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Sample records for waste glass study

  1. 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

  2. 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), ...

  3. 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

  4. 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.

  5. 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.

  6. 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.

  7. 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

  8. Natural glass analogues to alteration of nuclear waste glass: A review and recommendations for further study

    International Nuclear Information System (INIS)

    McKenzie, W.F.

    1990-01-01

    The purpose of this report is to review previous work on the weathering of natural glasses; and to make recommendations for further work with respect to studying the alteration of natural glasses as it relates quantifying rates of dissolution. the first task was greatly simplified by the published papers of Jercinovic and Ewing (1987) and Byers, Jercinovic, and Ewing (1987). The second task is obviously the more difficult of the two and the author makes no claim of completeness in this regard. Glasses weather in the natural environment by reacting with aqueous solutions producing a rind of secondary solid phases. It had been proposed by some workers that the thickness of this rind is a function of the age of the glass and thus could be used to estimate glass dissolution rates. However, Jercinovic and Ewing (1987) point out that in general the rind thickness does not correlate with the age of the glass owing to the differences in time of contact with the solution compared to the actual age of the sample. It should be noted that the rate of glass dissolution is also a function of the composition of both the glass and the solution, and the temperature. Quantification of the effects of these parameters (as well as time of contact with the aqueous phase and flow rates) would thus permit a prediction of the consequences of glass-fluid interactions under varying environmental conditions. Defense high- level nuclear waste (DHLW), consisting primarily of liquid and sludge, will be encapsulated by and dispersed in a borosilicate glass before permanent storage in a HLW repository. This glass containing the DHLW serves to dilute the radionuclides and to retard their dispersion into the environment. 318 refs

  9. Natural glass analogues to alteration of nuclear waste glass: A review and recommendations for further study

    Energy Technology Data Exchange (ETDEWEB)

    McKenzie, W.F.

    1990-01-01

    The purpose of this report is to review previous work on the weathering of natural glasses; and to make recommendations for further work with respect to studying the alteration of natural glasses as it relates quantifying rates of dissolution. the first task was greatly simplified by the published papers of Jercinovic and Ewing (1987) and Byers, Jercinovic, and Ewing (1987). The second task is obviously the more difficult of the two and the author makes no claim of completeness in this regard. Glasses weather in the natural environment by reacting with aqueous solutions producing a rind of secondary solid phases. It had been proposed by some workers that the thickness of this rind is a function of the age of the glass and thus could be used to estimate glass dissolution rates. However, Jercinovic and Ewing (1987) point out that in general the rind thickness does not correlate with the age of the glass owing to the differences in time of contact with the solution compared to the actual age of the sample. It should be noted that the rate of glass dissolution is also a function of the composition of both the glass and the solution, and the temperature. Quantification of the effects of these parameters (as well as time of contact with the aqueous phase and flow rates) would thus permit a prediction of the consequences of glass-fluid interactions under varying environmental conditions. Defense high- level nuclear waste (DHLW), consisting primarily of liquid and sludge, will be encapsulated by and dispersed in a borosilicate glass before permanent storage in a HLW repository. This glass containing the DHLW serves to dilute the radionuclides and to retard their dispersion into the environment. 318 refs.

  10. 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

  11. Leach rate studies on glass containing actual radioactive waste

    International Nuclear Information System (INIS)

    Walker, D.D.; Wiley, J.R.; Dukes, M.D.; LeRoy, J.H.

    1980-01-01

    Borosilicate glass containing radioactive wastes from the Savannah River Plant have been leached for 900 days. The International Standards Organization's (ISO) static leach test procedure was used on glass buttons in various leachants. Leach rates based on 90 Sr and 137 Cs analyses were similar: 2 x 10 -8 to 3 x 10 -8 g/(cm 2 )(d) in distilled water, 1 x 10 -8 to 3 x 10 -7 g/(cm 2 )(d) in pH 7 buffer, 3 x 10 -7 to 7 x 10 -7 g/(cm 2 )(d) in pH 9 buffer, and 7 x 10 -6 to 8 x 10 -5 g/(cm 2 )(d) in pH 4 buffer. Rates based on Pu analyses were the same as above in distilled water and pH 9 buffer, but were lower by an order of magnitude in pH 4 and pH 7 buffers. Almost all leach rates remained constant between 200 and 900 days of leaching. Increasing the concentration of the buffering agents had no effect on the leach rates at pH 7 (phosphate) and pH 9 (carbonate), but dramatically increased the rates at pH 4 (acetate). Leach rates did not differ significantly between high aluminum and high iron waste glasses

  12. 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)

  13. 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.

  14. 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

  15. 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)

  16. 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)

  17. 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

  18. 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

  19. 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

  20. 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)

  1. 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

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  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. Structure study and properties of rare earth-rich glassed for the conditioning of nuclear waste

    International Nuclear Information System (INIS)

    Bardez, I.

    2004-11-01

    A new nuclear glass composition, able to immobilize highly radioactive liquid wastes from high burn-up UO 2 fuel, was established and its structure studied. The composition of the selected rare earth-rich glass is (molar %): 61.79 SiO 2 - 8.94 B 2 O 3 - 3.05 Al 2 O 3 - 14.41 Na 2 O - 6.32 CaO - 1.89 ZrO 2 - 3.60 RE 2 O 3 (with RE = La, Ce, Pr and Nd) The aim of this study was to determine the local environment of the rare earth in this glass and also to glean information about the effect of glass composition on the rare earth neighbouring (influence of Si, B, Al, Na and Ca contents). To this end, several series of glasses, prepared from the baseline glass, were studied by different characterisation methods such as EXAFS spectroscopy at the neodymium L III -edge, optical absorption spectroscopy, Raman spectroscopy and 29 Si, 27 Al and 11 B MAS-NMR. By coupling all the results obtained, several hypotheses about the nature of the rare earth neighbouring in the glass were proposed. (author)

  12. 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

  13. 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

  14. Wastes vitrification by plasma torch: study of a glass formulation compatible with a wide range of B wastes

    International Nuclear Information System (INIS)

    Poitou, S.; Richaud, D.; Fiquet, O.; Gramondi, P.; Massit, H.

    2001-01-01

    Within the context of radioactive waste management, CEA has equipped itself with a 'PLASMARC' device. The central element of this device is a plasma torch treatment furnace. It has been implemented and validated for the vitrification of low level radioactive wastes. Meanwhile, the plasma torch presents potentially interests for immobilizing under an inert form in vitreous matrices, B wastes which are generally divided and of complex chemical composition. The application of this process to this type of wastes has been studied here. The obtained results show that with the plasma torch it is possible to make glasses with a high amount of silicon and aluminium oxide and which are adapted to the treatment / packaging of the B wastes. (O.M.)

  15. 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)

  16. 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.

  17. 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

  18. Molecular dynamics study of a nuclear waste glass matrix with plutonium

    International Nuclear Information System (INIS)

    Meis, C.; Delaye, J.M.; Ghaleb, D.

    1999-01-01

    Molecular dynamics simulation techniques were applied to model the incorporation of plutonium in the French nuclear waste glass matrix. Born-Mayer-Huggins analytical potentials were established to characterize short-range interactions between Pu-O and Pu-Pu pairs; the potentials were fitted to the structural properties of plutonium dioxide in the light of a recent experimental study showing that plutonium is found as Pu(IV) in the glass. The transferability of the established potentials to the glass structure is discussed, and the potential parameters are further refined by molecular dynamics simulations in an aluminoborosilicate glass to obtain mean Pu-O interatomic distances and first-neighbor coordination numbers matching the experimental values as closely as possible. Previously published Born-Mayer-Huggins potentials supplemented by Stillinger-Weber three-body terms were used for oxygen-cation and cation-cation interactions. The difficulties encountered in establishing a Pu-O potential that provides satisfactory results in both oxides and glasses are also discussed

  19. 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

  20. 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

  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. X-ray spectrometric determination of glass content of melts incorporating radioactive waste: a feasibility study

    International Nuclear Information System (INIS)

    Slates, R.V.

    1978-09-01

    X-ray fluorescence spectrometry was evaluated for the determination of glass content and homogeneity of glass incorporating high-level radioactive waste. Accuracy and precision were determined for analyses of Al 2 O 3 , SiO 2 , CaO, TiO 2 , MnO, Fe 2 O 3 , and NiO in specimens of known composition. These specimens were prepared by fusing powdered glass with nonradioactive synthetic waste. Matrix effects of sodium on these analyses were specifically evaluated. X-ray fluorescence spectrometry was shown to be applicable to the proposed determinations by comparing the known glass contents of 14 glass waste compositions with those calculated from experimentally determined concentrations of SiO or TiO 2

  3. Structural and crystallisation study of a rare earth alumino borosilicate glass designed for nuclear waste confinement

    International Nuclear Information System (INIS)

    Quintas, A.

    2007-09-01

    This work is devoted to the study of a rare earth alumino borosilicate glass, which molar composition is 61,81 SiO 2 - 3,05 Al 2 O 3 - 8,94 B 2 O 3 - 14,41 Na 2 O - 6,33 CaO - 1,90 ZrO 2 - 3,56 Nd 2 O 3 , and envisaged for the immobilization of nuclear wastes originating from the reprocessing of high discharge burn up spent fuel. From a structural viewpoint, we investigated the role of the modifier cations on the arrangement of the glass network through different modifications of the glass composition: variation of the Na/Ca ratio and modification of the nature of the alkali and alkaline earth cations. The NMR and Raman spectroscopic techniques were useful to determine the distribution of modifier cations among the glass network and also to cast light on the competition phenomena occurring between alkali and alkaline earth cations for charge compensation of [AlO 4 ] - and [BO 4 ] - species. The neodymium local environment could be probed by optical absorption and EXAFS spectroscopies which enabled to better understand the insertion mode of Nd 3+ ions among the silicate domains of the glass network. Concerning the crystallization behavior we were interested in how the glass composition may influence the crystallization processes and especially the formation of the apatite phase of composition Ca 2 Nd 8 (SiO 4 ) 6 O 2 . In particular, this work underlined the important role of both alkaline earth and rare earth cations on the crystallization of the apatite phase. (author)

  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. XPS and ion beam scattering studies of leaching in simulated waste glass containing uranium

    International Nuclear Information System (INIS)

    Karim, D.P.; Pronko, P.P.; Marcuso, T.L.M.; Lam, D.J.; Paulikas, A.P.

    1980-01-01

    Glass samples (consisting of 2 mole % UO 3 dissolved in a number of complex borosilicate simulated waste glasses including Battelle 76-68) were leached for varying times in distilled water at 75 0 C. The glass surfaces were examined before and after leaching using x-ray photoemission spectroscopy and back-scattered ion beam profiling. Leached samples showed enhanced surface layer concentrations of several elements including uranium, titanium, zinc, iron and rare earths. An experiment involving the leaching of two glasses in the same vessel showed that the uranium surface enhancement is probably not due to redeposition from solution

  6. 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

  7. Study of the surface crystallization and resistance to dissolution of niobium phosphate glasses for nuclear waste immobilization

    International Nuclear Information System (INIS)

    Vieira, Heveline

    2008-01-01

    The surface crystallization and the dissolution rate of three phosphate glass compositions containing different amounts of niobium oxide were studied. The glasses were named Nb30, Nb37, and Nb44 according to the nominal content of niobium oxide in the glass composition. The three compositions were evaluated keeping the P 2 O 5 /K 2 O ratio constant and varying the amount of Nb 2 O 5 . These glasses were produced by melting appropriate chemical compounds at 1500 deg C for 0.5 hour. The crystalline phases which were nucleated on the glass surface after heat treatment were determined by X-ray diffraction. The crystalline structures depend on the amount of niobium oxide in the glass composition. The crystal morphologies were observed by using an optical microscope, and their characteristics are specific for each kind of crystalline phase. The crystal growth rate and the surface nuclei density were determined for each glass composition, and they depend on each crystalline phase nucleated on the surface. From the differential thermal analysis curves it was determined that the Nb44 glass containing 46.5 mol por cent of niobium oxide is the most thermally stable against crystallization when compared to the Nb30 and Nb37 glasses. According to the activation energies determined for crystal growth on the surface of each glass type, the Nb44 glass can also be considered the most resistant one against crystallization. The dissolution rate for the Nb44 glass after 14 days immersed in an aqueous solution with pH equals to 7 at 90 deg C is the lowest (9.0 x 10 -7 g. cm -2 . day -1 ) when compared to the other two glass compositions. The dissolution rates in acidic and neutral solutions of all studied glasses meet the international standards for materials which can be used in the immobilization of nuclear wastes. (author)

  8. X-ray absorption studies of chlorine valence and local environments in borosilicate waste glasses

    International Nuclear Information System (INIS)

    McKeown, David A.; Gan, Hao; Pegg, Ian L.; Stolte, W.C.; Demchenko, I.N.

    2011-01-01

    Chlorine (Cl) is a constituent of certain types of nuclear wastes and its presence can affect the physical and chemical properties of silicate melts and glasses developed for the immobilization of such wastes. Cl K-edge X-ray absorption spectra (XAS) were collected and analyzed to characterize the unknown Cl environments in borosilicate waste glass formulations, ranging in Cl-content from 0.23 to 0.94 wt.%. Both X-ray absorption near edge structure (XANES) and extended X-ray absorption fine structure (EXAFS) data for the glasses show trends dependent on calcium (Ca) content. Near-edge data for the Ca-rich glasses are most similar to the Cl XANES of CaCl 2 , where Cl - is coordinated to three Ca atoms, while the XANES for the Ca-poor glasses are more similar to the mineral davyne, where Cl is most commonly coordinated to two Ca in one site, as well as Cl and oxygen nearest-neighbors in other sites. With increasing Ca content in the glass, Cl XANES for the glasses approach that for CaCl 2 , indicating more Ca nearest-neighbors around Cl. Reliable structural information obtained from the EXAFS data for the glasses is limited, however, to Cl-Cl, Cl-O, and Cl-Na distances; Cl-Ca contributions could not be fit to the glass data, due to the narrow k-space range available for analysis. Structural models that best fit the glass EXAFS data include Cl-Cl, Cl-O, and Cl-Na correlations, where Cl-O and Cl-Na distances decrease by approximately 0.16 A as glass Ca content increases. XAS for the glasses indicates Cl - is found in multiple sites where most Cl-sites have Ca neighbors, with oxygen, and possibly, Na second-nearest neighbors. EXAFS analyses suggest that Cl-Cl environments may also exist in the glasses in minor amounts. These results are generally consistent with earlier findings for silicate glasses, where Cl - was associated with Ca 2+ and Na + in network modifier sites.

  9. A Study of the Closed-Loop Supply Chain Coordination on Waste Glass Bottles Recycling

    Directory of Open Access Journals (Sweden)

    Wenxue Ran

    2016-01-01

    Full Text Available The recycling of waste products can sharply save manufacturing cost and improve the economic efficiency and corporate-reputation. It also has a great effect on the environment and resources protection. In the management of the closed-loop supply chain, the recycling of waste products and decision-making on pricing often directly affect the supply and demand of products and the operation efficiency of supply chain. Therefore, first we take waste glass bottles as an example and establish a mathematical model to solve the profit of manufacturers and retailers solely. Then, we analyzed whole supply chain profit under a dual-channel recycling condition which is directly recycled by consumers or by retailers. Finally, we concluded that no matter what product’s price, quality, profit, or operational efficiency of supply chain is, the overall recycling is better than the single node recycling model. Based on the analysis, we developed a new model to coordinate the profit of manufacturers and retailers in the supply chain with revenue-sharing contract. A numerical study shows that this approach is applicable and effective.

  10. 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

  11. 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

  12. 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.

  13. 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)

  14. Cold crucible induction melter studies for making glass ceramic waste forms: A feasibility assessment

    International Nuclear Information System (INIS)

    Crum, Jarrod; Maio, Vince; McCloy, John; Scott, Clark; Riley, Brian; Benefiel, Brad; Vienna, John; Archibald, Kip; Rodriguez, Carmen; Rutledge, Veronica; Zhu, Zihua; Ryan, Joe; Olszta, Matthew

    2014-01-01

    Glass ceramics are being developed to immobilize fission products, separated from used nuclear fuel by aqueous reprocessing, into a stable waste form suitable for disposal in a geological repository. This work documents the glass ceramic formulation at bench scale and for a scaled melter test performed in a pilot-scale (∼1/4 scale) cold crucible induction melter (CCIM). Melt viscosity, electrical conductivity, and crystallization behavior upon cooling were measured on a small set of compositions to select a formulation for melter testing. Property measurements also identified a temperature range for melter operation and cooling profiles necessary to crystallize the targeted phases in the waste form. Bench scale and melter run results successfully demonstrate the processability of the glass ceramic using the CCIM melter technology

  15. 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

  16. 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

  17. 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...

