, precipitation of calcium silicate hydrates (C–S–H) or calcium incorporation in the altered layer. These findings highlight the impact of silicon–calcium interactions on glass durability and open the way for a better understanding of glass–cement mixing in civil engineering applications as well as in nuclear waste storage

  2. Time-Dependent Deformation Modelling for a Chopped-Glass Fiber Composite for Automotive Durability Design Criteria; FINAL

    International Nuclear Information System (INIS)

    Ren, W

    2001-01-01

    Time-dependent deformation behavior of a polymeric composite with chopped-glass-fiber reinforcement was investigated for automotive applications, The material under stress was exposed to representative automobile service environments. Results show that environment has substantial effects on time-dependent deformation behavior of the material. The data were analyzed and experimentally-based models developed for the time-dependent deformation behavior as a basis for automotive structural durability design criteria

  3. Molecular Environmental Science Using Synchrotron Radiation: Chemistry and Physics of Waste Form Materials. Final Report

    International Nuclear Information System (INIS)

    Lindle, Dennis W.

    2011-01-01

    Production of defense-related nuclear materials has generated large volumes of complex chemical wastes containing a mixture of radionuclides. The disposition of these wastes requires conversion of the liquid and solid-phase components into durable, solid forms suitable for long-term immobilization. Specially formulated glass compositions and ceramics such as pyrochlores and apatites are the main candidates for these wastes. An important consideration linked to the durability of waste-form materials is the local structure around the waste components. Equally important is the local structure of constituents of the glass and ceramic host matrix. Knowledge of the structure in the waste-form host matrices is essential, prior to and subsequent to waste incorporation, to evaluate and develop improved waste-form compositions based on scientific considerations. This project used the soft-x-ray synchrotron-radiation-based technique of near-edge x-ray-absorption fine structure (NEXAFS) as a unique method for investigating oxidation states and structures of low-Z elemental constituents forming the backbones of glass and ceramic host matrices for waste-form materials. In addition, light metal ions in ceramic hosts, such as titanium, are also ideal for investigation by NEXAFS in the soft-x-ray region. Thus, one of the main objectives was to understand outstanding issues in waste-form science via NEXAFS investigations and to translate this understanding into better waste-form materials, followed by eventual capability to investigate 'real' waste-form materials by the same methodology. We conducted several detailed structural investigations of both pyrochlore ceramic and borosilicate-glass materials during the project and developed improved capabilities at Beamline 6.3.1 of the Advanced Light Source (ALS) to perform the studies.

  4. High-level nuclear waste borosilicate glass: A compendium of characteristics

    International Nuclear Information System (INIS)

    Cunnane, J.C.; Bates, J.K.; Ebert, W.L.; Feng, X.; Mazer, J.J.; Wronkiewicz, D.J.; Sproull, J.; Bourcier, W.L.; McGrail, B.P.

    1992-01-01

    With the imminent startup, in the United States, of facilities for vitrification of high-level nuclear waste, a document has been prepared that compiles the scientific basis for understanding the alteration of the waste glass products under the range of service conditions to which they may be exposed during storage, transportation, and eventual geologic disposal. A summary of selected parts of the content of this document is provided. Waste glass alterations in a geologic repository may include corrosion of the glass network due to groundwater and/or water vapor contact. Experimental testing results are described and interpreted in terms of the underlying chemical reactions and physical processes involved. The status of mechanistic modeling, which can be used for long-term predictions, is described and the remaining uncertainties associated with long-term simulations are summarized

  5. Monitoring the Durability Performance of Concrete in Nuclear Waste Containment. Technical Progress Report No. 3

    International Nuclear Information System (INIS)

    Ulm, Franz-Josef

    2000-01-01

    OAK-B135 Monitoring the Durability Performance of Concrete in Nuclear Waste Containment. Technical Progress Report No. 3(NOTE: Part II A item 1 indicates ''PAPER'', but a report is attached electronically)

  6. Ceramic Single Phase High-Level Nuclear Waste Forms: Hollandite, Perovskite, and Pyrochlore

    Science.gov (United States)

    Vetter, M.; Wang, J.

    2017-12-01

    The lack of viable options for the safe, reliable, and long-term storage of nuclear waste is one of the primary roadblocks of nuclear energy's sustainable future. The method being researched is the incorporation and immobilization of harmful radionuclides (Cs, Sr, Actinides, and Lanthanides) into the structure of glasses and ceramics. Borosilicate glasses are the main waste form that is accepted and used by today's nuclear industry, but they aren't the most efficient in terms of waste loading, and durability is still not fully understood. Synroc-phase ceramics (i.e. hollandite, perovskite, pyrochlore, zirconolite) have many attractive qualities that glass waste forms do not: high waste loading, moderate thermal expansion and conductivity, high chemical durability, and high radiation stability. The only downside to ceramics is that they are more complex to process than glass. New compositions can be discovered by using an Artificial Neural Network (ANN) to have more options to optimize the composition, loading for performance by analyzing the non-linear relationships between ionic radii, electronegativity, channel size, and a mineral's ability to incorporate radionuclides into its structure. Cesium can be incorporated into hollandite's A-site, while pyrochlore and perovskite can incorporate actinides and lanthanides into their A-site. The ANN is used to predict new compositions based on hollandite's channel size, as well as the A-O bond distances of pyrochlore and perovskite, and determine which ions can be incorporated. These new compositions will provide more options for more experiments to potentially improve chemical and thermodynamic properties, as well as increased waste loading capabilities.

  7. Colloid formation during waste glass corrosion

    International Nuclear Information System (INIS)

    Mertz, C.J.; Buck, E.C.; Fortner, J.A.; Bates, J.K.

    1996-01-01

    The long-term behavior of nuclear waste glass in a geologic repository may require a technical consideration of the role of colloids in the release and transport of radionuclides. The neglect of colloidal properties in assessing the near- and far-field migration behavior of actinides may lead to significant underestimates and poor predictions of biosphere exposure from high-level waste (HLW) disposal. Existing data on colloid-facilitated transport suggests that radionuclide migration may be enhanced, but the importance of colloids is not adequately assessed. Indeed, the occurrence of radionuclide transport, attributed to colloidal species, has been reported at Mortandad Canyon, Los Alamos and at the Nevada Test Site; both unsaturated regions are similar to the proposed HLW repository at Yucca Mountain. Although some developments have been made on understanding the transport characteristics of colloids, the characterization of colloids generated from the corrosion of the waste form has been limited. Colloids are known to incorporate radionuclides either from hydrolysis of dissolved species (real colloids) or from adsorption of dissolved species onto existing groundwater colloids (pseudocolloids); however, these colloids may be considered secondary and solubility limited when compared to the colloids generated during glass alteration

  8. Nuclear waste glass melter design including the power and control systems

    International Nuclear Information System (INIS)

    Chapman, C.C.

    1982-01-01

    An energy balance of a joule-heated nuclear waste glass melter is used to discuss the problems in the design of the melter geometry and in the specifications of the power and control systems. The relationships between geometry, electrode current density, production rate, load voltage, and load power are presented graphically. The influence of liquid feeding on the surface of the glass and the variability of nuclear waste glass on the design and control during operation is discussed. 10 refs

  9. Assessment of water/glass interactions in waste glass melter operation

    International Nuclear Information System (INIS)

    Postma, A.K.; Chapman, C.C.; Buelt, J.L.

    1980-04-01

    A study was made to assess the possibility of a vapor explosion in a liquid-fed glass melter and during off-standard conditions for other vitrification processes. The glass melter considered is one designed for the vitrification of high-level nuclear wastes and is comprised of a ceramic-lined cavity with electrodes for joule heating and processing equipment required to add feed and withdraw glass. Vapor explosions needed to be considered because experience in other industrial processes has shown that violent interactions can occur if a hot liquid is mixed with a cooler, vaporizable liquid. Available experimental evidence and theoretical analyses indicate that destructive glass/water interactions are low probability events, if they are possible at all. Under standard conditions, aspects of liquid-fed melter operation which work against explosive interactions include: (1) the aqueous feed is near its boiling point; (2) the feed contains high concentrations of suspended particles; (3) molten glass has high viscosity (greater than 20 poise); and (4) the glass solidifies before film boiling can collapse. While it was concluded that vapor explosions are not expected in a liquid-fed melter, available information does not allow them to be ruled out altogether. Several precautionary measures which are easily incorporated into melter operation procedures were identified and additional experiments were recommended

  10. Zero photoelastic and water durable ZnO–SnO–P2O5–B2O3 glasses

    Directory of Open Access Journals (Sweden)

    Akira Saitoh

    2015-04-01

    Full Text Available We report properties of zero birefringent xZnO–(67–xSnO–(33–yP2O5–y B2O3 glasses, within 18.5 ≤ x ≤ 22 and y = 0, 3, and 10 mol. %. These compositions of boro-phosphate glasses provide both zero photoelastic constant (PEC and improved water durability. x = 19 and y = 3 compositions show minimum PEC of −0.002 × 10−12 Pa−1, which can contribute to candidate material for fiber current sensor devise without lead. The structures of zero photoelastic glasses were investigated by Raman scattering and nuclear magnetic resonance spectroscopies. Compositions of zero PEC glasses are explained by the empirical model proposed by Zwanziger et al. [Chem. Mater. 19, 286-290 (2007].

  11. Use of waste glass in highway construction (update--1992).

    Science.gov (United States)

    1993-01-01

    Increasing pressures to recycle more wastes and minimize the amount of materials placed in landfills are forcing reconsideration of potential uses of waste glass in highway construction and maintenance operations. The federal government and many stat...

  12. Processing constraints on high-level nuclear waste glasses for Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Hrma, P.R.

    1993-09-01

    The work presented in this paper is a part of a major technology program supported by the U.S. Department of Energy (DOE) in preparation for the planned operation of the Hanford Waste Vitrification Plant (HWVP). Because composition of Hanford waste varies greatly, processability is a major concern for successful vitrification. This paper briefly surveys general aspects of waste glass processability and then discusses their ramifications for specific examples of Hanford waste streams

  13. Effects of heavy weight waste glass recycled as fine aggregate on the mechanical properties of mortar specimens

    International Nuclear Information System (INIS)