  18. 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

  19. 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.

  20. 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.)

  1. 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

  2. Study of the incorporation of marble and granite wastes in the raw material to produce glass wool

    International Nuclear Information System (INIS)

    Rodrigues, Girley Ferreira; Junca, Eduardo; Telles, Victor Bridi; Espinosa, Denise Crocce Romano; Tenorio, Jorge Alberto Soares; Alves, Joner Oliveira

    2010-01-01

    The study aimed to characterize materials obtained from the melted mixture containing marble and granite wastes, and also chemical reagents. Using the characterization results was defined the feasibility of reuse of the marble and granite wastes, through the incorporation in the raw material to produce glass wool (a material with great consumer market as thermo-acoustic insulator). The batch was poured in a water-filled recipient and also in a Herty viscometer at temperatures of 1400, 1450 and 1500 °C. Samples of produced materials were characterized by morphology using Scanning Electron Microscopy, by atomic structure using X-ray Diffraction, and by thermal behavior using Differential Thermal Analysis. The total amount of marble and granite wastes can reach about 79% replacement in relation to the total weight of the raw material used in the glass wool production. (author)

  3. 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.

  4. 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.

  5. 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

  6. Study of powellite-rich glass-ceramics for nuclear waste immobilization

    International Nuclear Information System (INIS)

    Taurines, T.

    2012-01-01

    MoO 3 is poorly soluble in borosilicate glasses which can lead to the crystallization of undesired phases when its concentration or the charge load (minor actinides and fission products concentration) is too high. Crystallization control is needed to guarantee good immobilization properties. We studied powellite-rich glass-ceramics obtained from a simplified nuclear glass in the system SiO 2 - B 2 O 3 - Na 2 O - CaO - Al 2 O 3 - MoO 3 - RE 2 O 3 (RE = Gd, Eu, Nd) by various heat treatments. Rare earth elements (REE) were added as minor actinides surrogates and as spectroscopic probes. The influence of MoO 3 and RE 2 O 3 content on powellite (CaMoO 4 ) crystallization was investigated. Various glass-ceramics (similar residual glass + powellite) were obtained with large crystal size distributions. Phase separation due to molybdenum occurs during quenching when [MoO 3 ] ≥ 2.5 mol%. We showed that increasing the rare earth content can suppress the phase separation due to molybdenum but it leads to spinodal decomposition of the residual glass. Furthermore, we studied the effects of parent glass complexifying and the insertion of Gd 3+ ions into the powellite structure. In order to understand the influence of microstructure on evolutions under β-irradiation, we studied point defects creation and structural changes. We showed that the damage induced by electronic excitations in the glass-ceramics is driven by the damage in the residual glass. (author) [fr

  7. 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

  8. X-ray absorption and Raman spectroscopy studies of molybdenum environments in borosilicate waste glasses

    Science.gov (United States)

    McKeown, David A.; Gan, Hao; Pegg, Ian L.

    2017-05-01

    Mo-containing high-level nuclear waste borosilicate glasses were investigated as part of an effort to improve Mo loading while avoiding yellow phase crystallization. Previous work showed that additions of vanadium decrease yellow phase formation and increases Mo solubility. X-ray absorption spectroscopy (XAS) and Raman spectroscopy were used to characterize Mo environments in HLW borosilicate glasses and to investigate possible structural relationships between Mo and V. Mo XAS spectra for the glasses indicate isolated tetrahedral Mo6+O4 with Mo-O distances near 1.75 Å. V XANES indicate tetrahedral V5+O4 as the dominant species. Raman spectra show composition dependent trends, where Mo-O symmetrical stretch mode frequencies (ν1) are sensitive to the mix of alkali and alkaline earth cations, decreasing by up to 10 cm-1 for glasses that change from Li+ to Na+ as the dominant network-modifying species. This indicates that MoO4 tetrahedra are isolated from the borosilicate network and are surrounded, at least partly, by Na+ and Li+. Secondary ν1 frequency effects, with changes up to 7 cm-1, were also observed with increasing V2O5 and MoO3 content. These secondary trends may indicate MoO4-MoO4 and MoO4-VO4 clustering, suggesting that V additions may stabilize Mo in the matrix with respect to yellow phase formation.

  9. 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.

  10. 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

  11. Natural analogue of nuclear waste glass in a geologic formation. Study on long-term behavior of volcanic glass shards collected from drill cores

    International Nuclear Information System (INIS)

    Yoshikawa, Hideki; Yui, Mikazu; Futakuchi, Katsuhito; Hiroki, Minenari

    2005-01-01

    Alteration of the volcanic glass in geologic formation was investigated as one of the natural analog for a glass of high-level nuclear waste in geological disposal. We analyzed some volcanic glasses included in the core sample of the bore hole and estimated the history of its burying and observed its alteration using the polarizing microscope. Some information at the piling up temperature and the piling up time was collected. (author)

  12. 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

  13. 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

  14. 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

  15. 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.

  16. X-ray absorption and Raman spectroscopy studies of molybdenum environments in borosilicate waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    McKeown, David A., E-mail: davidm@vsl.cua.edu; Gan, Hao; Pegg, Ian L.

    2017-05-15

    Mo-containing high-level nuclear waste borosilicate glasses were investigated as part of an effort to improve Mo loading while avoiding yellow phase crystallization. Previous work showed that additions of vanadium decrease yellow phase formation and increases Mo solubility. X-ray absorption spectroscopy (XAS) and Raman spectroscopy were used to characterize Mo environments in HLW borosilicate glasses and to investigate possible structural relationships between Mo and V. Mo XAS spectra for the glasses indicate isolated tetrahedral Mo{sup 6+}O{sub 4} with Mo-O distances near 1.75 Å. V XANES indicate tetrahedral V{sup 5+}O{sub 4} as the dominant species. Raman spectra show composition dependent trends, where Mo-O symmetrical stretch mode frequencies (ν{sub 1}) are sensitive to the mix of alkali and alkaline earth cations, decreasing by up to 10 cm{sup −1} for glasses that change from Li{sup +} to Na{sup +} as the dominant network-modifying species. This indicates that MoO{sub 4} tetrahedra are isolated from the borosilicate network and are surrounded, at least partly, by Na{sup +} and Li{sup +}. Secondary ν{sub 1} frequency effects, with changes up to 7 cm{sup −1}, were also observed with increasing V{sub 2}O{sub 5} and MoO{sub 3} content. These secondary trends may indicate MoO{sub 4}-MoO{sub 4} and MoO{sub 4}-VO{sub 4} clustering, suggesting that V additions may stabilize Mo in the matrix with respect to yellow phase formation. - Highlights: •Raman and XAS data indicate isolated MoO{sub 4} tetrahedra from the borosilicate network. •Mo-O stretch frequency is sensitive to network-modifying Li and Na concentrations. •Mo-O stretch frequency shifts were also seen with increasing V{sub 2}O{sub 5} and MoO{sub 3} content. •The Raman trends may indicate MoO{sub 4}-MoO{sub 4} and MoO{sub 4}-VO{sub 4} clustering in the glass. •V may stabilize Mo in the glass matrix avoiding yellow phase formation.

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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.)

  2. 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

  3. 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.%

  4. 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.

  5. 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.

  6. Liquid-Liquid Phase Separation in Model Nuclear Waste Glasses: A Solid-State Double-Resonance NMR Study

    Energy Technology Data Exchange (ETDEWEB)

    Martineau, Ch.; Michaelis, V.K.; Kroeker, S. [Univ Manitoba, Dept Chem, Winnipeg, MB R3T 2N2 (Canada); Schuller, S. [CEA Valrho Marcoule, LDMC, SECM, DTCD, DEN, F-30207 Bagnols Sur Ceze (France)

    2010-07-01

    Double-resonance nuclear magnetic resonance (NMR) techniques are used in addition to single-resonance NMR experiments to probe the degree of mixing between network-forming cations Si and B, along with the modifier cations Cs{sup +} and Na{sup +} in two molybdenum-bearing model nuclear waste glasses. The double-resonance experiments involving {sup 29}Si in natural abundance are made possible by the implementation of a CPMG pulse-train during the acquisition period of the usual REDOR experiments. For the glass with lower Mo content, the NMR results show a high degree of Si-B mixing, as well as an homogeneous distribution of the cations within the borosilicate network, characteristic of a non-phase-separated glass. For the higher-Mo glass, a decrease of B-Si(Q{sup 4}) mixing is observed, indicating phase separation. {sup 23}Na and {sup 133}Cs NMR results show that although the Cs{sup +} cations, which do not seem to be influenced by the molybdenum content, are spread within the borate network, there is a clustering of the Na{sup +} cations, very likely around the molybdate units. The segregation of a Mo-rich region with Na{sup +} cations appears to shift the bulk borosilicate glass composition toward the metastable liquid liquid immiscibility region and induce additional phase separation. Although no crystallization is observed in the present case, this liquid liquid phase separation is likely to be the first stage of crystallization that can occur at higher Mo loadings or be driven by heat treatment. From this study emerges a consistent picture of the nature and extent of such phase separation phenomena in Mo-bearing glasses, and demonstrates the potential of double-resonance NMR methods for the investigation of phase separation in amorphous materials. (authors)

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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.

  12. 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

  13. 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.

  14. 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

  15. 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

  16. Silicon isotope ratio measurements by inductively coupled plasma tandem mass spectrometry for alteration studies of nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Gourgiotis, Alkiviadis, E-mail: alkiviadis.gourgiotis@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PRP-DGE/SRTG/LT2S, Fontenay-aux-Roses (France); Ducasse, Thomas [CEA, DEN, DTCD, SECM, F-30207 Bagnols-sur-Cèze (France); Barker, Evelyne [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PRP-DGE/SRTG/LT2S, Fontenay-aux-Roses (France); Jollivet, Patrick; Gin, Stéphane [CEA, DEN, DTCD, SECM, F-30207 Bagnols-sur-Cèze (France); Bassot, Sylvain; Cazala, Charlotte [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PRP-DGE/SRTG/LT2S, Fontenay-aux-Roses (France)

    2017-02-15

    High-level, long-lived nuclear waste arising from spent fuel reprocessing is vitrified in silicate glasses for final disposal in deep geologic formations. In order to better understand the mechanisms driving glass dissolution, glass alteration studies, based on silicon isotope ratio monitoring of {sup 29}Si-doped aqueous solutions, were carried out in laboratories. This work explores the capabilities of the new type of quadrupole-based ICP-MS, the Agilent 8800 tandem quadrupole ICP-MS/MS, for accurate silicon isotope ratio determination for alteration studies of nuclear waste glasses. In order to avoid silicon polyatomic interferences, a new analytical method was developed using O{sub 2} as the reaction gas in the Octopole Reaction System (ORS), and silicon isotopes were measured in mass-shift mode. A careful analysis of the potential polyatomic interferences on SiO{sup +} and SiO{sub 2}{sup +} ion species was performed, and we found that SiO{sup +} ion species suffer from important polyatomic interferences coming from the matrix of sample and standard solutions (0.5M HNO{sub 3}). For SiO{sub 2}{sup +}, no interferences were detected, and thus, these ion species were chosen for silicon isotope ratio determination. A number of key settings for accurate isotope ratio analysis like, detector dead time, integration time, number of sweeps, wait time offset, memory blank and instrumental mass fractionation, were considered and optimized. Particular attention was paid to the optimization of abundance sensitivity of the quadrupole mass filter before the ORS. We showed that poor abundance sensitivity leads to a significant shift of the data away from the Exponential Mass Fractionation Law (EMFL) due to the spectral overlaps of silicon isotopes combined with different oxygen isotopes (i.e. {sup 28}Si{sup 16}O{sup 18}O{sup +}, {sup 30}Si{sup 16}O{sup 16}O{sup +}). The developed method was validated by measuring a series of reference solutions with different {sup 29}Si

  17. 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

  18. 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

  19. Silicon isotope ratio measurements by inductively coupled plasma tandem mass spectrometry for alteration studies of nuclear waste glasses.

    Science.gov (United States)

    Gourgiotis, Alkiviadis; Ducasse, Thomas; Barker, Evelyne; Jollivet, Patrick; Gin, Stéphane; Bassot, Sylvain; Cazala, Charlotte

    2017-02-15

    High-level, long-lived nuclear waste arising from spent fuel reprocessing is vitrified in silicate glasses for final disposal in deep geologic formations. In order to better understand the mechanisms driving glass dissolution, glass alteration studies, based on silicon isotope ratio monitoring of 29 Si-doped aqueous solutions, were carried out in laboratories. This work explores the capabilities of the new type of quadrupole-based ICP-MS, the Agilent 8800 tandem quadrupole ICP-MS/MS, for accurate silicon isotope ratio determination for alteration studies of nuclear waste glasses. In order to avoid silicon polyatomic interferences, a new analytical method was developed using O 2 as the reaction gas in the Octopole Reaction System (ORS), and silicon isotopes were measured in mass-shift mode. A careful analysis of the potential polyatomic interferences on SiO + and SiO 2 + ion species was performed, and we found that SiO + ion species suffer from important polyatomic interferences coming from the matrix of sample and standard solutions (0.5M HNO 3 ). For SiO 2 + , no interferences were detected, and thus, these ion species were chosen for silicon isotope ratio determination. A number of key settings for accurate isotope ratio analysis like, detector dead time, integration time, number of sweeps, wait time offset, memory blank and instrumental mass fractionation, were considered and optimized. Particular attention was paid to the optimization of abundance sensitivity of the quadrupole mass filter before the ORS. We showed that poor abundance sensitivity leads to a significant shift of the data away from the Exponential Mass Fractionation Law (EMFL) due to the spectral overlaps of silicon isotopes combined with different oxygen isotopes (i.e. 28 Si 16 O 18 O + , 30 Si 16 O 16 O + ). The developed method was validated by measuring a series of reference solutions with different 29 Si enrichment. Isotope ratio trueness, uncertainty and repeatability were found to be

  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. 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. 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.

  4. Platinoids and molybdenum in nuclear waste containment glasses: a structural study

    International Nuclear Information System (INIS)

    Le Grand, M.

    2000-01-01

    This work deals with the structure of borosilicate nuclear glasses and with some relationships between structure and macroscopic properties. Two types of elements which may disturb the industrial process - platinoids (Ru and Pd) and molybdenum - are central to this work. Platinoids induce weak modifications on the structure of the glass, causing a depolymerization of the glassy network, an increase of the [3] B/ [4] B ratio and a modification of the medium range order around Si between 3.3 and 4.5 angstrom. The modifications of viscosity and density induced by platinoids in the glass are not due to the structural effect of the platinoids. The increase of viscosity is attributed to needle shaped RuO 2 . It can be moderated by imposing reducing conditions during the elaboration of the glass. The slight difference between experimental and calculated densities is due to the increase of the volume percentage of bubbles in the glass with increasing platinoid content. Mo is either present in the glass as molybdic groupings, or mobilized in chemically complex molybdic crystalline phases. The chemical composition and mineralogy of these phases has been obtained using electronic microprobe data and XRD with Rietveld analysis. The distribution of the different elements between the crystalline phases and the glass is strongly influenced by the structural role of the various cations in the glass. The Mo present in the glass appears as MoO 4 tetrahedra, independent of the borosilicate network. The formation of the crystalline phases can be explained by the existence of a precursor in which the MoO 4 tetrahedra are concentrated in rich alkali and earth-alkali bearing areas of the glass. (author)

  5. 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.

  6. 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

  7. 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

  8. 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

  9. Gamma radiation induced changes in nuclear waste glass containing Eu

    Science.gov (United States)

    Mohapatra, M.; Kadam, R. M.; Mishra, R. K.; Kaushik, C. P.; Tomar, B. S.; Godbole, S. V.