    Choi, So Yeong; Choi, Yoon Suk; Yang, Eun Ik

    2017-01-01

    Highlights: • The properties of mortar used heavy weight waste glass as fine aggregate were compared. • Unit volume weight and shielding performance increased with the content of waste glass. • However, the strength decreased as the waste glass substitution increased. • The waste glass substitution affected on pores ranging from 10–100 nm. - Abstract: The quantities of heavy weight waste glass have increased over time due to rapid industrialization and changes in the quality of life. Moreover, most of this waste is not recycled. Concrete is the most widely used construction material, the huge amounts of natural resources are required to make concrete. Therefore, it is necessary to investigate the possibility of recycling of heavy weight waste glass as an ingredient in the manufacturing of concrete. In this study, the suitability of heavy weight waste glass as a fine aggregate material is considered. The results of flow test, unit volume weight, radiation shielding performance, compressive strength, flexural strength, and micropore and macropore distribution of mortar are compared and evaluated. It was found that when the heavy weight waste glass substitution ratio increases, the fluidity, unit volume weight and radiation shielding performance also increase. However, the compressive and flexural strength of mortar gradually decrease with an increase in the substitution ratio of heavy weight waste glass. Moreover, the micro pore size distribution is significantly affected by the substitution of heavy weight waste glass.

  14. Durability of incinerator ash waste encapsulated in modified sulfur cement

    International Nuclear Information System (INIS)

    Kalb, P.D.; Heiser, J.H. III; Pietrzak, R.; Colombo, P.

    1991-01-01

    Waste form stability under anticipated disposal conditions is an important consideration for ensuring continued isolation of contaminants from the accessible environment. Modified sulfur cement is a relatively new material and has only recently been applied as a binder for encapsulation of mixed wastes. Little data are available concerning its long-term durability. Therefore, a series of property evaluation tests for both binder and waste-binder combinations have been conducted to examine potential waste form performance under storage and disposal conditions. These tests include compressive strength, biodegradation, radiation stability, water immersion, thermal cycling, and leaching. Waste form compressive strength increased with ash waste loadings to 30.5 MPa at a maximum incinerator ash loading of 43 wt %. Biodegradation testing resulted in no visible microbial growth of either bacteria or fungi. Initial radiation stability testing did not reveal statistically significant deterioration in structural integrity. Results of 90 day water immersion tests were dependent on the type of ash tested. There were no statistically significant changes in compressive strength detected after completion of thermal cycle testing. Radionuclides from ash waste encapsulated in modified sulfur cement leached between 5 and 8 orders of magnitude slower than the leach index criterion established by the Nuclear Regulatory Commission (NRC) for low-level radioactive waste. Modified sulfur cement waste forms containing up to 43 wt % incinerator fly ash passed EPA Toxicity Characteristic Leaching Procedure (TCLP) criteria for lead and cadmium leachability. 11 refs., 2 figs., 5 tabs

  15. Evaluation the microwave heating of spinel crystals in high-level waste glass

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Ecology Lab. (SREL); Washington, A. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Ecology Lab. (SREL)

    2015-08-18

    In this report, the microwave heating of a crystal-free and a partially (24 wt%) trevorite-crystallized waste glass simulant were evaluated. The results show that a 500 mg piece of partially crystallized waste glass can be heated from room-temperature to above 1600 °C (as measured by infrared radiometry) within 2 minutes using a single mode, highly focused, 2.45 GHz microwave, operating at 300 W. X-ray diffraction measurements show that the partially crystallized glass experiences an 87 % reduction in trevorite following irradiation and thermal quenching. When a crystal-free analogue of the same waste glass simulant composition is exposed to the same microwave radiation it could not be heated above 450 °C regardless of the heating time.

  16. Control of DWPF [Defense Waste Processing Facility] melter feed composition

    International Nuclear Information System (INIS)

    Edwards, R.E. Jr.; Brown, K.G.; Postles, R.L.

    1990-01-01

    The Defense Waste Processing Facility will be used to immobilize Savannah River Site high-level waste into a stable borosilicate glass for disposal in a geologic repository. Proper control of the melter feed composition in this facility is essential to the production of glass which meets product durability constraints dictated by repository regulations and facility processing constraints dictated by melter design. A technique has been developed which utilizes glass property models to determine acceptable processing regions based on the multiple constraints imposed on the glass product and to display these regions graphically. This system along with the batch simulation of the process is being used to form the basis for the statistical process control system for the facility. 13 refs., 3 figs., 1 tab

  17. Investigation on Stress-Rupture Behavior of a Chopped-Glass-Fiber Composite for Automotive Durability Design Criteria

    Energy Technology Data Exchange (ETDEWEB)

    Ren, W

    2001-08-24

    Practical and inexpensive testing methods were developed to investigate stress-rupture properties of a polymeric composite with chopped glass fiber reinforcement for automotive applications. The material was tested in representative automotive environments to generate experimental data. The results indicate that environments have substantial effects on the stress-rupture behavior. The data were analyzed and developed into stress-rupture design criteria to address one of the durability aspects of the material for automotive structural applications.

  18. Glass formation, properties, and structure of soda-yttria-silicate glasses

    Science.gov (United States)

    Angel, Paul W.; Hann, Raiford E.

    1991-01-01

    The glass formation region of the soda yttria silicate system was determined. The glasses within this region were measured to have a density of 2.4 to 3.1 g/cu cm, a refractive index of 1.50 to 1.60, a coefficient of thermal expansion of 7 x 10(exp -6)/C, softening temperatures between 500 and 780 C, and Vickers hardness values of 3.7 to 5.8 GPa. Aqueous chemical durability measurements were made on select glass compositions while infrared transmission spectra were used to study the glass structure and its effect on glass properties. A compositional region was identified which exhibited high thermal expansion, high softening temperatures, and good chemical durability.

  19. Glass formation, properties and structure of soda-yttria-silica glasses

    Science.gov (United States)

    Angel, Paul W.; Hann, Raiford E.

    1992-01-01

    The glass formation region of the soda yttria silicate system was determined. The glasses within this region were measured to have a density of 2.4 to 3.1 g/cu cm, a refractive index of 1.50 to 1.60, a coefficient of thermal expansion of 7 x 10(exp -6)/C, softening temperatures between 500 and 780 C, and Vickers hardness values of 3.7 to 5.8 GPa. Aqueous chemical durability measurements were made on select glass compositions while infrared transmission spectra were used to study the glass structure and its effect on glass properties. A compositional region was identified which exhibited high thermal expansion, high softening temperatures, and good chemical durability.

  20. A review and discussion of candidate ceramics for immobilization of high-level fuel reprocessing wastes

    International Nuclear Information System (INIS)

    Hayward, P.J.

    1982-08-01

    This review discusses and attempts to evaluate 11 of the leading ceramic processes for hosting the high-level and high-level plus medium-level wastes which would arise from the reprocessing of used UO 2 , (Th,Pu)O 2 and (Th,U)O 2 fuels. The wasteform materials considered include glass ceramics, supercalcine ceramics, SYNROC ceramics, 'stuffed glass', titanate ceramics, cermets, clay ceramics, cement-based materials and multibarrier wasteforms. Although no attempt has been made to rank these candidates in order of superiority, the conclusion is drawn that, of the materials proposed so far, a glass ceramic appears to be best suited to the Canadian program, taking into account durability in the potential environment of a flooded vault, ability to withstand radiation and transmutation damage without serious loss of durability, ability to accommodate variable waste compositions, and ease of processing and quality control. This conclusion does not necessarily apply to other national waste management programs. However, many of the points raised might be included in any critical assessment of alternative wasteform materials

  1. Evaluation on the long-term durability and leachability of cemented waste form

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Hong; Lee, Jae Won; Ryue, Young Gerl [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-08-01

    The long-term durability and leachability on the cemented waste form containing boric acids produced in domestic nuclear nuclear power plants were evaluated. Compressive strength of waste form after durability test such as thermal stability and water immersion cycle was higher than before test and consistently increased with increasing of test time in range of 83 of 286 kgf/cm{sup 2}. Long-term leachability was evaluated by standard test methods, leach affecting factors, and prediction of long-term leachability with result data of short-term leach test. In all leach tests, the release of Cs-137 was controlled by diffusion, whereas release of Co-60 was not controlled by diffusion. Leach rate of Cs-137 was relatively constant at standard leach test methods such as NAS 16.1, IAEA, ISO-6961, and MCC-1, but that of Co-60 increased with leachant-renewal frequencies. The leach rate of both Cs-137 and Co-60 increased as test temperature raised. The release of Co-137 decreased in simulated seawater as leachant, but increased with increasing leachant volume. The prediciton of long-term release of Cs-137 from large-scale waste form using the results from short-term leach test of small-scale waste form were within {+-} 5% of actual release. The leachability indexes of Cs-137 were between 6.5 and 7.5 and those of Co-60 were ranged from 11.6 to 13.3, increasing as cumulative leaching time increased. (author). 22 refs., 14 figs., 15 tabs.

  2. Molecular glasses for nuclear waste encapsulation

    International Nuclear Information System (INIS)

    Ropp, R.C.

    1982-01-01

    The use of a molecular glass based upon a polymerized phosphate of aluminum (PAP), indium or gallium overcomes all of the prior objections to use of glass as a high-level nuclear waste (HLW) encapsulation agent. This HLW glass product could not be made to devitrify, dissolved all of the oxides found in calcine, including the difficultly soluble ones, did not form microcrystallites in the melt or subsequent glass-casting, and possessed a hydrolytic etching rate to boiling water even lower than that of HLW-ZBS glass. A precursor compound, M(H 2 PO 4 ) 3 , is prepared, where M is a trivalent metal selected from the group consisting of aluminum, indium and gallium. The impurity level is carefully controlled so as not to exceed 300 ppm total. The precursor crystals may be washed to remove excess phosphoric acid as desired. HLW is added to the crystals and the mixture is then heated at a controlled heating rate to induce solid state polymerization and to form a melt at 1350 degrees C in which the HLW oxides dissolve rapidly

  3. On the Durability of Nuclear Waste Forms from the Perspective of Long-Term Geologic Repository Performance

    Directory of Open Access Journals (Sweden)

    Yifeng Wang

    2013-12-01

    Full Text Available High solid/water ratios and slow water percolation cause the water in a repository to quickly (on a repository time scale reach radionuclide solubility controlled by the equilibrium with alteration products; the total release of radionuclides then becomes insensitive to the dissolution rates of primary waste forms. It is therefore suggested that future waste form development be focused on conditioning waste forms or repository environments to minimize radionuclide solubility, rather than on marginally improving the durability of primary waste forms.

  4. Analyses of SRS waste glass buried in granite in Sweden and salt in the United States

    International Nuclear Information System (INIS)

    Williams, J.P.; Wicks, G.G.; Clark, D.E.; Lodding, A.R.