    2011-10-01

    Gamma radiation induced changes were investigated in sodium-barium borosilicate glasses containing Eu. The glass composition was similar to that of nuclear waste glasses used for vitrifying Trombay research reactor nuclear waste at Bhabha Atomic Research Centre, India. Photoluminescence (PL) and electron paramagnetic resonance (EPR) techniques were used to study the speciation of the rare earth (RE) ion in the matrix before and after gamma irradiation. Judd-Ofelt ( J- O) analyses of the emission spectra were done before and after irradiation. The spin counting technique was employed to quantify the number of defect centres formed in the glass at the highest gamma dose studied. PL data suggested the stabilisation of the trivalent RE ion in the borosilicate glass matrix both before and after irradiation. It was also observed that, the RE ion distributes itself in two different environments in the irradiated glass. From the EPR data it was observed that, boron oxygen hole centre based radicals are the predominant defect centres produced in the glass after irradiation along with small amount of E’ centres. From the spin counting studies the concentration of defect centres in the glass was calculated to be 350 ppm at 900 kGy. This indicated the fact that bulk of the glass remained unaffected after gamma irradiation up to 900 kGy.

  10. 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.)

  11. 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

  12. Waste vitrification: prediction of acceptable compositions in a lime-soda-silica glass-forming system

    International Nuclear Information System (INIS)

    Gilliam, T.M.; Jantzen, C.M.

    1996-10-01

    A model is presented based upon calculated bridging oxygens which allows the prediction of the region of acceptable glass compositions for a lime-soda-silica glass-forming system containing mixed waste. The model can be used to guide glass formulation studies (e.g., treatability studies) or assess the applicability of vitrification to candidate waste streams

  13. 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

  14. 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

  15. 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.

  16. 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

  17. Elaboration and experimental study of the Borosilicate glass GP 98/12 for the vitrification of the radioactive wastes of KfKarlsruhe Centre (R.F.A.)

    International Nuclear Information System (INIS)

    Ghezal, A.

    1987-09-01

    The transformation into a vitrified block of highly radioactive liquid wastes is actually the best solution for the storage in long run. In West Germany, the research institute in the field of nuclear energy (KfK) has been oriented in this way by developing industrial processes of vitrification and by following studies on the behaviour of the final products. For the fission products, the chosen glasses present good stability characteristics and are used as a first barrier during confinement. Our work, which is part of the research program on radioactive waste vitrification, consists of preparing borosilicate glass GP 98/12 and studying physical and chemical characteristics. We have also contributed to the development and the realization of glass blocks sampling system prepared at pilot scale

  18. 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.

  19. 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)

  20. 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)

  1. Research on the Properties of the Waste Glass Concrete Composite Foundation

    Science.gov (United States)

    Jia, Shilong; Chen, Kaihui; Chen, Zhongliang

    2018-02-01

    The composite foundation of glass concrete can not only reuse the large number of waste glass, but also improve the bearing capacity of weak foundation and soil with special properties. In this paper, the engineering properties of glass concrete composite foundation are studied based on the development situation of glass concrete and the technology of composite foundation.

  2. 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

  3. 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)

  4. 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

  5. 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

  6. Vanadium and Chromium Redox Behavior in borosilicate Nuclear Waste Glasses

    International Nuclear Information System (INIS)

    McKeown, D.; Muller, I.; Gan, H.; Feng, Z.; Viragh, C.; Pegg, I.

    2011-01-01

    X-ray absorption spectroscopy (XAS) was used to characterize vanadium (V) and chromium (Cr) environments in low activity nuclear waste (LAW) glasses synthesized under a variety of redox conditions. V 2 O 5 was added to the melt to improve sulfur incorporation from the waste; however, at sufficiently high concentrations, V increased melt foaming, which lowered melt processing rates. Foaming may be reduced by varying the redox conditions of the melt, while small amounts of Cr are added to reduce melter refractory corrosion. Three parent glasses were studied, where CO-CO 2 mixtures were bubbled through the corresponding melt for increasing time intervals so that a series of redox-adjusted-glasses was synthesized from each parent glass. XAS data indicated that V and Cr behaviors are significantly different in these glasses with respect to the cumulative gas bubbling times: V 4+ /V total ranges from 8 to 35%, while Cr 3+ /Cr total can range from 15 to 100% and even to population distributions including Cr 2+ . As Na-content decreased, V, and especially, Cr became more reduced, when comparing equivalent glasses within a series. The Na-poor glass series show possible redox coupling between V and Cr, where V 4+ populations increase after initial bubbling, but as bubbling time increases, V 4+ populations drop to near the level of the parent glass, while Cr becomes more reduced to the point of having increasing Cr 2+ populations.

  7. 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

  8. 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.

  9. Glass-crystalline materials for active waste incorporation

    International Nuclear Information System (INIS)

    Kulichenko, V.V.; Krylova, N.V.; Vlasov, V.I.; Polyakov, A.S.

    1979-01-01

    This paper presents the results of investigations into the possibility and conditions for using glass-crystalline materials for the incorporation of radionuclides. Materials of a cast pyroxene type that are obtained by smelting calcined wastes with acid blast furnace slags are described. A study was also made of materials of a basalt type prepared from wastes with and without alkali metal salt. Changes in the structure and properties of materials in the process of storage at different temperatures have been studied

  10. 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

  11. 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.

  12. Obsidian: alteration study under hydrothermal-like conditions for its assessment as a nuclear waste glass

    International Nuclear Information System (INIS)

    Rania, Nishi; Shrivastava, J.P.; Bajpai, R.K.

    2013-01-01

    Alteration experiments of obsidian (from Osham Hill, Gujarat, India) were performed under hydrothermal-like conditions. Neo-formed minerals were compared with naturally altered minerals to assess its performance. Altered specimens show partial to complete leaching of glass, where ionic release is of the order of Na>Si>K>Ca>Al = Mg>Mn>Ti. SEM-BSE images show distinct microstructures and mineral paragenesis of smectite, chlorite, nontronite, and illite inside and outside of the secondary layers - show retention of Si, Al, and Mg ions, fixation in the alteration products after their meager release to the solution. Secondary minerals-palagonite, chlorite, calcite, zeolite and white colored clays - formed after experiments largely correspond to altered obsidian in the natural environment since ∼ 65 Ma. (authors)

  13. Obsidian: alteration study under hydrothermal-like conditions for its assessment as a nuclear waste glass

    Energy Technology Data Exchange (ETDEWEB)

    Rania, Nishi; Shrivastava, J.P. [Department of Geology, University of Delhi, Delhi - 110007 (India); Bajpai, R.K. [BETDD, Nuclear Recycle Group, BARC, Mumbai - 400008 (India)

    2013-07-01

    Alteration experiments of obsidian (from Osham Hill, Gujarat, India) were performed under hydrothermal-like conditions. Neo-formed minerals were compared with naturally altered minerals to assess its performance. Altered specimens show partial to complete leaching of glass, where ionic release is of the order of Na>Si>K>Ca>Al = Mg>Mn>Ti. SEM-BSE images show distinct microstructures and mineral paragenesis of smectite, chlorite, nontronite, and illite inside and outside of the secondary layers - show retention of Si, Al, and Mg ions, fixation in the alteration products after their meager release to the solution. Secondary minerals-palagonite, chlorite, calcite, zeolite and white colored clays - formed after experiments largely correspond to altered obsidian in the natural environment since ∼ 65 Ma. (authors)

  14. Thermal phase stability of some simulated Defense waste glasses

    International Nuclear Information System (INIS)

    May, R.P.

    1981-04-01

    Three simulated defense waste glass compositions developed by Savannah River Laboratories were studied to determine viscosity and compositional effects on the comparative thermal phase stabilities of these glasses. The glass compositions are similar except that the 411 glasses are high in lithium and low in sodium compared to the 211 glass, and the T glasses are high in iron and low in aluminum compared to the C glass. Specimens of these glasses were heat treated using isothermal anneals as short as 10 min and up to 15 days over the temperature range of 450 0 C to 1100 0 C. Additionally, a specimen of each glass was cooled at a constant cooling rate of 7 0 C/hour from an 1100 0 C melt down to 500 0 C where it was removed from the furnace. The following were observed. The slow cooling rate of 7 0 C/hour is possible as a canister centerline cooling rate for large canisters. Accordingly, it is important to note that a short range diffusion mechanism like cooperative growth phenomena can result in extensive devitrification at lower temperatures and higher yields than a long-range diffusion mechanism can; and can do it without the growth of large crystals that can fracture the glass. Refractory oxides like CeO 2 and (Ni, Mn, Fe) 2 O 4 form very rapidly at higher temperatures than silicates and significant yields can be obtained at sufficiently high temperatures that settling of these dense phases becomes a major microstructural feature during slow cooling of some glasses. These annealing studies further show that below 500 0 C there is but little devitrification occurring implying that glass canisters stored at 300 0 C may be kinetically stable despite not being thermodynamically so

  15. Thermal phase stability of some simulated Defense waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    May, R.P.

    1981-04-01

    Three simulated defense waste glass compositions developed by Savannah River Laboratories were studied to determine viscosity and compositional effects on the comparative thermal phase stabilities of these glasses. The glass compositions are similar except that the 411 glasses are high in lithium and low in sodium compared to the 211 glass, and the T glasses are high in iron and low in aluminum compared to the C glass. Specimens of these glasses were heat treated using isothermal anneals as short as 10 min and up to 15 days over the temperature range of 450/sup 0/C to 1100/sup 0/C. Additionally, a specimen of each glass was cooled at a constant cooling rate of 7/sup 0/C/hour from an 1100/sup 0/C melt down to 500/sup 0/C where it was removed from the furnace. The following were observed. The slow cooling rate of 7/sup 0/C/hour is possible as a canister centerline cooling rate for large canisters. Accordingly, it is important to note that a short range diffusion mechanism like cooperative growth phenomena can result in extensive devitrification at lower temperatures and higher yields than a long-range diffusion mechanism can; and can do it without the growth of large crystals that can fracture the glass. Refractory oxides like CeO/sub 2/ and (Ni, Mn, Fe)/sub 2/O/sub 4/ form very rapidly at higher temperatures than silicates and significant yields can be obtained at sufficiently high temperatures that settling of these dense phases becomes a major microstructural feature during slow cooling of some glasses. These annealing studies further show that below 500/sup 0/C there is but little devitrification occurring implying that glass canisters stored at 300/sup 0/C may be kinetically stable despite not being thermodynamically so.

  16. 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

  17. 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.

  18. 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

  19. 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.

  20. 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)

  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. Producing glass-ceramics from waste materials

    Energy Technology Data Exchange (ETDEWEB)

    Boccaccini, A.R.; Rawlings, R.D. [Imperial College, London (United Kingdom)

    2002-10-01

    An overview is given of recent research at the Department of Materials of Imperial College, London, UK, concerning the production of useful glass-ceramic products from industrial waste materials. The new work, using controlled crystallisation to improve the properties of vitrified products, could help to solve the problem of what to do with increasing amounts of slag, fly ash and combustion dust. The results show, that it is possible to produce new materials with interesting magnetic and constructive properties.

  3. 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.

  4. 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

  5. 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

  6. 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

  7. Study of parameters of heat treatment in obtaining glass ceramic materials with addition of the industrial waste

    International Nuclear Information System (INIS)

    Kniess, C.T.; Prates, P.B.; Martins, G.J.M.; Riella, H.G.; Matsinhe, Jonas; Kuhnen, N.C.

    2012-01-01

    The production of materials from crystallization of glass, called glass ceramic, have proved interesting by the possibility of development of different microstructures, with reduced grain size and the presence of residual amorphous phase in different quantities. The method that uses the differential thermal analysis (DTA) provides research on the material properties over a wide temperature range, it's widely applied to crystallization processes of glass ceramic materials. Within this context, this paper aims to study the kinetics of nucleation and crystal growth in glass ceramic materials in the system SiO 2 - Al 2 O 3 -Li 2 O, obtained with the addition of mineral coal bottom ash as source of aluminosilicates, through the technique of differential thermal analysis. (author)

  8. 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

  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. 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. Ceramics and glasses for radioactive waste storage

    International Nuclear Information System (INIS)

    Baudin, G.

    1984-06-01

    Borosilicate glasses are mainly choosen for the confinement of fission products; industrial plants are either in operation (AVM) or in construction. Studies of ceramics as a matrix haven't received real application [fr

  12. 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

  13. 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.

  14. 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

  15. Study of the mechanisms underlying resumptions of alteration. Modeling and evaluation of the impact on nuclear waste glasses

    International Nuclear Information System (INIS)

    Fournier, Maxime

    2015-01-01

    A sudden and still poorly understood phenomenon, the resumption of alteration results in a sudden acceleration of the glass alteration rate due to the destabilization of the amorphous passivating layer formed on the glass surface. Understanding the origin and the consequences of this phenomenon is a major issue for the prediction of nuclear glass long-term behavior. This study quantitatively links the alteration degree of a six-oxide reference glass and the formation mechanisms of zeolites and C-S-H that control the solution chemistry. The role played by the decrease in aluminum concentration as an indicator of resumption is highlighted. It appears that the resumption occurrence and rate are correlated to the couple (T, pH), but even in the most adverse situations the resumption rate is lower than the initial alteration rate, which remains the fastest kinetic regime. Previously limited to alkaline pH, the characterization of alteration resumptions was extended to conditions more representative of those found in a geological repository. This approach required the development of a new tool: seeding, that reduces or eliminates the latency period preceding a resumption. The results obtained demonstrate its usefulness in understanding the role of zeolites in amorphous layer destabilization and for modeling alteration resumptions. A geochemical modeling approach to alteration resumption is proposed, based on the formalism of the GRAAL glass alteration model. It is based on the calculation of zeolite thermodynamic constants, on the implementation of their nucleation and growth kinetics, and on assumptions related to the solubility of the amorphous layer. When zeolite precipitation consumes alkali, glass alteration - driven by zeolite precipitation - releases alkali. The model highlights the importance of such chemical couplings and shows that, in the stoichiometry of French nuclear reference glass, their cumulative effects are the cause of a pH decrease which limits

  16. 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

  17. 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.

  18. 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.

  19. 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

  20. 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

  1. 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.

  2. 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.

  3. 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...

  4. 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

  5. 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

  6. 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.

  7. 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

  8. 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.

  9. 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

  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. 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.

  12. 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

  13. 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

  14. 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

  15. 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

  16. Lead recovery and glass microspheres synthesis from waste CRT funnel glasses through carbon thermal reduction enhanced acid leaching process.

    Science.gov (United States)

    Mingfei, Xing; Yaping, Wang; Jun, Li; Hua, Xu

    2016-03-15

    In this study, a novel process for detoxification and reutilization of waste cathode ray tube (CRT) funnel glass was developed by carbon thermal reduction enhanced acid leaching process. The key to this process is removal of lead from the CRT funnel glass and synchronous preparation of glass microspheres. Carbon powder was used as an isolation agent and a reducing agent. Under the isolation of the carbon powder, the funnel glass powder was sintered into glass microspheres. In thermal reduction, PbO in the funnel glass was first reduced to elemental Pb by carbon monoxide and then located on the surface of glass microspheres which can be removed easily by acid leaching. Experimental results showed that temperature, carbon adding amount and holding time were the major parameters that controlled lead removal rate. The maximum lead removal rate was 94.80% and glass microspheres that measured 0.73-14.74μm were obtained successfully by setting the temperature, carbon adding amount and holding time at 1200°C, 10% and 30min, respectively. The prepared glass microspheres may be used as fillers in polymer materials and abrasive materials, among others. Accordingly, this study proposed a practical and economical process for detoxification and recycling of waste lead-containing glass. Copyright © 2015 Elsevier B.V. All rights reserved.