    1991-01-01

    Simulated Savannah River Site (SRS) waste glass forms have been buried in the granite geology of the Stirpa mine in Sweden for two years. Analyses of glass surfaces provided a measure of the performance of the waste glasses as a function of time. Similar SRS waste glass compositions have also been buried in salt at the WIPP facility in Carlsbad, New Mexico for a similar time period. Analyses of the SRS waste glasses buried in-situ in granite will be presented and compared to the performance of these same compositions buried in salt at WIPP

  5. The incorporation of technetium into a representative low-activity waste glass

    International Nuclear Information System (INIS)

    Ebert, W.L.; Bakel, A.J.; Bowers, D.L.; Buck, E.C.; Emery, J.W.

    1997-01-01

    A glass that has been tested to understand the corrosion behavior of waste glasses with high soda contents for immobilizing Hanford incidental wastes has been made by melting crushed glass with either TcO 2 or NaTcO 4 at 1,100--1,300 C. Incorporation of technetium in the glass was affected by solubility or kinetic effects. Metallic technetium inclusions formed in all the TcO 2 -doped glasses. Inclusions also formed in glasses with added NaTcO 4 that were melted at 1,100 C, but a glass melted at 1,200 C did not contain detectable inclusions. The presence of Tc-bearing inclusions complicates the interpretation of results from dissolution tests because of the simultaneous release of technetium from more than one phase, the unknown surface areas of each phase, and the possible incorporation of technetium that is released from one phase into another phase. A glass containing about 0.15 mass % Tc dissolved in the glass is being used in dissolution tests to study the release behavior of technetium

  6. Advanced waste form and Melter development for treatment of troublesome high-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (also with high Al2O3 concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both

  7. Glasses used for the high level radioactive wastes storage

    International Nuclear Information System (INIS)

    Sombret, C.

    1983-06-01

    High level radioactive wastes generated by the reprocessing of spent fuels is an important concern in the conditioning of radioactive wastes. This paper deals with the status of the knowledge about glasses used for the treatment of these liquids [fr

  8. Office of River Protection Advanced Low-Activity Waste Glass Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    Peeler, David K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kim, Dong-Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vienna, John D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schweiger, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Piepel, Gregory F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-11-01

    The U.S. Department of Energy Office of River Protection (ORP) has initiated and leads an integrated Advanced Waste Glass (AWG) program to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product performance requirements. The integrated ORP program is focused on providing a technical, science-based foundation for making key decisions regarding the successful operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) facilities in the context of an optimized River Protection Project (RPP) flowsheet. The fundamental data stemming from this program will support development of advanced glass formulations, key product performance and process control models, and tactical processing strategies to ensure safe and successful operations for both the low-activity waste (LAW) and high-level waste vitrification facilities. These activities will be conducted with the objective of improving the overall RPP mission by enhancing flexibility and reducing cost and schedule. The purpose of this advanced LAW glass research and development plan is to identify the near-term, mid-term, and longer-term research and development activities required to develop and validate advanced LAW glasses, property-composition models and their uncertainties, and an advanced glass algorithm to support WTP facility operations, including both Direct Feed LAW and full pretreatment flowsheets. Data are needed to develop, validate, and implement 1) new glass property-composition models and 2) a new glass formulation algorithm. Hence, this plan integrates specific studies associated with increasing the Na2O and SO3/halide concentrations in glass, because these components will ultimately dictate waste loadings for LAW vitrification. Of equal importance is the development of an efficient and economic strategy for 99Tc management. Specific and detailed studies are being implemented to understand the fate of Tc throughout

  9. Heterogeneities in nuclear waste glass

    International Nuclear Information System (INIS)

    Ladirat, Ch.

    1997-01-01

    The industrial vitrification of high level radioactive wastes is a 2 stage process. During the first stage, the concentrated solution is heated in a spinning resistance oven at the temperature of 400 Celsius degrees till evaporation and calcination. The second stage begins when the dry residue falls into a melting pot that is maintained at a temperature of 1100-1150 Celsius degrees. Glass fretting is added and the glass is elaborated through the fusion of the different elements present in the melting pot. Heterogeneities in the glass may be associated to: - the presence in the solution to vitrify of insoluble elements from the dissolution of the fuel (RuO 2 , Rh, Pd), - the presence of minuscule metal scraps (Zr) that have been produced during the cutting of the fuel element, - the failures to conform to the technical specifications of the vitrification process, for instance, temperatures or flow rates when introducing the different elements in the melting pot. (A.C.)

  10. Glass composite waste forms for iodine confined in bismuth-embedded SBA-15

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Jae Hwan [Nuclear Fuel Cycle Process Development Division, Korea Atomic Energy Research Institute, 989-111 Daeduk-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of); Park, Hwan Seo; Ahn, Do-Hee [Nuclear Fuel Cycle Process Development Division, Korea Atomic Energy Research Institute, 989-111 Daeduk-daero, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of); Yim, Man-Sung, E-mail: msyim@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 291 Daehak-ro, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