  17. Leaching behavior of simulated high-level waste glass

    International Nuclear Information System (INIS)

    Kamizono, Hiroshi

    1987-03-01

    The author's work in the study on the leaching behavior of simulated high-level waste (HLW) glass were summarized. The subjects described are (1) leach rates at high temperatures, (2) effects of cracks on leach rates, (3) effects of flow rate on leach rates, and (4) an in-situ burial test in natural groundwater. In the following section, the leach rates obtained by various experiments were summarized and discussed. (author)

  18. Physical Characteristics and Technology of Glass Foam from Waste Cathode Ray Tube Glass

    Directory of Open Access Journals (Sweden)

    G. Mucsi

    2013-01-01

    Full Text Available This paper deals with the laboratory investigation of cathode-ray-tube- (CRT- glass-based glass foam, the so-called “Geofil-Bubbles” which can be applied in many fields, mainly in the construction industry (lightweight concrete aggregate, thermal and sound insulation, etc.. In this study, the main process engineering material properties of raw materials, such as particle size distribution, moisture content, density, and specific surface area, are shown. Then, the preparation of raw cathode ray tube glass waste is presented including the following steps: crushing, grinding, mixing, heat curing, coating, and sintering. Experiments were carried out to optimize process circumstances. Effects of sintering conditions—such as temperature, residence time, and particle size fraction of green pellet—on the mechanical stability and particle density of glass foam particles were investigated. The mechanical stability (abrasion resistance was tested by abrasion test in a Deval drum. Furthermore, the cell structure was examined with optical microscopy and SEM. We found that it was possible to produce foam glass (with proper mechanical stability and particle density from CRT glass. The material characteristics of the final product strongly depend on the sintering conditions. Optimum conditions were determined: particle size fraction was found to be 4–6 mm, temperature 800°C, and residence time 7.5 min.

  19. 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

  20. 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

  1. 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...

  2. Waste glass/metal interactions in brines

    International Nuclear Information System (INIS)

    Shade, J.W.; Pederson, L.R.; McVay, G.L.

    1983-05-01

    Leaching studies of MCC 76-68 glass in synthetic brines high in NaCl were performed from 50 to 150 0 C and included interactive testing with ductile iron and titanium. Hydrolysis of the glass matrix was generally slower in saturated brines than in deionized water, due to a lower solubility of silica in the brines. Inclusion of ductile iron in the tests resulted in accelerated leach rates because irion-silica reactions occurred which reduced the silica saturation fraction. At 150 0 C, iron also accelerated the rate of crystalline reaction product formation which were primarily Fe-bearing sepiolite and talc. 16 references

  3. 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

  4. 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.

  5. The behavior of silicon and boron in the surface of corroded nuclear waste glasses: an EFTEM study

    International Nuclear Information System (INIS)

    Buck, E. C.; Smith, K. L.; Blackford, M. G.

    1999-01-01

    Using electron energy-loss filtered transmission electron microscopy (EFTEM), we have observed the formation of silicon-rich zones on the corroded surface of a West Valley (WV6) glass. This layer is approximately 100-200 nm thick and is directly underneath a precipitated smectite clay layer. Under conventional (C)TEM illumination, this layer is invisible; indeed, more commonly used analytical techniques, such as x-ray energy dispersive spectroscopy (EDS), have failed to describe fully the localized changes in the boron and silicon contents across this region. Similar silicon-rich and boron-depleted zones were not found on corroded Savannah River Laboratory (SRL) borosilicate glasses, including SRL-EA and SRL-51, although they possessed similar-looking clay layers. This study demonstrates a new tool for examining the corroded surfaces of materials

  6. 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

  7. 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

  8. 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

  9. 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.

  10. Experimental Design for Hanford Low-Activity Waste Glasses with High Waste Loading

    Energy Technology Data Exchange (ETDEWEB)

    Piepel, Gregory F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cooley, Scott K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vienna, John D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-07-24

    This report discusses the development of an experimental design for the initial phase of the Hanford low-activity waste (LAW) enhanced glass study. This report is based on a manuscript written for an applied statistics journal. Appendices A, B, and E include additional information relevant to the LAW enhanced glass experimental design that is not included in the journal manuscript. The glass composition experimental region is defined by single-component constraints (SCCs), linear multiple-component constraints (MCCs), and a nonlinear MCC involving 15 LAW glass components. Traditional methods and software for designing constrained mixture experiments with SCCs and linear MCCs are not directly applicable because of the nonlinear MCC. A modification of existing methodology to account for the nonlinear MCC was developed and is described in this report. One of the glass components, SO3, has a solubility limit in glass that depends on the composition of the balance of the glass. A goal was to design the experiment so that SO3 would not exceed its predicted solubility limit for any of the experimental glasses. The SO3 solubility limit had previously been modeled by a partial quadratic mixture model expressed in the relative proportions of the 14 other components. The partial quadratic mixture model was used to construct a nonlinear MCC in terms of all 15 components. In addition, there were SCCs and linear MCCs. This report describes how a layered design was generated to (i) account for the SCCs, linear MCCs, and nonlinear MCC and (ii) meet the goals of the study. A layered design consists of points on an outer layer, and inner layer, and a center point. There were 18 outer-layer glasses chosen using optimal experimental design software to augment 147 existing glass compositions that were within the LAW glass composition experimental region. Then 13 inner-layer glasses were chosen with the software to augment the existing and outer

  11. Experimental Design for Hanford Low-Activity Waste Glasses with High Waste Loading

    International Nuclear Information System (INIS)

    Piepel, Gregory F.; Cooley, Scott K.; Vienna, John D.; Crum, Jarrod V.

    2015-01-01

    This report discusses the development of an experimental design for the initial phase of the Hanford low-activity waste (LAW) enhanced glass study. This report is based on a manuscript written for an applied statistics journal. Appendices A, B, and E include additional information relevant to the LAW enhanced glass experimental design that is not included in the journal manuscript. The glass composition experimental region is defined by single-component constraints (SCCs), linear multiple-component constraints (MCCs), and a nonlinear MCC involving 15 LAW glass components. Traditional methods and software for designing constrained mixture experiments with SCCs and linear MCCs are not directly applicable because of the nonlinear MCC. A modification of existing methodology to account for the nonlinear MCC was developed and is described in this report. One of the glass components, SO 3 , has a solubility limit in glass that depends on the composition of the balance of the glass. A goal was to design the experiment so that SO 3 would not exceed its predicted solubility limit for any of the experimental glasses. The SO 3 solubility limit had previously been modeled by a partial quadratic mixture model expressed in the relative proportions of the 14 other components. The partial quadratic mixture model was used to construct a nonlinear MCC in terms of all 15 components. In addition, there were SCCs and linear MCCs. This report describes how a layered design was generated to (i) account for the SCCs, linear MCCs, and nonlinear MCC and (ii) meet the goals of the study. A layered design consists of points on an outer layer, and inner layer, and a center point. There were 18 outer-layer glasses chosen using optimal experimental design software to augment 147 existing glass compositions that were within the LAW glass composition experimental region. Then 13 inner-layer glasses were chosen with the software to augment the existing and outer-layer glasses. The experimental

  12. 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

  13. 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...

  14. Rhyolitic glasses as natural analogues of nuclear waste glasses: behaviour of an Icelandic glass upon natural aqueous corrosion

    International Nuclear Information System (INIS)

    Magonthier, M.-C.; Petit, J.-C.; Dran, J.-C.

    1992-01-01

    A detailed study of the altered rims present in narrow fissures of a 52 ka-old Icelandic obsidian reveals the behaviour of transition and heavy elements, as well as the mechanism and kinetics of alteration, during glass/solution interaction. These complex altered rims are alkali depleted and consist of alternating layers of Fe-rich aluminosilicate and aluminium thihydroxide. The elemental partitioning observed on this naturally corroded obsidian is supported by laboratory experiments performed on the same glass, the elemental accumulation being explained by the formation of a hydrosilicate. A good correlation exists between the thickness of the altered rims and that calculated from the amounts of Fe and Ti accumulated locally. Thus, immobile elements can be used reliably as indices of the extent of alteration because only near-equilibrium conditions occur. The good agreement between the experimental hydration rate of obsidians and the progress of natural corrosion, leads to the assumption that ion diffusion is the long-term controlling mechanism of corrosion. Such an assumption is supported by the particular distribution of the immobile elements which is due to ion diffusion and coprecipitation processes (self-organization genesis). These observations have implications for nuclear waste disposal topics and support the validity of obsidians as analogues of nuclear waste glasses with respect to some local environmental constraints induced by waste packaging and disposal. (author)

  15. 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)

  16. 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.

  17. 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

  18. Application of a glass furnace system to low-level radioactive and mixed waste disposal

    International Nuclear Information System (INIS)

    Klinger, L.; Armstrong, K.

    1986-01-01

    In 1981 Mound began a study to determine the feasibility of using an electrically heated glass furnace for the treatment of low-level radioactive wastes generated at commercial nuclear power facilities. Experiments were designed to determine: Whether the technology offered solutions to industry waste disposal problems, and if so; whether is could meet what were thought to be critical requirements for radioactive thermal waste processing. These requirements include: high quality combustion of organic constituents, capture and immobilization of radioactivity, integrity of final waste form, and cost effectiveness. To address these questions a variety of wastes typical of the types generated by nuclear power facilities, including not only standard trash but also wastes of high aqueous and/or inorganic content, were spiked with waste radioisotopes predominant in plant wastes and processed in the glass furnace. The results of this study indicate that the unit is capable of fully meeting the addressed needs of the nuclear industry for power plant waste processing

  19. 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

  20. 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)

  1. 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

  2. Study of the surface crystallization and resistance to dissolution of niobium phosphate glasses for nuclear waste immobilization; Estudo da cristalizacao superficial e da resistencia a dissolucao de vidros niobofosfatos visando a imobilizacao de rejeitos radioativos

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, Heveline

    2008-07-01

    The surface crystallization and the dissolution rate of three phosphate glass compositions containing different amounts of niobium oxide were studied. The glasses were named Nb30, Nb37, and Nb44 according to the nominal content of niobium oxide in the glass composition. The three compositions were evaluated keeping the P{sub 2}O{sub 5}/K{sub 2}O ratio constant and varying the amount of Nb{sub 2}O{sub 5}. These glasses were produced by melting appropriate chemical compounds at 1500 deg C for 0.5 hour. The crystalline phases which were nucleated on the glass surface after heat treatment were determined by X-ray diffraction. The crystalline structures depend on the amount of niobium oxide in the glass composition. The crystal morphologies were observed by using an optical microscope, and their characteristics are specific for each kind of crystalline phase. The crystal growth rate and the surface nuclei density were determined for each glass composition, and they depend on each crystalline phase nucleated on the surface. From the differential thermal analysis curves it was determined that the Nb44 glass containing 46.5 mol por cent of niobium oxide is the most thermally stable against crystallization when compared to the Nb30 and Nb37 glasses. According to the activation energies determined for crystal growth on the surface of each glass type, the Nb44 glass can also be considered the most resistant one against crystallization. The dissolution rate for the Nb44 glass after 14 days immersed in an aqueous solution with pH equals to 7 at 90 deg C is the lowest (9.0 x 10{sup -7} g. cm{sup -2} . day{sup -1}) when compared to the other two glass compositions. The dissolution rates in acidic and neutral solutions of all studied glasses meet the international standards for materials which can be used in the immobilization of nuclear wastes. (author)

  3. 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

  4. Sinter recrystalization and properties evaluation of glass-ceramic from waste glass bottle and magnesite for extended application

    Directory of Open Access Journals (Sweden)

    As'mau Ibrahim Gebi

    2016-12-01

    Full Text Available In a bid to address environmental challenges associated with the management of waste Coca cola glass bottle, this study set out to develop glass ceramic materials using waste coca cola glass bottles and magnesite from Sakatsimta in Adamawa state. A reagent grade chrome (coloring agent were used to modify the composition of the coca cola glass bottle;  X-ray fluorescence(XRF, X-ray diffraction (XRD and Thermo gravimetric analysis (TGA were used to characterize raw materials, four batches GC-1= Coca cola glass frit +1%Cr2O3, GC-2=97% Coca cola glass frit+ 2% magnesite+1%Cr2O3, GC-3=95% Coca cola glass frit+ 4%magnesite+1%Cr2O3, GC-4=93%Coca cola glass frit+ 6%magnesite+ 1%Cr2O3 were formulated and prepared. Thermal Gradient Analysis (TGA results were used as a guide in selection of three temperatures (7000C, 7500C and 8000C used for the study, three particle sizes -106+75, -75+53, -53µm and 2 hr sintering time were also used, the sinter crystallization route of glass ceramic production was adopted. The samples were characterized by X-ray diffraction (XRD and Scanning Electron Microscope (SEM, the density, porosity, hardness and flexural strength of the resulting glass ceramics were also measured. The resulting glass ceramic materials composed mainly of wollastonite, diopside and anorthite phases depending on composition as indicated by XRD and SEM, the density of the samples increased with increasing sintering temperature and decreasing particle size. The porosity is minimal and it decreases with increasing sintering temperature and decreasing particle size. The obtained glass ceramic materials possess appreciable hardness and flexural strength with GC-3 and GC-4 having the best combination of both properties.

  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. Investigation of the use of waste crushed glass in the production of asphalt mixes

    CSIR Research Space (South Africa)

    Anochie-Boateng, Joseph

    2017-06-01

    Full Text Available Recent study at the Council for Scientific and Industrial Research in South Africa has revealed the potential to substitute depleting natural aggregates with waste crushed glass in asphalt mixes. This paper presents an investigation of the use...

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  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. 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

  17. 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.

  18. 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

  19. 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

  20. 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.

  1. 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%.

  2. 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.

  3. 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.)

  4. 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.

  5. 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.)

  6. 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

  7. Development of a glass matrix for vitrification of sulphate bearing high level radioactive liquid waste

    International Nuclear Information System (INIS)

    Kaushik, C.P.; Mishra, R.K.; Thorat, Vidya; Ramchandran, M.; Amar Kumar; Ozarde, P.D.; Raj, Kanwar; Das, D.

    2004-07-01

    High level radioactive liquid waste (HLW) is generated during reprocessing of spent nuclear fuel. In the earlier reprocessing flow sheet ferrous sulphamate has been used for valancy adjustment of Pu from IV to III for effective separation. This has resulted in generation of HLW containing significance amount of sulphate. Internationally borosilicate glass matrix has been adopted for vitrification of HLW. The first Indian vitrification facility at Waste Immobilislition Plant (WIP), Tarapur a five component borosilicate matrix (SiO 2 :B 2 O 3 :Na 2 O : MnO : TiO 2 ) has been used for vitrification of waste. However at Trombay HLW contain significant amount of sulphate which is not compatible with standard borosilicate formulation. Extensive R and D efforts were made to develop a glass formulation which can accommodate sulphate and other constituents of HLW e.g., U, Al, Ca, etc. This report deals with development work of a glass formulations for immobilization of sulphate bearing waste. Different glass formulations were studied to evaluate the compatibility with respect to sulphate and other constituents as mentioned above. This includes sodium, lead and barium borosilicate glass matrices. Problems encountered in different glass matrices for containment of sulphate have also been addressed. A glass formulation based on barium borosilicate was found to be effective and compatible for sulphate bearing high level waste. (author)

  8. 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

  9. 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.

  10. DHLW Glass Waste Package Criticality Analysis (SCPB:N/A)

    International Nuclear Information System (INIS)

    Davis, J.W.

    1996-01-01

    This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to determine the viability of the Defense High-Level Waste (DHLW) Glass waste package concept with respect to criticality regulatory requirements in compliance with the goals of the Waste Package Implementation Plan (Ref. 5.1) for conceptual design. These design calculations are performed in sufficient detail to provide a comprehensive comparison base with other design alternatives. The objective of this evaluation is to show to what extent the concept meets the regulatory requirements or indicate additional measures that are required for the intact waste package

  11. Electronic waste: chemical characterization glasses of tubes cathode rays with viability for recycling

    International Nuclear Information System (INIS)

    Lima, Norma Maria O.; Morais, Crislene R. Silva; Lima, Lenilde Mergia Ribeiro

    2011-01-01

    Electronic waste, or e-waste, often makes incorrect destinations, which causes serious environmental problems. The aim of this study was to analyze the X-ray fluorescence to study the recycling technology for the glass of Cathode Ray Tubes or, popularly, 'picture tubes', identified by the acronym CRT (Cathode Ray Tubes), which integrate computer monitors. It was observed that the glass screen and funnel analyzed have different chemical compositions. As the silicon oxide (SiO2), the largest component of these glasses percentage 59.89% and 48.63% respectively for the screen and funnel this oxide is responsible for forming the vitreous network. The study of recycling of computer monitors it is important, since about 45% of existing materials on a monitor are made of glass, since it is 100% recyclable and can be reused, thus reducing the amount of waste deposited in the environment. (author)

  12. 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

  13. 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

  14. 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)

  15. 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

  16. 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

  17. 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

  18. Growth of hydrated gel layers in nuclear waste glasses

    International Nuclear Information System (INIS)

    Sullivan, T.M.; Machiels, A.J.