    2016-11-15

    The aim of this study was to stabilize bismuth-embedded SBA-15 that captured iodine gas by fabrication of monolithic waste forms. The iodine containing waste was mixed with Bi{sub 2}O{sub 3} (a stabilizing additive) and low-temperature sintering glass followed by pelletizing and the sintering process to produce glass composite materials. Iodine volatility during the sintering process was significantly affected by the ratio of Bi{sub 2}O{sub 3} and the glass composition. It was confirmed that BiI{sub 3}, the main iodine phase within bismuth-embedded SBA-15, was effectively transformed to the mixed phases of Bi{sub 5}O{sub 7}I and BiOI. The initial leaching rates of iodine from the glass composite waste forms ranged 10{sup −3}–10{sup −2} g/m{sup 2} day, showing the stability of the iodine phases encapsulated by the glassy networks. It was also observed that common groundwater anions (e.g., chloride, carbonate, sulfite, and fluoride) elevated the iodine leaching rate by anion exchange reactions. The present results suggest that the glass composite waste form of bismuth-embedded SBA-15 could be a candidate material for stable storage of {sup 129}I. - Highlights: • Glass composite waste forms were developed to stabilize iodine confined in Bi-embedded SBA-15. • BiI{sub 3} within Bi-embedded SBA-15 was transformed to BiOI and Bi{sub 5}O{sub 7}I during sintering process. • Iodine volatility was significantly affected by glass composition and Bi{sub 2}O{sub 3} additive. • Iodine leaching rates were 10{sup −3}–10{sup −2} g/m{sup 2} day due to the stable iodine phases encapsulated by glassy networks. • Glass composite waste form of Bi-embedded SBA-15 is expected to be a candidate material for stable storage of {sup 129}I.

  11. Production and characterization of red mud based on glasses for the immobilization of nuclear wastes

    International Nuclear Information System (INIS)

    Vieira, Heveline

    2015-01-01

    Glasses based on red mud, a residual material from bauxite processing, were developed and characterized in this work. In order to promote its use, a minimum 60 wt% of red mud was used in the production of the glasses. According to XRD results, materials containing considerable amorphous phases were produced when using red mud as raw material. These amorphous phases were observed even though crystalline phases associated to Fe coming from the red mud itself were present. The material denominated 60L40S, which has a nominal composition of 60 wt% red mud showed the best properties comparing with the others compositions studied. However, these materials presented a high melting temperature. Changes in the composition of this material were made with the objective of lowering this temperature. Results indicated that the changes made to the material were successful in the reduction of the melting temperature. However, a reduction in the chemical properties of the resulting material was observed. Elements usually found in the chemical composition of nuclear wastes were added to the glasses produced. It was done with the objective of determining the effect of these elements on the chemical and physical properties of the red mud based glasses obtained. It was found that it was possible to add up to 15 wt% of these elements to the materials produced. The addition of these simulant materials promoted a reduction in the melting temperature of the resulting material. Above 15 wt%, the added elements precipitate in the structure of the resulting material. Even though the reduction in the chemical durability of the 60L40S material when simulant elements were added, it was observed that this material contained the simulant elements confined in its structure when in contact with water. This is a promising result, since it indicates that the 60L40S has the potential to immobilize elements from nuclear wastes . (author)

  12. Joule-heated glass-furnace system for the incineration of low-level radioactive wastes

    International Nuclear Information System (INIS)

    Armstrong, K.M.; Klingler, L.M.; Doty, J.W.; Kramer, D.P.

    1982-01-01

    For the past 1-1/2 years, Mound has been preparing and evaluating a commercially available joule-heated glass furnace unit, coupled with a wet scrubbing system. The purpose of the glass furnace evaluation is to advance and document incinerator technology for such combustibles as solids, resins, and sludges, and to develop a stable waste form for subsequent disposal. Four (4) waste nonradioactive types were selected to determine the combustion efficiency of the furnace unit: (1) dry solid waste composed of paper, plastics, rubber, and cloth, (2) ion exchange resin of both the anionic and cationic type, (3) filter sludge composed of diatomaceous earth, organic cellulosic filter aid, and powdered ion exchange resin, and (4) cartridge filters having glass and plastic filter surfaces and nonmetallic cores. When completed, the combustion efficiency experiments for the proposed nonradioactive waste-types revealed the ability of the furnace to easily incinerate waste at feedrates of up to 150 lb/hr. During the course of the experiments, combustibles in the offgas remained consistently low, suggesting excellent combustion efficiency. Furthermore, ash produced by the combustion process was effectively incorporated into the melt by convective currents in the glass. Future work on the glass furnace incinerator will include spiking the waste to determine radioisotope behavior in the furnace

  13. Development of Models to Predict the Redox State of Nuclear Waste Containment Glass

    Energy Technology Data Exchange (ETDEWEB)

    Pinet, O.; Guirat, R.; Advocat, T. [Commissariat a l' Energie Atomique (CEA), Departement de Traitement et de Conditionnement des Dechets, Marcoule, BP 71171, 30207 Bagnols-sur-Ceze Cedex (France); Phalippou, J. [Universite de Montpellier II, Laboratoire des Colloides, Verres et Nanomateriaux, 34095 Montpellier Cedex 5 (France)

    2008-07-01

    Vitrification is one of the recommended immobilization routes for nuclear waste, and is currently implemented at industrial scale in several countries, notably for high-level waste. To optimize nuclear waste vitrification, research is conducted to specify suitable glass formulations and develop more effective processes. This research is based not only on experiments at laboratory or technological scale, but also on computer models. Vitrified nuclear waste often contains several multi-valent species whose oxidation state can impact the properties of the melt and of the final glass; these include iron, cerium, ruthenium, manganese, chromium and nickel. Cea is therefore also developing models to predict the final glass redox state. Given the raw materials and production conditions, the model predicts the oxygen fugacity at equilibrium in the melt. It can also estimate the ratios between the oxidation states of the multi-valent species contained in the molten glass. The oxidizing or reductive nature of the atmosphere above the glass melt is also taken into account. Unlike the models used in the conventional glass industry based on empirical methods with a limited range of application, the models proposed are based on the thermodynamic properties of the redox species contained in the waste vitrification feed stream. The thermodynamic data on which the model is based concern the relationship between the glass redox state and the oxygen fugacity in the molten glass. The model predictions were compared with oxygen fugacity measurements for some fifty glasses. The experiments carried out at laboratory and industrial scale with a cold crucible melter. The oxygen fugacity of the glass samples was measured by electrochemical methods and compared with the predicted value. The differences between the predicted and measured oxygen fugacity values were generally less than 0.5 Log unit. (authors)

  14. Leaching of actinides from simulated nuclear waste glass

    International Nuclear Information System (INIS)

    Pickering, S.; Walker, C.T.; Offermann, P.

    1982-01-01

    Two types of simulated nuclear waste glass doped with actinides were leached at 200 0 C in distilled water and salt solutions. Am, Np, Pu and U were all preferentially retained in the surface layer on the glass. Leaching ratios of 0.1 to 0.2 for Np and approx. 0.02 for Am were measured. The losses of Am and Np to the leachant were proportional to the total weight loss of the glass and were larger at 10 ml leachant/cm 2 glass than at 5 ml/cm 2 . Weight loss from the glass occurred only at the start of the experiments for periods ranging from 10 h to 10 days according to leachant composition and volume. Wt losses from the C31-3-EC glass were much greater in saturated NaCl solution than in distilled water. Enrichment in the outer surface layer of Al or Ca according to glass type could be correlated with leachant pH, glass composition and weight loss measurements

  15. MIIT: International in-situ testing of nuclear waste glasses-performance of SRS simulated waste glass after 5 years of burial at the waste isolation pilot plant (WIPP)

    International Nuclear Information System (INIS)

    Wicks, G.G.; Lodding, A.R.; Macedo, P.B.; Clark, D.E.

    1993-01-01

    In July of 1986, the first in-situ test involving burial of simulated high-level waste [HLW] forms conducted in the United States was started. This program, called the Materials Interface Interactions Test or MIIT, comprises the largest, most cooperative field-testing venture in the international waste management community. Included in the study are over 900 waste form samples comprising 15 different systems supplied by seven nations. Also included are about 300 potential canister or overpack metal samples along with more than 500 geologic and backfill specimens. There are almost 2000 relevant interactions that characterize this effort which has been conducted in the bedded salt site at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico. The MIIT program represents a joint effort managed by the Savannah River Technology Center (SRTC) in Aiken, S.C., and Sandia National Laboratories (SNL) in Albuquerque, N.M.. and sponsored by the US Department of Energy. Involved in MIIT are participants from national and federal laboratories, universities, and representatives from laboratories in France, Germany, Canada, Belgium, Japan, Sweden, the United Kingdom, and the United States. In July of 1991, the experimental portion of the 5-year MIIT study was completed on schedule. During this time interval, many in-situ measurements were performed, thousands of brine analyses conducted, and hundreds of waste glass and package components exhumed and evaluated after 6 mo., 1 yr., 2 yr. and 5 yr. burial periods. Although analyses are still in progress, the performance of SRS waste glass based on all data currently available has been seen to be excellent thus far. Initial analyses and assessment of Savannah River (SR) waste glass after burial in WIPP at 90 degrees C for 5 years is presented

  16. Glass composite waste forms for iodine confined in bismuth-embedded SBA-15

    Science.gov (United States)

    Yang, Jae Hwan; Park, Hwan Seo; Ahn, Do-Hee; Yim, Man-Sung

    2016-11-01

    The aim of this study was to stabilize bismuth-embedded SBA-15 that captured iodine gas by fabrication of monolithic waste forms. The iodine containing waste was mixed with Bi2O3 (a stabilizing additive) and low-temperature sintering glass followed by pelletizing and the sintering process to produce glass composite materials. Iodine volatility during the sintering process was significantly affected by the ratio of Bi2O3 and the glass composition. It was confirmed that BiI3, the main iodine phase within bismuth-embedded SBA-15, was effectively transformed to the mixed phases of Bi5O7I and BiOI. The initial leaching rates of iodine from the glass composite waste forms ranged 10-3-10-2 g/m2 day, showing the stability of the iodine phases encapsulated by the glassy networks. It was also observed that common groundwater anions (e.g., chloride, carbonate, sulfite, and fluoride) elevated the iodine leaching rate by anion exchange reactions. The present results suggest that the glass composite waste form of bismuth-embedded SBA-15 could be a candidate material for stable storage of 129I.