    1984-01-01

    The hydration kinetics of waste glasses in contact with an aqueous solution has been studied by using three different approaches. Emphasis has been placed on modeling processes in the transition zone defined as the region in which the nature of the glass changes from the original dry glass to an open hydrated structure. The first model relies on concentration-dependent diffusion coefficients to obtain a transition zone in which the ions mobility is extremely low compared to that in the gel layer. In the second model, the transition zone and hydrated layer are treated as distinct phases and it is assumed that ion exchange at their common boundary is the rate-controlling process. The third model treats the transition zone as a thin film of constant thickness and low diffusivity. In the absence of appreciable network dissolution, all three models indicate that growth of the gel layer becomes eventually proportional to the square root of time; however, as long as processes in the transition zone are rate controlling, growth is linearly proportional to time

  19. 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.

  20. 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.

  1. 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

  2. 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).

  3. 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)

  4. Characterization of raw materials to obtain the mass for white ware, using waste glass

    International Nuclear Information System (INIS)

    Cavalcanti, M.S.L.; Porto, V.S.; Meneses, R.L; Albuquerque, A.V.; Guedes, B.F.R.; Morais, C.R.S.; Santana, L.N.L.

    2009-01-01

    A major problem faced in the post modern society is the huge amount of glass, accumulated in landfills cities. The glass material is one hundred percent recyclable and has the property to act as fluxes as well as feldspar. Given this premise, this study aimed to characterize materials - raw materials and waste glass regional plan for development of ceramic bodies with the similar behavior produced industrially, using shards of glass to partially replace the feldspar. The materials - raw materials used were clay, ball clay, kaolin, quartz, feldspar and shard of glass, being characterized by the techniques: chemical analysis, size analysis, differential thermal analysis vibrational spectroscopy in the infrared region, the Ray-Diffraction X and scanning electron microscopy. The results showed that the waste had higher rates of vitreous oxides fluxes and similar. (author)

  5. 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

  6. 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

  7. Melter feed viscosity during conversion to glass: Comparison between low-activity waste and high-level waste feeds

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Tongan [Pacific Northwest National Laboratory, Richland Washington; Chun, Jaehun [Pacific Northwest National Laboratory, Richland Washington; Dixon, Derek R. [Pacific Northwest National Laboratory, Richland Washington; Kim, Dongsang [Pacific Northwest National Laboratory, Richland Washington; Crum, Jarrod V. [Pacific Northwest National Laboratory, Richland Washington; Bonham, Charles C. [Pacific Northwest National Laboratory, Richland Washington; VanderVeer, Bradley J. [Pacific Northwest National Laboratory, Richland Washington; Rodriguez, Carmen P. [Pacific Northwest National Laboratory, Richland Washington; Weese, Brigitte L. [Pacific Northwest National Laboratory, Richland Washington; Schweiger, Michael J. [Pacific Northwest National Laboratory, Richland Washington; Kruger, Albert A. [U.S. Department of Energy, Office of River Protection, Richland Washington; Hrma, Pavel [Pacific Northwest National Laboratory, Richland Washington

    2017-12-07

    During nuclear waste vitrification, a melter feed (generally a slurry-like mixture of a nuclear waste and various glass forming and modifying additives) is charged into the melter where undissolved refractory constituents are suspended together with evolved gas bubbles from complex reactions. Knowledge of flow properties of various reacting melter feeds is necessary to understand their unique feed-to-glass conversion processes occurring within a floating layer of melter feed called a cold cap. The viscosity of two low-activity waste (LAW) melter feeds were studied during heating and correlated with volume fractions of undissolved solid phase and gas phase. In contrast to the high-level waste (HLW) melter feed, the effects of undissolved solid and gas phases play comparable roles and are required to represent the viscosity of LAW melter feeds. This study can help bring physical insights to feed viscosity of reacting melter feeds with different compositions and foaming behavior in nuclear waste vitrification.

  8. 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...

  9. 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

  10. 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

  11. Long-term leach rates of glasses containing actual waste

    International Nuclear Information System (INIS)

    Wiley, J.R.; LeRoy, J.H.

    1979-01-01

    Leach rates of borosilicate glasses that contained actual Savannah River Plant waste were measured. Leaching was done by water and by buffer solutions of pH 4, 7, and 9. Leach rates were then determined from the amount of 137 Cs, 90 Sr, and Pu released into the leach solutions. The cumulative fractions leached were fit to a mathematical model that included leaching by diffusion and glass dissolution

  12. Long-term leach rates of glasses containing actual waste

    International Nuclear Information System (INIS)

    Wiley, J.R.; LeRoy, J.H.

    1979-01-01

    Leach rates of borosilicate glasses that contained actual Savannah River Plant waste were measured. Leaching was done by water and by buffer solutions of pH 4, 7, and 9. Leach rates were then determined from the amount of 137 Cs, 90 Sr, and plutonium released into the leach solutions. The cumulative fractions leached were fit to a mathematical model that included leaching by diffusion and glass dissolution. 5 figures, 3 tables

  13. Alteration of national glass in radioactive waste repository host rocks: A conceptional review

    International Nuclear Information System (INIS)

    Apps, J.A.

    1987-01-01

    The storage of high-level radioactive wastes in host rocks containing natural glass has potential chemical advantages, especially if the initial waste temperatures are as high as 250 0 C. However, it is not certain how natural glasses will decompose when exposed to an aqueous phase in a repository environment. The hydration and devitrification of both rhyolitic and natural basaltic natural glasses are reviewed in the context of hypothetical thermodynamic phase relations, infrared spectroscopic data and laboratory studies of synthetic glasses exposed to steam. The findings are compared with field observations and laboratory studies of hydrating and devitrifying natural glasses. The peculiarities of the dependence of hydration and devitrification behavior on compositional variation is noted. There is substantial circumstantial evidence to support the belief that rhyolitic glasses differ from basaltic glasses in their thermodynamic stability and their lattice structure, and that this is manifested by a tendency of the former to hydrate rather than devitrify when exposed to water. Further research remains to be done to confirm the differences in glass structure, and to determine both physically and chemically dependent properties of natural glasses as a function of composition

  14. 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.

  15. 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.)

  16. 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.

  17. 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.

  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. Fracture during cooling of cast borosilicate glass containing nuclear wastes

    International Nuclear Information System (INIS)

    Smith, P.K.; Baxter, C.A.

    1981-09-01

    Procedures and techniques were evaluated to mitigate thermal stress fracture in waste glass as the glass cools after casting. The two principal causes of fracture identified in small-scale testing are internal thermal stresses arising from excessive thermal gradients when cooled too fast, and shear fracturing in the surface of the glass because the stainless steel canister shrinks faster than the glass on cooling. Acoustic emission and ceramographic techniques were used to outline an annealing schedule that requires at least three weeks of controlled cooling below 550 0 C to avoid excessive thermal gradients and corresponding stresses. Fracture arising from canister interactions cannot be relieved by slow cooling, but can be eliminated for stainless steel canisters by using ceramic paper, ceramic or graphite paste linings, or by choosing a canister material with a thermal expansion coefficient comparable to, or less than, that of the glass

  20. 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

  1. Glasses obtained from industrial wastes; Vidros obtidos a partir de residuos industriais

    Energy Technology Data Exchange (ETDEWEB)

    Bortoluzzi, D.; Oliveira Fillho, J.; Uggioni, E. [Universidade do Extremo Sul Catarinense (UNESC), Criciuma, SC (Brazil). Curso de Engenharia de Materiais; Bernardin, A.M., E-mail: amb@unesc.ne [Servico Nacional de Aprendizagem Industrial (SENAI), Tijucas, SC (Brazil). Tecnologia em Ceramica

    2009-07-01

    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{sub 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)

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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)

  7. 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

  8. 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

  9. High insulation foam glass material from waste cathode ray tube panel glass

    DEFF Research Database (Denmark)

    König, Jakob; Petersen, Rasmus Rosenlund; Yue, Yuanzheng

    . In general CRT consists of two types of glasses: barium/strontium containing glass (panel glass) and lead containing glass (funnel and panel glass). In this work we present the possibility to produce high performance insulation material from the recycled lead-free glass. We studied the influence of foaming...... between 750 and 850°C. We investigated the influence of milling time, particle size, foaming and oxidizing agent concentrations, temperature and time on the foaming process, foam density, foam porosity and homogeneity. Only moderate foaming was observed in carbon containing samples, while the addition...... of the oxidizing agent greatly improved the foaming quality. The results showed that the amount of oxygen available from the glass is not sufficient to combust all of the added carbon, therefore, additional oxygen was supplied via manganese reduction. In general, a minimum in the foam glass density was observed...

  10. 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

  11. 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

  12. Investigation of waste glass pouring behavior over a knife edge

    International Nuclear Information System (INIS)

    Ebadian, M.A.

    1998-01-01

    The development of vitrification technology for converting radioactive waste into a glass solid began in the early 1960s. Some problems encountered in the vitrification process are still waiting for a solution. One of them is wicking. During pouring, the glass stream flows down the wall of the pour spout until it reaches an angled cut in the wall. At this point, the stream is supposed to break cleanly away from the wall of the pour spout and fall freely into the canister. However, the glass stream is often pulled toward the wall and does not always fall into the canister, a phenomenon known as wicking. Phase 1 involves the assembly, construction, and testing of a melter capable of supplying molten glass at operational flow rates over a break-off point knife edge. Phase 2 will evaluate the effects of glass and pour spout temperatures as well as glass flow rates on the glass flow behavior over the knife edge. Phase 3 will identify the effects on wicking resulting from varying the knife edge diameter and height as well as changing the back-cut angle of the knife edge. The following tasks were completed in FY97: Design the experimental system for glass melting and pouring; Acquire and assemble the melter system; and Perform initial research work

  13. Disposition of actinides released from high-level waste glass

    International Nuclear Information System (INIS)

    Ebert, W.L.; Bates, J.K.; Buck, E.C.; Gong, M.; Wolf, S.F.

    1994-01-01

    A series of static leach tests was conducted using glasses developed for vitrifying tank wastes at the Savannah River Site to monitor the disposition of actinide elements upon corrosion of the glasses. In these tests, glasses produced from SRL 131 and SRL 202 frits were corroded at 90 degrees C in a tuff groundwater. Tests were conducted using crushed glass at different glass surface area-to-solution volume (S/V) ratios to assess the effect of the S/V on the solution chemistry, the corrosion of the glass, and the disposition of actinide elements. Observations regarding the effects of the S/V on the solution chemistry and the corrosion of the glass matrix have been reported previously. This paper highlights the solution analyses performed to assess how the S/V used in a static leach test affects the disposition of actinide elements between fractions that are suspended or dissolved in the solution, and retained by the altered glass or other materials

  14. 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.)

  15. MELT RATE ENHANCEMENT FOR HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASS FORMULATION FINAL REPORT 08R1360-1

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; KOT W; PEGG IL; JOSEPH I; BARDAKCI T; GAN H; GONG W; CHAUDHURI M

    2010-01-04

    This report describes the development and testing of new glass formulations for high aluminum waste streams that achieve high waste loadings while maintaining high processing rates. The testing was based on the compositions of Hanford High Level Waste (HLW) with limiting concentrations of aluminum specified by the Office of River Protection (ORP). The testing identified glass formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts and small scale melt rate screening tests. The results were used to select compositions for subsequent testing in a DuraMelter 100 (DM100) system. These tests were used to determine processing rates for the selected formulations as well as to examine the effects of increased glass processing temperature, and the form of aluminum in the waste simulant. Finally, one of the formulations was selected for large-scale confirmatory testing on the HLW Pilot Melter (DM1200), which is a one third scale prototype of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW melter and off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy (DOE) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same high-aluminum waste composition used in the present work and other Hanford HLW compositions. The scope of this study was outlined in a Test Plan that was prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the WTP is about 13,500 (equivalent to 40,500 MT glass). This estimate is based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form

  16. MELT RATE ENHANCEMENT FOR HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASS FORMULATION. FINAL REPORT 08R1360-1

    International Nuclear Information System (INIS)

    Kruger, A.A.; Matlack, K.S.; Kot, W.; Pegg, I.L.; Joseph, I.; Bardakci, T.; Gan, H.; Gong, W.; Chaudhuri, M.

    2010-01-01

    This report describes the development and testing of new glass formulations for high aluminum waste streams that achieve high waste loadings while maintaining high processing rates. The testing was based on the compositions of Hanford High Level Waste (HLW) with limiting concentrations of aluminum specified by the Office of River Protection (ORP). The testing identified glass formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts and small scale melt rate screening tests. The results were used to select compositions for subsequent testing in a DuraMelter 100 (DM100) system. These tests were used to determine processing rates for the selected formulations as well as to examine the effects of increased glass processing temperature, and the form of aluminum in the waste simulant. Finally, one of the formulations was selected for large-scale confirmatory testing on the HLW Pilot Melter (DM1200), which is a one third scale prototype of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW melter and off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy (DOE) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same high-aluminum waste composition used in the present work and other Hanford HLW compositions. The scope of this study was outlined in a Test Plan that was prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the WTP is about 13,500 (equivalent to 40,500 MT glass). This estimate is based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form

  17. 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

  18. Glass-ceramic nuclear waste forms obtained by crystallization of SiO 2-Al 2O 3-CaO-ZrO 2-TiO 2 glasses containing lanthanides (Ce, Nd, Eu, Gd, Yb) and actinides (Th): Study of the crystallization from the surface

    Science.gov (United States)

    Loiseau, P.; Caurant, D.

    2010-07-01

    Glass-ceramic materials containing zirconolite (nominally CaZrTi 2O 7) crystals in their bulk can be envisaged as potential waste forms for minor actinides (Np, Am, Cm) and Pu immobilization. In this study such matrices are synthesized by crystallization of SiO 2-Al 2O 3-CaO-ZrO 2-TiO 2 glasses containing lanthanides (Ce, Nd, Eu, Gd, Yb) and actinides (Th) as surrogates. A thin partially crystallized layer containing titanite and anorthite (nominally CaTiSiO 5 and CaAl 2Si 2O 8, respectively) growing from glass surface is also observed. The effect of the nature and concentration of surrogates on the structure, the microstructure and the composition of the crystals formed in the surface layer is presented in this paper. Titanite is the only crystalline phase able to significantly incorporate trivalent lanthanides whereas ThO 2 precipitates in the layer. The crystal growth thermal treatment duration (2-300 h) at high temperature (1050-1200 °C) is shown to strongly affect glass-ceramics microstructure. For the system studied in this paper, it appears that zirconolite is not thermodynamically stable in comparison with titanite growing form glass surface. Nevertheless, for kinetic reasons, such transformation (i.e. zirconolite disappearance to the benefit of titanite) is not expected to occur during interim storage and disposal of the glass-ceramic waste forms because their temperature will never exceed a few hundred degrees.

  19. 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

  20. Utilization of waste glass in translucent and photocatalytic concrete

    NARCIS (Netherlands)

    Spiesz, P.; Rouvas, S.; Brouwers, H.J.H.

    2016-01-01

    Abstract This article addresses the development of a translucent and air purifying concrete containing waste glass. The concrete composition was optimized applying the modified Andreasen & Andersen model to obtain a densely packed system of granular ingredients. Both untreated (unwashed) and washed

  1. 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

  2. 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)

  3. Mix proportions and properties of CLSC made from thin film transition liquid crystal display optical waste glass.

    Science.gov (United States)

    Wang, Her-Yung; Chen, Jyun-Sheng

    2010-01-01

    In this study, controlled low-strength concrete (CLSC) is mixed using different water-to-binder (W/B) ratios (1.1, 1.3 and 1.5) and various percentages of sand substituted by waste LCD glass sand (0%, 10%, 20% and 30%). The properties of the fresh concrete, including compressive strength, electrical resistivity, ultrasonic pulse velocity, permeability ratio and shrinking of the CLSC, are examined. Results show that increases in amount of waste glass added result in better slump and slump flow, longer initial setting time and smaller unit weight. Compressive strength decreases with increasing W/B ratio and greater amounts of waste glass added. Both electrical resistivity and ultrasonic pulse velocity increase with increases in amount of waste glass and decreases in W/B ratio. On the contrary, the permeability ratio increases with increases in W/B ratio, but decreases with greater amounts of waste glass added. CLSC specimens cured for different durations show little changes in length with shrinkage below 0.025%. Our findings reveal that CLSC mixed using waste LCD glass in place of sand can meet design requirements. Recycling of waste LCD glass not only offers an economical substitute for aggregates, but also an ecological alternative for waste management. 2009 Elsevier Ltd. All rights reserved.

  4. 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

  5. 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.

  6. A comparison of the performance of nuclear waste glasses by modeling

    International Nuclear Information System (INIS)

    Grambow, B.; Strachan, D.M.

    1988-12-01

    Through a combination of data collection and computer modeling, the dissolution mechanism of nuclear waste glasses has been investigated and more clearly defined. Glass dissolution can be described as a dissolution/precipitation process in which glass dissolves in aqueous solution and solids precipitate as the solubility products are exceeded. The dissolution process is controlled by activity of the rate-limiting specie H 4 SiO 4 . As a concentration of H 4 SiO 4 increases, the rate of dissolution decreases until a final reaction rate is reached. Between the forward reaction rate (early time) and final reaction rate (very long time), glasses may exhibit an intermediate root time dependence caused by a transport resistance for the diffusion of H 4 SiO 4 within the gel layer on the glass surface. In this report, three glasses are studied: JSS-A, PNL 76-68, and SRL-131. Data from static and dynamic leach tests are assembled, plotted, and successfully modeled. The kinetic parameters for these glasses are reported. With four parameters derived from experiments for each glass, the model can be used to calculate the effects of changes in the initial composition of the water contacting the glass. The effects of convective flow can also be modeled. Furthermore, glasses of different compositions can be readily compared. 49 refs., 27 figs., 5 tabs

  7. Relating structural parameters to leachability in a glass-bonded ceramic waste form

    International Nuclear Information System (INIS)

    Frank, S. M.; Johnson, S. G.; Moschetti, T. L.

    1998-01-01

    Lattice parameters for a crystalline material can be obtained by several methods, notably by analyzing x-ray powder diffraction patterns. By utilizing a computer program to fit a pattern, one can follow the evolution or subtle changes in a structure of a crystalline species in different environments. This work involves such a study for an essential component of the ceramic waste form that is under development at Argonne National Laboratory. Zeolite 4A and zeolite 5A are used to produce two different types of waste forms: a glass-bonded sodalite and a glass-bonded zeolite, respectively. Changes in structure during production of the waste forms are discussed. Specific salt-loadings in the sodalite waste form are related to relative peak intensities of certain reflections in the XRD patterns. Structural parameters for the final waste forms will also be given and related to leachability under standard conditions

  8. The use of positrons to survey alteration layers on synthetic nuclear waste glasses

    International Nuclear Information System (INIS)

    Reiser, Joelle T.; Parruzot, Benjamin; Weber, Marc H.; Ryan, Joseph V.; McCloy, John S.; Wall, Nathalie A.

    2017-01-01

    In order to safeguard society and the environment, understanding radioactive waste glass alteration mechanisms in interactions with solutions and near-field materials, such as Fe, is essential to nuclear waste repository performance assessments. Alteration products are formed at the surface of glasses after reaction with solution. In this study, glass altered in the presence of Fe 0 in aqueous solution formed two alteration layers: one embedded with Fe closer to the surface and one without Fe found deeper in the sample. Both layers were found to be thinner than the alteration layer found in glass altered in aqueous solution only. For the first time, Doppler Broadening Positron Annihilation Spectroscopy (DB-PAS) is used to non-destructively characterize the pore structures of glass altered in the presence of Fe 0 . Advantages and disadvantages of DB-PAS compared to other techniques used to analyze pore structures for altered glass samples are discussed. Ultimately, DB-PAS has shown to be an excellent choice for pore structure characterization for glasses with multiple alteration layers. Monte Carlo modeling predicted positron trajectories through the layers, and helped explain DB-PAS data, which showed that the deeper alteration layer without Fe had a similar composition and pore structure to layers on glass altered in water only.

  9. The use of positrons to survey alteration layers on synthetic nuclear waste glasses

    Science.gov (United States)

    Reiser, Joelle T.; Parruzot, Benjamin; Weber, Marc H.; Ryan, Joseph V.; McCloy, John S.; Wall, Nathalie A.

    2017-07-01

    In order to safeguard society and the environment, understanding radioactive waste glass alteration mechanisms in interactions with solutions and near-field materials, such as Fe, is essential to nuclear waste repository performance assessments. Alteration products are formed at the surface of glasses after reaction with solution. In this study, glass altered in the presence of Fe0 in aqueous solution formed two alteration layers: one embedded with Fe closer to the surface and one without Fe found deeper in the sample. Both layers were found to be thinner than the alteration layer found in glass altered in aqueous solution only. For the first time, Doppler Broadening Positron Annihilation Spectroscopy (DB-PAS) is used to non-destructively characterize the pore structures of glass altered in the presence of Fe0. Advantages and disadvantages of DB-PAS compared to other techniques used to analyze pore structures for altered glass samples are discussed. Ultimately, DB-PAS has shown to be an excellent choice for pore structure characterization for glasses with multiple alteration layers. Monte Carlo modeling predicted positron trajectories through the layers, and helped explain DB-PAS data, which showed that the deeper alteration layer without Fe had a similar composition and pore structure to layers on glass altered in water only.

  10. The use of positrons to survey alteration layers on synthetic nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Reiser, Joelle T. [Washington State University, Chemistry Department, Pullman, WA 99164 (United States); Pacific Northwest National Laboratory, Energy and Environment Directorate, Richland, WA 99352 (United States); Parruzot, Benjamin [Pacific Northwest National Laboratory, Energy and Environment Directorate, Richland, WA 99352 (United States); Weber, Marc H. [Washington State University, Center for Materials Research, Pullman, WA 99164 (United States); Ryan, Joseph V., E-mail: joe.ryan@pnnl.gov [Pacific Northwest National Laboratory, Energy and Environment Directorate, Richland, WA 99352 (United States); McCloy, John S. [Washington State University, Chemistry Department, Pullman, WA 99164 (United States); Pacific Northwest National Laboratory, Energy and Environment Directorate, Richland, WA 99352 (United States); Washington State University, School of Mechanical and Materials Engineering, Pullman, WA 99164 (United States); Wall, Nathalie A., E-mail: nawall@wsu.edu [Washington State University, Chemistry Department, Pullman, WA 99164 (United States)

    2017-07-15

    In order to safeguard society and the environment, understanding radioactive waste glass alteration mechanisms in interactions with solutions and near-field materials, such as Fe, is essential to nuclear waste repository performance assessments. Alteration products are formed at the surface of glasses after reaction with solution. In this study, glass altered in the presence of Fe{sup 0} in aqueous solution formed two alteration layers: one embedded with Fe closer to the surface and one without Fe found deeper in the sample. Both layers were found to be thinner than the alteration layer found in glass altered in aqueous solution only. For the first time, Doppler Broadening Positron Annihilation Spectroscopy (DB-PAS) is used to non-destructively characterize the pore structures of glass altered in the presence of Fe{sup 0}. Advantages and disadvantages of DB-PAS compared to other techniques used to analyze pore structures for altered glass samples are discussed. Ultimately, DB-PAS has shown to be an excellent choice for pore structure characterization for glasses with multiple alteration layers. Monte Carlo modeling predicted positron trajectories through the layers, and helped explain DB-PAS data, which showed that the deeper alteration layer without Fe had a similar composition and pore structure to layers on glass altered in water only.

  11. 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

  12. Chemical composition analysis and product consistency tests to support enhanced Hanford waste glass models. Results for the third set of high alumina outer layer matrix glasses

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States)

    2015-12-01

    In this report, the Savannah River National Laboratory provides chemical analyses and Product Consistency Test (PCT) results for 14 simulated high level waste glasses fabricated by the Pacific Northwest National Laboratory. The results of these analyses will be used as part of efforts to revise or extend the validation regions of the current Hanford Waste Treatment and Immobilization Plant glass property models to cover a broader span of waste compositions. The measured chemical composition data are reported and compared with the targeted values for each component for each glass. All of the measured sums of oxides for the study glasses fell within the interval of 96.9 to 100.8 wt %, indicating recovery of all components. Comparisons of the targeted and measured chemical compositions showed that the measured values for the glasses met the targeted concentrations within 10% for those components present at more than 5 wt %. The PCT results were normalized to both the targeted and measured compositions of the study glasses. Several of the glasses exhibited increases in normalized concentrations (NCi) after the canister centerline cooled (CCC) heat treatment. Five of the glasses, after the CCC heat treatment, had NCB values that exceeded that of the Environmental Assessment (EA) benchmark glass. These results can be combined with additional characterization, including X-ray diffraction, to determine the cause of the higher release rates.

  13. 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

  14. Effect of Technetium-99 sources on its retention in low activity waste glass

    Science.gov (United States)

    Luksic, Steven A.; Kim, Dong-Sang; Um, Wooyong; Wang, Guohui; Schweiger, Michael J.; Soderquist, Chuck Z.; Lukens, Wayne; Kruger, Albert A.

    2018-05-01

    Small-scale crucible melting tests on simulated waste glass were performed with technetium-99 (Tc-99) introduced as different species in a representative low activity waste simulant. The glass saw an increase in Tc-99 retention when TcO2•2H2O and various Tc-minerals containing reduced tetravalent Tc were used compared to tests in which pertechnetate with heptavalent Tc was used. We postulate that the increase of Tc retention is likely caused by different reaction paths for Tc incorporation into glass during early stages of melting, rather than the low volatility of reduced tetravalent Tc compounds, which has been a generally accepted idea. Additional studies are needed to clarify the exact mechanisms relevant to the effect of reduced Tc compounds on Tc incorporation into or volatilization from the glass melt.

  15. Effect of Technetium-99 Sources on Its Retention in Low Activity Waste Glass

    Energy Technology Data Exchange (ETDEWEB)

    Luksic, Steven A.; Kim, Dong-Sang; Um, Wooyong; Wang, Guohui; Schweiger, Michael J.; Soderquist, Chuck Z.; Lukens, Wayne W.; Kruger, Albert A.

    2018-05-01

    Small-scale crucible melting tests on simulated waste glass were performed with technetium-99 (Tc-99) introduced as different species in a representative low activity waste simulant. The glass saw an increase in Tc-99 retention when TcO2∙2H2O and various Tc-minerals containing reduced tetravalent Tc were used compared to tests in which pertechnetate with hexavalent Tc was used. We postulate that the increase of Tc retention is likely caused by different reaction paths for Tc incorporation into glass during early stages of melting, rather than the low volatility of reduced tetravalent Tc compounds, which has been a generally accepted idea. Additional studies are needed to clarify the exact mechanisms relevant to the effect of reduced Tc compounds on Tc incorporation into or volatilization from glass melt.

  16. Studies on the thermal decomposition of nitrates found in highly active waste and of chemicals used to convert the waste to glass

    International Nuclear Information System (INIS)

    Chun, K.S.

    1977-05-01

    The decomposition of all the individual chemicals used in the Harwell inactive vitrification pilot plant has been studied by means of a thermal balance. Weight loss curves to 1100 0 C have been obtained. The four materials sodium nitrate, cesium nitrate, lithium nitrate and ruthenium nitroso-nitrate (solution) showed a greater weight loss than that based on an oxide yield, and hence these compounds or their products of decomposition are volatile below 1100 0 C. The remaining materials suffered a weight loss no more than that corresponding to a full yield of the oxide, and hence they were not volatile below 1100 0 C. Most of the chemicals begin to decompose at less than 75 0 C but the nitrates of cesium, strontium, barium and sodium not until 295 0 to 590 0 C. The results obtained can be used in the analysis of process conditions in the vitrification and calcination of highly radioactive wastes and also of the thermal decomposition behaviour of mixtures containing those materials. The materials tested were: Al(NO 3 ) 3 .9H 2 O, Ba(NO 3 ) 2 , CaNo 3 , Cr(NO 3 ) 3 .9H 2 O, Fe(NO 3 ) 3 .9H 2 O. Mg(NO 3 ) 2 .6H 2 O, Ni(NO 3 ) 2 .6H 2 O, R.E. Nitrates, Ruthenium Solution, Sr(NO 3 ) 2 , UO 2 (NO 3 ) 2 .6H 2 O, Zn(NO 3 ) 2 .6H 2 O. Zirconium Solution, 'Gasil WP' Silica, 'Neosyl' Silica, LiOH.H 2 O. LiNO 3 .3H 2 O, Na 2 CO 3 , NaNO 3 , Na 2 B 4 O 7 .10H 2 O. (author)

  17. Lead recovery and high silica glass powder synthesis from waste CRT funnel glasses through carbon thermal reduction enhanced glass phase separation process

    Energy Technology Data Exchange (ETDEWEB)

    Xing, Mingfei [Henan Key Laboratory Cultivation Base of Mine Environmental Protection and Ecological Remediation, Henan Polytechnic University, Jiaozuo 454000 Henan China (China); Institute of Resource and Environment, Henan Polytechnic University, Jiaozuo 454000 Henan China (China); Fu, Zegang [Institute of Resource and Environment, Henan Polytechnic University, Jiaozuo 454000 Henan China (China); Wang, Yaping, E-mail: wangyp326@163.com [School of Surveying and Land Information Engineering, Henan Polytechnic University, Jiaozuo 454000, Henan China (China); Wang, Jingyu [Institute of Resource and Environment, Henan Polytechnic University, Jiaozuo 454000 Henan China (China); Zhang, Zhiyuan [Research Center for Eco-Environmental Sciences, Chinese Academy of Sciences, Beijing 100085 (China)

    2017-01-15

    Highlights: • CRT funnel glass was remelted with B{sub 2}O{sub 3} in reducing atmosphere. • A part of PbO was reduced into Pb and detached from the glass phase. • The rest of PbO and other metal oxides were mainly concentrated in the B{sub 2}O{sub 3} phase. • PbO enriched in the interconnected B{sub 2}O{sub 3} phase can be completely leached out by HNO{sub 3}. • High silica glass powder(SiO{sub 2} purity >95%) was obtained after the leaching process. - Abstract: In this study, a novel process for the removal of toxic lead from the CRT funnel glass and synchronous preparation of high silica glass powder was developed by a carbon-thermal reduction enhanced glass phase separation process. CRT funnel glass was remelted with B{sub 2}O{sub 3} in reducing atmosphere. In the thermal process, a part of PbO contained in the funnel glass was reduced into metallic Pb and detached from the glass phase. The rest of PbO and other metal oxides (including Na{sub 2}O, K{sub 2}O, Al{sub 2}O{sub 3,} BaO and CaO) were mainly concentrated in the boric oxide phase. The metallic Pb phase and boric oxide phase were completely leached out by 5 mol/L HNO{sub 3}. The lead removal rate was 99.80% and high silica glass powder (SiO{sub 2} purity >95 wt%) was obtained by setting the temperature, B{sub 2}O{sub 3} added amount and holding time at 1000 °C, 20% and 30 mins, respectively. The prepared high silicate glass powders can be used as catalyst carrier, semipermeable membranes, adsorbents or be remelted into high silicate glass as an ideal substitute for quartz glass. Thus this study proposed an eco-friendly and economical process for recycling Pb-rich electronic glass waste.

  18. Thermo-chemistry of nuclear waste glasses: a new approach

    International Nuclear Information System (INIS)

    Linard, Y.; Neuville, D.R.; Richet, P.