  17. Study of Wettability of Clayey Ceramic and Fluorescent Lamp Glass Waste Powders

    Science.gov (United States)

    Morais, Alline Sardinha Cordeiro; Monteiro, Sergio Neves; Ribeiro, Sebastião; Sardinha, Leonardo Carneiro; Vieira, Carlos Maurício Fontes

    The glass tube of spent fluorescent lamps is contaminated with mercury, which might be a serious hazard in the case of conventional recycling by melting with other glasses. A possible solution could be its incorporation into a clay body to fabricate common fired ceramics such as bricks and tiles. The objective of this work is to characterize a type of fluorescent lamp glass waste to be incorporated into a clayey ceramic. The characterization was performed in terms of wettability tests to evaluate the interaction between the surface of the clayey ceramic and glass waste as a function of the firing temperature. The results showed that the contact angle decreased with increasing temperature, reaching a value of 79°, at a temperature of 1100°C, but not sufficient to completely wet the ceramic. However, compatible chemical composition and reduction of porosity by the flow of soft glass waste between the clay particles favor the consolidation of the ceramics structure above 900°C.

  18. Alternative design concept for the second Glass Waste Storage Building

    International Nuclear Information System (INIS)

    Rainisch, R.

    1992-10-01

    This document presents an alternative design concept for storing canisters filled with vitrified waste produced at the Defense Waste Processing Facility (DWPF). The existing Glass Waste Storage Building (GWSB1) has the capacity to store 2,262 canisters and is projected to be completely filled by the year 2000. Current plans for glass waste storage are based on constructing a second Glass Waste Storage Building (GWSB2) once the existing Glass Waste Storage Building (GWSB1) is filled to capacity. The GWSB2 project (Project S-2045) is to provide additional storage capacity for 2,262 canisters. This project was initiated with the issue of a basic data report on March 6, 1989. In response to the basic data report Bechtel National, Inc. (BNI) prepared a draft conceptual design report (CDR) for the GWSB2 project in April 1991. In May 1991 WSRC Systems Engineering issued a revised Functional Design Criteria (FDC), the Rev. I document has not yet been approved by DOE. This document proposes an alternative design for the conceptual design (CDR) completed in April 1991. In June 1992 Project Management Department authorized Systems Engineering to further develop the proposed alternative design. The proposed facility will have a storage capacity for 2,268 canisters and will meet DWPF interim storage requirements for a five-year period. This document contains: a description of the proposed facility; a cost estimate of the proposed design; a cost comparison between the proposed facility and the design outlined in the FDC/CDR; and an overall assessment of the alternative design as compared with the reference FDC/CDR design

  19. Effect of Na{sub 2}O on aqueous dissolution of nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Farooqi, Rahmat Ullah, E-mail: rufarooqi@live.com [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, 77 Cheongam-Ro, Nam-Gu, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Hrma, Pavel [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, 77 Cheongam-Ro, Nam-Gu, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Pacific Northwest National Laboratory, Richland, WA (United States)

    2017-04-15

    Sodium oxide is present in the majority of commercial and waste glasses as a viscosity-reducing component. In some nuclear waste glasses, its source is the waste itself. As such, it can limit the waste loading because of its deleterious effect on the resistance of the glass to attack by aqueous media. The maximum tolerable content of Na{sub 2}O in glass depends on the presence and concentration of components that interact with it. To assess the acceptability limits of Na{sub 2}O in the composition region of nuclear waste glasses, we formulated 11 baseline compositions by varying the content of oxides of Si, B, Al, Ca, Zr, and Li. In each of these compositions, we varied the Na{sub 2}O fraction from 8–16 mass% to 23–30 mass%. To each of 146 glasses thus formulated, we applied the seven-day Product Consistency Test (PCT) to determine normalized B and Na releases (r{sub i}, where i ≡ B or Na). Fitting approximation functions ln(r{sub i}/gm{sup −2}) = Σb{sub ij}g{sub j} to r{sub i} data (g{sub j} is the j-th component mass fraction and b{sub ij} the corresponding component coefficient), we showed that the r{sub B} (and, consequently, the initial glass alteration rate) was proportional to the glass component mass fractions in the order Al{sub 2}O{sub 3}glass structure would fall apart or beyond which a continuous nondurable phase would be separated. Specific examples are given to demonstrate restrictions imposed on the boundary of the composition region of acceptable glasses by the maximum allowable r{sub B} and by the melt viscosity required for glass melter operation. Finally, the role that PCT data may play in understanding the evolution of the glass alteration process is discussed.

  20. Formulation of portland composite cement using waste glass as a supplementary cementitious material

    Science.gov (United States)

    Manullang, Ria Julyana; Samadhi, Tjokorde Walmiki; Purbasari, Aprilina

    2017-09-01

    Utilization of waste glass in cement is an attractive options because of its pozzolanic behaviour and the market of glass-composite cement is potentially available. The objective of this research is to evaluate the formulation of waste glass as supplementary cementitious material (SCM) by an extreme vertices mixture experiment, in which clinker, waste glass and gypsum proportions are chosen as experimental variables. The composite cements were synthesized by mixing all of powder materials in jar mill. The compressive strength of the composite cement mortars after being cured for 28 days ranges between 229 to 268 kg/cm2. Composite cement mortars exhibit lower compressive strength than ordinary Portland cement (OPC) mortars but is still capable of meeting the SNI 15-7064-2004 standards. The highest compressive strength is obtained by shifting the cement blend composition to the direction of increasing clinker and gypsum proportions as well as reducing glass proportion. The lower compressive strength of composite cement is caused by expansion due to ettringite and ASR gel. Based on the experimental result, the composite cement containing 80% clinker, 15% glass and 5% gypsum has the highest compressive strength. As such, the preliminary technical feasibility of reuse of waste glass as SCM has been confirmed.