    1997-01-01

    Understanding of the stability and weathering of glasses used for storing fission products is hampered by a general lack of basic thermochemical information. Models have been setup to predict Gibbs free energies of dissolution of glasses, but ascertaining their accuracy is made difficult by the very lack of reliable experimental data with which model results should be compared. As enthalpies of formation can in principle be determined from usual solution calorimetry experiments, the lack of Gibbs-free energy data for glasses mainly stems from the fact that, as disordered substances, glasses do not obey the third principle and have indeed large configurational entropies. These entropies can be determined from thermochemical measurements only when there exist a congruently melting crystalline compound with the same composition. Using available data, we have calculated the Gibbs-free energies of formation of a series of silicate glasses for which such a calorimetric determination is possible. With these results, we assess the predictions of Paul's model (1977) for calculating Gibbs-free energies of dissolution. As the complex compositions of the borosilicate glasses used for nuclear waste storage prevent determining configurational entropies by calorimetric methods, we point out how these can be determined instead from viscosity measurements. We finally discuss the implications of this approach for modeling of water-glass interactions. (authors)

  19. Conceptual process for conversion of high level waste to glass

    International Nuclear Information System (INIS)

    1975-01-01

    During a ten-year period highly radioactive wastes amounting to 22 million gallons of salt cake and 5 million gallons of wet sludge are to be converted to 1.2 million gallons of glass and 24 million gallons of decontaminated salt cake and placed in the new storage facilities which will provide high assurance of containment with minimal reliance on maintenance and surveillance. The glass will contain nearly all of the radioactivity in a form that is highly resistant to leaching and dispersion. The salt cake will contain a small amount of residual radioactivity. The process is shown in Figure 1 and the facilities may be arranged in seven modules to accomplish seven tasks, (1) remove wastes from tanks, (2) separate sludge and salt, (3) decontaminate salt, (4) solidify and package sludge and 137 Cs, (5) solidify and package decontaminated salt, (6) store high level waste, and (7) store decontaminated salt cake

  20. 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

  1. Chemical composition measurements of the low activity waste (LAW) EPA-Series glasses

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. [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)

    2016-03-01

    In this report, the Savannah River National Laboratory provides chemical analysis results for a series of simulated low activity waste glasses provided by Pacific Northwest National Laboratory as part of an ongoing development task. The measured chemical composition data are reported and compared with the targeted values for each component for each glass. A detailed review showed no indications of errors in the preparation or measurement of the study glasses. All of the measured sums of oxides for the study glasses fell within the interval of 100.2 to 100.8 wt %, indicating recovery of all components. Comparisons of the targeted and measured chemical compositions showed that the measured values for the glasses met the targeted concentrations within 10% for those components present at more than 5 wt %.

  2. The effect of devitrification on leaching rate of glass containing simulated high level liquid waste (HLLW)

    International Nuclear Information System (INIS)

    Suryantoro; Sumarbagiono; Martono, H.

    1996-01-01

    Effect of devitrification on leaching rate of glass named G1 and G2 each contains 20 wt% and 30wt% of waste has been studied. devitrification of waste - glass has been carried out by heating those specimens at 850 o C for 10, 18, 26, 34, 42 and 50 hours respectively. The weight percentage of crystal in waste glass was determined by X-ray diffractometer and leaching rate was determined by soxhlet apparatus at 100 o C for 24 hours. The longer heating time, the more weight percentage of crystal is formed. The results show that leaching rate of G2 specimens are higher than those of G1. For G1 the leaching rate at 850 o C in 20 times than without heating, and for G2 leaching rate is 15.7 times than without heating. (author)

  3. 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)

  4. 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

  5. Fracture toughness measurements on a glass bonded sodalite high-level waste form

    International Nuclear Information System (INIS)

    DiSanto, T.; Goff, K. M.; Johnson, S. G.; O'Holleran, T. P.

    1999-01-01

    The electrometallurgical treatment of metallic spent nuclear fuel produces two high-level waste streams; cladding hulls and chloride salt. Argonne National Laboratory is developing a glass bonded sodalite waste form to immobilize the salt waste stream. The waste form consists of 75 Vol.% crystalline sodalite (containing the salt) with 25 Vol.% of an ''intergranular'' glassy phase. Microindentation fracture toughness measurements were performed on representative samples of this material using a Vickers indenter. Palmqvist cracking was confirmed by post-indentation polishing of a test sample. Young's modulus was measured by an acoustic technique. Fracture toughness, microhardness, and Young's modulus values are reported, along with results from scanning electron microscopy studies

  6. Development of iodine waste forms using low-temperature sintering glass

    International Nuclear Information System (INIS)

    Krumhansl, James Lee; Nenoff, Tina Maria; Garino, Terry J.; Rademacher, David

    2010-01-01

    This presentation will describe our recent work on the use of low temperature-sintering glass powders mixed with either AgI or AgI-zeolite to produce a stable waste form. Radioactive iodine ( 129 I, half-life of 1.6 x 10 7 years) is generated in the nuclear fuel cycle and is of particular concern due to its extremely long half-life and its effects on human health. As part of the DOE/NE Advanced Fuel Cycle Initiative (AFCI), the separation of 129 I from spent fuel during fuel reprocessing is being studied. In the spent fuel reprocessing scheme under consideration, the iodine is released in gaseous form and collected using Ag-loaded zeolites, to form AgI. Although AgI has extremely low solubility in water, it has a relatively high vapor pressure at moderate temperatures (>550 C), thus limiting the thermal processing. Because of this, immobilization using borosilicate glass is not feasible. Therefore, a bismuth oxide-based glasses are being studied due to the low solubility of bismuth oxide in aqueous solution at pH > 7. These waste forms were processed at 500 C, where AgI volatility is low but the glass powder is able to first densify by viscous sintering and then crystallize. Since the glass is not melted, a more chemically stable glass can be used. The AgI-glass mixture was found to have high iodine leach resistance in these initial studies.

  7. Radioactive wastes immobilization in glasses and ceramics

    International Nuclear Information System (INIS)

    Zanotto, E.D.

    1983-01-01

    A review on the several options available for encapsulation and disposal of high level radioactive liquid wastes is presented, along with the relative merits and disadvantages of each material to be encapsulated. Some of the main fields requiring further advancements in both scientific and technological research are discussed and a few suggestions for the solution of the brazilian problem are given. (Author) [pt

  8. Use of glass-ceramic materials for the fixation of radioactive wastes

    International Nuclear Information System (INIS)

    Minaev, A.A.; Oziraner, S.N.; Prokhorova, N.P.

    1979-01-01

    This paper is concerned with the study of the crystallization of phosphate and silicate glasses. It was shown that temperature and time of storage influence considerably the crystallization of glasses and that crystallization very often increases their rates of leaching to a great extent. However, there are glasses in which crystallization does not result in leaching rate increase. It seems reasonable to use these materials for the fixation of radioactive wastes. The main reasons for the increase in the leaching rate during crystallization are the formation of porosity and soluble crystal phases

  9. Neutron diffraction studies of glasses

    International Nuclear Information System (INIS)

    Wright, A.C.

    1987-01-01

    A survey is given of the application of neutron diffraction to structural studies of oxide and halide glasses. As with crystalline materials, neutron and X-ray diffraction are the major structural probes for glasses and other amorphous solids, particularly in respect of intermediate range order. The glasses discussed mostly have structures which are dominated by a network in which the bonding is predominantly covalent. The examples discussed demonstrate the power of the neutron diffraction technique in the investigation of the structures of inorganic glasses. The best modern diffraction experiments are capable of providing accurate data with high real space resolution, which if used correctly, are an extremely fine filter for the various structural models proposed in the literature. 42 refs

  10. Production of sodalite waste forms by addition of glass

    International Nuclear Information System (INIS)

    Pereira, C.

    1995-01-01

    Spent nuclear fuel can be treated in a molten salt electrorefiner for conversion into metal and mineral waste forms for geologic disposal. Sodalite is one of the mineral waste forms under study. Fission products in the molten salt are ion-exchanged into zeolite A, which is converted to sodalite and consolidated. Sodalite can be formed directly from mixtures of salt and zeolite A at temperatures above 975 K; however, nepheline is usually produced as a secondary phase. Addition of small amounts of glass frit to the mixture reduced nepheline formation significantly. Loss of fission products was not observed for reaction below 1000 K. Hot-pressing of the sodalite powders yielded dense pellets (∼2.3 g/cm 3 ) without any loss of fission product species. Normalized release rates were below 1 g/m 2 ·day for pre-washed samples in 28-day leach tests based on standard MCC-1 tests but increased with the presence of free salt on the sodalite

  11. A literature review of surface alteration layer effects on waste glass behavior

    International Nuclear Information System (INIS)

    Feng, X.; Cunnane, J.C.; Bates, J.K.

    1993-01-01

    When in contact with an aqueous solution, nuclear waste glass is subject to a chemical attack that results in progressive alteration. During tills alteration, constituent elements of the glass pass into the solution; elements initially in solution diffuse into, or are adsorbed onto, the solid; and new phases appear. This results in the formation of surface layers on the reacted glass. The glass corrosion and radionuclide release can be better understood by investigating these surface layer effects. In the past decade, there have been numerous studies regarding the effects of surface layers on glass reactions. This paper presents a systematic analysis and summary of the past knowledge regarding the effects of surface layers on glass-water interaction. This paper describes the major formation mechanisms of surface layers; reviews the role of surface layers in controlling mass transport and glass reaction affinity (through crystalline phases, an amorphous silica, a gel layer, or all the components in the glass); and discusses how the surface layers contribute to the retention of radionuclides during glass dissolution

  12. Borosilicate glasses for the high activity waste vetrification

    International Nuclear Information System (INIS)

    Cantale, C.; Donato, A.; Guidi, G.

    1984-01-01

    Some results concerning the researches carried out on the high-level wastes vitrification at ENEA, Comb-Mepis-Rifiu laboratory are reported. A fission product solution referred to power plant nuclear fuel reprocessing has been selected and simulated with no radioactive chemicals. Some glass composition have been tested for the vitrification of this solution, the best of them being taken into consideration for real active tests at the hot bench scale plant ESTER in Ispra. The final glasses have been characterized from the chemical and physical point of view; moreover some microstructural investigations have been performed in order to identify few microsegregations and to test the degree of amorphousness of the products

  13. Calculation of the viscosity of nuclear waste glass systems

    International Nuclear Information System (INIS)

    Shah, R.; Behrman, E.C.; Oksoy, D.

    1990-01-01

    Viscosity is one of the most important processing parameters and one of the most difficult to calculate theoretically, particularly for multicomponent systems like nuclear waste glasses. Here, the authors propose a semi-empirical approach based on the Fulcher equation, involving identification of key variables, for which coefficients are then determined by regression analysis. Results are presented for two glass systems, and compared to results of previous workers and to experiment. The authors also sketch a first-order statistical mechanical perturbation theory calculation for the effects on viscosity of a change in composition of the melt

  14. 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

  15. A hazardous waste from secondary aluminium metallurgy as a new raw material for calcium aluminate glasses.

    Science.gov (United States)

    López-Delgado, Aurora; Tayibi, Hanan; Pérez, Carlos; Alguacil, Francisco José; López, Félix Antonio

    2009-06-15

    A solid waste coming from the secondary aluminium industry was successfully vitrified in the ternary CaO-Al(2)O(3)-SiO(2) system at 1500 degrees C. This waste is a complex material which is considered hazardous because of its behaviour in the presence of water or moisture. In these conditions, the dust can generate gases such as H(2), NH(3), CH(4), H(2)S, along with heat and potential aluminothermy. Only silica sand and calcium carbonate were added as external raw materials to complete the glasses formula. Different nominal compositions of glasses, with Al(2)O(3) ranging between 20% and 54%, were studied to determine the glass forming area. The glasses obtained allow the immobilisation of up to 75% of waste in a multicomponent oxide system in which all the components of the waste are incorporated. The microhardness Hv values varied between 6.05 and 6.62GPa and the linear thermal expansion coefficient, alpha, varied between (62 and 139)x10(-7)K(-1). Several glasses showed a high hydrolytic resistance in deionised water at 98 degrees C.

  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. X-ray tomography of feed-to-glass transition of simulated borosilicate waste glasses

    International Nuclear Information System (INIS)

    Harris, William H.; Guillen, Donna P.; Klouzek, Jaroslav; Pokorny, Richard; Yano, Tetsuji

    2017-01-01

    The feed composition of a high level nuclear waste (HLW) glass melter affects the overall melting rate by influencing the chemical, thermophysical, and morphological properties of a relatively insulating cold cap layer over the molten phase where the primary feed vitrification reactions occur. Data from X ray computed tomography imaging of melting pellets comprised of a simulated high-aluminum HLW feed heated at a rate of 10°C/min reveal the distribution and morphology of bubbles, collectively known as primary foam, within this layer for various SiO 2 /(Li 2 CO 3 +H 3 BO 3 +Na 2 CO 3 ) mass fractions at temperatures between 600°C and 1040°C. To track melting dynamics, cross-sections obtained through the central profile of the pellet were digitally segmented into primary foam and a condensed phase. Pellet dimensions were extracted using Photoshop CS6 tools while the DREAM.3D software package was used to calculate pellet profile area, average and maximum bubble areas, and two-dimensional void fraction. The measured linear increase in the pellet area expansion rates – and therefore the increase in batch gas evolution rates – with SiO 2 /(Li 2 CO 3 +H 3 BO 3 +Na 2 CO 3 ) mass fraction despite an exponential increase in viscosity of the final waste glass at 1050°C and a lower total amount of gas-evolving species suggest that the retention of primary foam with large average bubble size at higher temperatures results from faster reaction kinetics rather than increased viscosity. However, viscosity does affect the initial foam collapse temperature by supporting the growth of larger bubbles. Because the maximum bubble size is limited by the pellet dimensions, larger scale studies are needed to understand primary foam morphology at high temperatures. This temperature-dependent morphological data can be used in future investigations to synthetically generate cold cap structures for use in models of heat transfer within a HLW glass melter.

  18. 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

  19. Characterization of leached surface layers on simulated high-level waste glasses by sputter-induced optical emission

    International Nuclear Information System (INIS)

    Houser, C.; Tsong, I.S.T.; White, W.B.

    1979-01-01

    The leaching process in simulated waste encapsulant glasses was studied by measuring the compositional depth-profiles of H (from water), the glass framework formers Si and B, the alkalis Na and Cs, the alkaline earths Ca and Sr, the transition metals Mo and Fe, the rare-earths La, Ce, and Nd, using the technique of sputter-induced optical emission. The leaching process of these glasses is highly complex. In addition to alkali/hydrogen exchange, there is breakdown of the glass framework, build-up of barrier layers on the surface, and formation of layered reaction zones of distinctly different chemistry all within the outer micrometer of the glass

  20. Immobilization of high-level wastes into sintered glass: 1

    International Nuclear Information System (INIS)

    Russo, D.O.; Messi de Bernasconi, N.; Audero, M.A.

    1987-01-01

    In order to immobilize the high-level radioactive wastes from fuel elements reprocessing, borosilicate glass was adopted. Sintering experiments are described with the variety VG 98/12 (SiO 2 , TiO 2 , Al 2 O 3 , B 2 O 3 , MgO, CaO and Na 2 O) (which does not present devitrification problems) mixed with simulated calcinated wastes. The hot pressing line (sintering under pressure) was explored in two variants 1: In can; 2: In graphite matrix with sintered pellet extraction. With scanning electron microscopy it is observed that the simulated wastes do not disolve in the vitreous matrix, but they remain dispersed in the same. The results obtained point out that the leaching velocities are independent from the density and from the matrix type employed, as well as from the fact that the wastes do no dissolve in the matrix. (M.E.L.) [es

  1. Calcium titanium silicate based glass-ceramic for nuclear waste immobilisation

    Science.gov (United States)

    Sharma, K.; Srivastav, A. P.; Goswami, M.; Krishnan, Madangopal

    2018-04-01

    Titanate based ceramics (synroc) have been studied for immobilisation of nuclear wastes due to their high radiation and thermal stability. The aim of this study is to synthesis glass-ceramic with stable phases from alumino silicate glass composition and study the loading behavior of actinides in glass-ceramics. The effects of CaO and TiO2 addition on phase evolution and structural properties of alumino silicate based glasses with nominal composition x(10CaO-9TiO2)-y(10Na2O-5 Al2O3-56SiO2-10B2O3); where z = x/y = 1.4-1.8 are reported. The glasses are prepared by melt-quench technique and characterized for thermal and structural properties using DTA and Raman Spectroscopy. Glass transition and peak crystallization temperatures decrease with increase of CaO and TiO2 content, which implies the weakening of glass network and increased tendency of glasses towards crystallization. Sphene (CaTiSiO5) and perovskite (CaTiO3) crystalline phases are confirmed from XRD which are well known stable phase for conditioning of actinides. The microsturcture and elemental analysis indicate the presence of actinide in stable crystalline phases.

  2. Processing glass-pyrochlore composites for nuclear waste encapsulation

    International Nuclear Information System (INIS)

    Pace, S.; Cannillo, V.; Wu, J.; Boccaccini, D.N.; Seglem, S.; Boccaccini, A.R.

    2005-01-01

    Glass matrix composites have been developed as alternative materials to immobilize nuclear solid waste, in particular actinides. These composites are made of soda borosilicate glass matrix, into which particles of lanthanum zirconate pyrochlore are encapsulated in concentrations of 30 vol.%. The fabrication process involves powder mixing followed by hot-pressing. At the relatively low processing temperature used (620 deg. C), the pyrochlore crystalline structure of the zirconate, which is relevant for containment of radioactive nuclei, remains unaltered. The microstructure of the composites exhibits a homogeneous distribution of isolated pyrochlore particles in the glass matrix and strong bonding at the matrix-particle interfaces. Hot-pressing was found to lead to high densification (95% th.d.) of the composite. The materials are characterized by relatively high elastic modulus, flexural strength, hardness and fracture toughness. A numerical approach using a microstructure-based finite element solver was used in order to investigate the mechanical properties of the composites

  3. 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

  4. Elaboration of new ceramic composites containing glass fibre production wastes

    International Nuclear Information System (INIS)

    Rozenstrauha, I.; Sosins, G.; Krage, L.; Sedmale, G.; Vaiciukyniene, D.

    2013-01-01

    Two main by-products or waste from the production of glass fibre are following: sewage sludge containing montmorillonite clay as sorbent material and ca 50 % of organic matter as well as waste glass from aluminium borosilicate glass fibre with relatively high softening temperature (> 600 degree centigrade). In order to elaborate different new ceramic products (porous or dense composites) the mentioned by-products and illitic clay from two different layers of Apriki deposit (Latvia) with illite content in clay fraction up to 80-90 % was used as a matrix. The raw materials were investigated by differential-thermal (DTA) and XRD analysis. Ternary compositions were prepared from mixtures of 15 - 35 wt % of sludge, 20 wt % of waste glass and 45 - 65 wt % of clay and the pressed green bodies were thermally treated in sintering temperature range from 1080 to 1120 degree centigrade in different treatment conditions. Materials produced in temperature range 1090 - 1100 degree centigrade with the most optimal properties - porosity 38 - 52 %, water absorption 39 -47 % and bulk density 1.35 - 1.67 g/cm 3 were selected for production of porous ceramics and materials showing porosity 0.35 - 1.1 %, water absorption 0.7 - 2.6 % and bulk density 2.1 - 2.3 g/cm 3 - for dense ceramic composites. Obtained results indicated that incorporation up to 25 wt % of sewage sludge is beneficial for production of both ceramic products and glass-ceramic composites according to the technological properties. Structural analysis of elaborated composite materials was performed by scanning electron microscopy(SEM). By X-ray diffraction analysis (XRD) the quartz, diopside and anorthite crystalline phases were detected. (Author)

  5. An approach to thermochemical modeling of nuclear waste glass

    International Nuclear Information System (INIS)

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

    1998-01-01

    This initial work is aimed at developing a basic understanding of the phase equilibria and solid solution behavior of the constituents of waste glass. Current, experimentally determined values are less than desirable since they depend on measurement of the leach rate under non-realistic conditions designed to accelerate processes that occur on a geologic time scale. The often-used assumption that the activity of a species is either unity or equal to the overall concentration of the metal can also yield misleading results. The associate species model, a recent development in thermochemical modeling, will be applied to these systems to more accurately predict chemical activities in such complex systems as waste glasses

  6. Identification of glass compositions suitable for disposal of waste reactive metal

    International Nuclear Information System (INIS)

    Varma, R.; Brown, A.P.; Kumar, R.; San-Pedro, R.; Freeman, C.J.; Helt, J.E.

    1988-09-01

    This study was conducted in support of a project to convert waste sodium to a form that is amenable to easy disposal in ordinary landfills. This waste sodium will be from reactor and other operations at the US Department of Energy and will contain small amounts of radioactive species that must not be released to the environment in an uncontrolled manner. The sodium will be converted into a glass that will contain and isolate the radionuclides present in it. This study was conducted to define acceptable glass compositions that (1) are resistant to leaching of sodium by groundwater and rainwater, (2) contain a relatively large proportion of sodium so that unreasonably large volumes of the glass for disposal will not be produced, and (3) are conveniently prepared from the waste sodium. For this purpose, glass samples containing varying amounts of the oxides of sodium, calcium, boron, aluminum, and silicon were prepared in the laboratory. The samples were subjected to the accelerated MCC-1 test to determine resistance to leaching by water at 60/degree/C. Soda-silica glasses were observed to dissolve in the water rather quickly. Addition of the other ingredients was found to impart significant leach resistance to the glasses. Among the high-Na 2 O glasses, those containing alumina (3% Al 2 O 3 -10% CaO-30% Na 2 O and 6% Al 2 O 3 -10% B 2 O 3 -30% Na 2 O) were found to be most resistant to leaching. Lowering the Na 2 O content to 20% made these glasses even more leach resistant. 8 refs., 6 figs., 2 tabs

  7. 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

  8. Radiation effects in vitreous and devitrified simulated waste glass

    International Nuclear Information System (INIS)

    Weber, W.J.; Turcotte, R.P.; Bunnell, L.R.; Roberts, F.P.; Westsik, J.H. Jr.

    1979-01-01

    The long-term radiation stability of vitreous and partially devitrified forms of high-level waste glass was investigated in accelerated experiments by 266 Cm doping. The effects of radiation on microstructure, phase behavior, density, impact strength, stored energy, and leachability are reported to a cumulative radiation dose of 5 x 10 18 α decays/cm 3 . This dose produces saturation of radiation effects in most properties. 4 figures

  9. Measurement of Solute Diffusion Behavior in Fractured Waste Glass Media

    International Nuclear Information System (INIS)

    Saripalli, Kanaka P.; Lindberg, Michael J.; Meyer, Philip D.

    2008-01-01

    Determination of aqueous phase diffusion coefficients of solutes through fractured media is essential for understanding and modeling contaminants transport at many hazardous waste disposal sites. No methods for earlier measurements are available for the characterization of diffusion in fractured glass blocks. We report here the use of time-lag diffusion experimental method to assess the diffusion behavior of three different solutes (Cs, Sr and Pentafluoro Benzoic Acid or PFBA) in fractured, immobilized low activity waste (ILAW) glass forms. A fractured media time-lag diffusion experimental apparatus that allows the measurement of diffusion coefficients has been designed and built for this purpose. Use of time-lag diffusion method, a considerably easier experimental method than the other available methods, was not previously demonstrated for measuring diffusion in any fractured media. Hydraulic conductivity, porosity and diffusion coefficients of a solute were experimentally measured in fractured glass blocks using this method for the first time. Results agree with the range of properties reported for similar rock media earlier, indicating that the time-lag experimental method can effectively characterize the diffusion coefficients of fractured ILAW glass media

  10. 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

  11. 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

  12. High-level waste solidification: why we chose glass

    International Nuclear Information System (INIS)

    Grover, J.R.

    1980-01-01

    This paper considers the desirable properties and factors to be assessed in the selection of a solidified waste product, surveys the possible product options and then analyzes in detail their suitability in meeting the criteria. It concludes that glasses are currently the preferred choice for the following reasons: their ability to fix the full spectrum of elements contained in the waste; their tolerance of the composition variations that will occur on a day to day basis in practice; their relatively low formation temperatures that lead to simpler and hence safer processing; their radiation stability; and their adequate leach rates. Suitable compositions are available for the wastes that will arise in the UK and techniques are available for manufacture on a production scale. Lower leach rates might be obtained by choosing glasses with higher formation temperatures or ceramics. However, these latter generally also have higher formation temperatures, have less tolerance for composition variations and their radiation stability is unproven. Supercalcines and synthetic rocks (SYNROC) may eventually be demonstrated to have some advantageous properties, but present indications are that these could be major disadvantages which more than offset any gains. Other advanced concepts (for example, the dispersion of glass beads in a metal matrix) have lower leach rates, but lead to additional complexity in manufacture

  13. 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.

  14. 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.

  15. 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

  16. 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.)

  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. 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

  19. Disposition of actinides released from high-level waste glass

    International Nuclear Information System (INIS)

    Ebert, W.L.; Bates, J.K.; Buck, E.C.; Gong, M.; Wolf, S.F.

    1994-01-01

    The disposition of actinide elements released from high-level waste glasses into a tuff groundwater in laboratory tests at 90 degrees C at various glass surface area/leachant volume ratios (S/V) between dissolved, suspended, and sorbed fractions has been measured. While the maximum release of actinides is controlled by the corrosion rate of the glass matrix, their solubility and sorption behavior affects the amounts present in potentially mobile phases. Actinide solubilities are affected by the solution pH and the presence of complexants released from the glass, such as sulfate, phosphate, and chloride, radiolytic products, such as nitrate and nitrite, and carbonate. Sorption onto inorganic colloids formed during lass corrosion may increase the amounts of actinides in solution, although subsequent sedimentation of these colloids under static conditions leads to a significant reduction in the amount of actinides in solution. The solution chemistry and observed actinide behavior depend on the S/V of the test. Tests at high S/V lead to higher pH values, greater complexant concentrations, and generate colloids more quickly than tests at low S/V. The S/V also affects the rate of glass corrosion

  20. Dense and porous glass and glass ceramics from natural and waste raw materials

    OpenAIRE

    Marangoni, Mauro

    2016-01-01

    The main goal of the herewith presented research activities was to develop innovative processes and materials for building applications adapted to the needs of Saudi Arabia according to the information exchanged with the partners from KACST (King Abdulaziz City of Science and Technology). The research activity focused on the development of a wide range of ceramic components via sinter-crystallization of glasses produced from waste (fly ash, slag, sludge) with or without the addition of vit...

  1. 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

  2. Silicate Based Glass Formulations for Immobilization of U.S. Defense Wastes Using Cold Crucible Induction Melters

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Gary L.; Kim, Dong-Sang; Schweiger, Michael J.; Marra, James C.; Lang, Jesse B.; Crum, Jarrod V.; Crawford, Charles L.; Vienna, John D.

    2014-05-22

    The cold crucible induction melter (CCIM) is an alternative technology to the currently deployed liquid-fed, ceramic-lined, Joule-heated melter for immobilizing of U.S. tank waste generated from defense related reprocessing. In order to accurately evaluate the potential benefits of deploying a CCIM, glasses must be developed specifically for that melting technology. Related glass formulation efforts have been conducted since the 1990s including a recent study that is first documented in this report. The purpose of this report is to summarize the silicate base glass formulation efforts for CCIM testing of U.S. tank wastes. Summaries of phosphate based glass formulation and phosphate and silicate based CCIM demonstration tests are reported separately (Day and Ray 2013 and Marra 2013, respectively). Combined these three reports summarize the current state of knowledge related to waste form development and process testing of CCIM technology for U.S. tank wastes.

  3. 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

  4. Influence of the waste glass in the axial compressive strength of Portland cement concrete

    International Nuclear Information System (INIS)

    Miranda Junior, E.J.P.; Paiva, A.E.M.

    2012-01-01

    In this work, was studied the influence of the incorporation of waste glass, coming from the stage of thinning and polishing of a company of thermal glass treatments, in the axial compressive strength of Portland cement concrete. The coarse and ground aggregates used was crushed stone and sand, respectively. For production of the concrete, percentages of glass residues of 5%, 10% and 20% had been used in substitution to the sand, and relations water/cement (a/c) 0,50, 0,55 and 0,58. The cure of the test bodies was carried through in 7, 14 and 28 days. The statistics analysis of the results was carried out through of the analysis of variance for each one of the cure times. From the results of the compressive strength of the concrete, it could be observed that the concrete has structural application for the relation a/c 0,5, independently of waste glass percentage used, and for the relation a/c 0,55 with 20% of waste glass. (author)

  5. Progress toward bridging from atomistic to continuum modeling to predict nuclear waste glass dissolution.

    Energy Technology Data Exchange (ETDEWEB)

    Zapol, Peter (Argonne National Laboratory, Argonne, IL); Bourg, Ian (Lawrence Berkeley National Laboratories, Berkeley, CA); Criscenti, Louise Jacqueline; Steefel, Carl I. (Lawrence Berkeley National Laboratories, Berkeley, CA); Schultz, Peter Andrew

    2011-10-01

    This report summarizes research performed for the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Subcontinuum and Upscaling Task. The work conducted focused on developing a roadmap to include molecular scale, mechanistic information in continuum-scale models of nuclear waste glass dissolution. This information is derived from molecular-scale modeling efforts that are validated through comparison with experimental data. In addition to developing a master plan to incorporate a subcontinuum mechanistic understanding of glass dissolution into continuum models, methods were developed to generate constitutive dissolution rate expressions from quantum calculations, force field models were selected to generate multicomponent glass structures and gel layers, classical molecular modeling was used to study diffusion through nanopores analogous to those in the interfacial gel layer, and a micro-continuum model (K{mu}C) was developed to study coupled diffusion and reaction at the glass-gel-solution interface.

  6. Material interactions relating to long-term geologic disposal of nuclear waste glass

    International Nuclear Information System (INIS)

    Bibler, N.E.; Jantzen, C.M.

    1986-01-01

    This review paper systematizes the additional interactions that materials in a geologic repository will impose on the borosilicate glass waste form-groundwater interactions. These materials are the steel canister that holds the glass, the steel overpack over the canister, backfill materials that may be used, and last, the repository host rock. 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 interaction(s) is examined. The interactions are evaluated based on their effect on the release and speciation of various elements including radionuclides from the glass. Repository relevant interactions testing that requires further study before long-term predictions can be made are noted. 62 refs

  7. Characteristics of colloids generated during the corrosion of nuclear waste glasses in groundwater

    International Nuclear Information System (INIS)

    Feng, X.; Buck, E.C.; Mertz, C.; Bates, J.K.; Cunnane, J.C.; Chaiko, D.

    1993-10-01

    Aqueous colloidal suspensions were generated by reacting nuclear waste glasses with groundwater at 90 degrees C at different ratios of the glass surface area to solution volume (S/V). The colloids have been characterized in terms of size, charge, identity, and stability with respect to salt concentration, pH, and time, by examination using dynamic light scattering, electrophoretic mobility, and transmission electron microscopy. The colloids are predominately produced by precipitation from solution, possibly with contribution from reacted layers that have spallated from the glass. These colloids are silicon-rich minerals. The colloidal suspensions agglomerate when the salinity of the solutions increase. The following implications for modeling the colloidal transport of contaminants have been derived from this study: (1) The sources of the colloids are not only solubility-limited real colloids and the pseudo colloids formed by adsorption of radionuclides onto a groundwater colloid, but also from the spalled surface layers of reacted waste glasses. (2) In a repository, the local environment is likely to be glass-reaction dominated and the salt concentration is likely to be high, leading to rapid colloid agglomeration and settling; thus, colloid transport may be insignificant. (3) If large volumes of groundwater contact the glass reaction site, the precipitated colloids may become resuspended, and colloid transport may become important. (4) Under most conditions, the colloids are negatively charged and will deposit readily on positively charged surfaces. Negatively charged surfaces will, in general, facilitate colloid stability and transport

  8. FTIR spectra and properties of iron borophosphate glasses containing simulated nuclear wastes

    Science.gov (United States)

    Liao, Qilong; Wang, Fu; Chen, Kuiru; Pan, Sheqi; Zhu, Hanzhen; Lu, Mingwei; Qin, Jianfa

    2015-07-01

    30 wt.% simulated nuclear wastes were successfully immobilized by B2O3-doped iron phosphate base glasses. The structure and thermal stability of the prepared wasteforms were characterized by Fourier transform infrared spectroscopy and differential thermal analysis, respectively. The subtle structural variations attributed to different B2O3 doping modes have been discussed in detail. The results show that the thermal stability and glass forming tendency of the iron borophosphate glass wasteforms are faintly affected by different B2O3 doping modes. The main structural networks of iron borophosphate glass wasteforms are PO43-, P2O74-, [BO4] groups. Furthermore, for the wasteform prepared by using 10B2O3-36Fe2O3-54P2O5 as base glass, the distributions of Fe-O-P bonds, [BO4], PO43- and P2O74- groups are optimal. In general, the dissolution rate (DR) values of the studied iron borophosphate wasteforms are about 10-8 g cm-2 min-1. The obtained conclusions can offer some useful information for the disposal of high-level radioactive wastes using boron contained phosphate glasses.

  9. The composition effect on the long-term corrosion of high-level waste glass

    Energy Technology Data Exchange (ETDEWEB)

    Hrma, P. [Pacific Northwest National Laboratory, Richland, Washington (United States)

    1997-07-01

    Waste glass can be optimized for long-term corrosion behavior if the key parameters that control the rate of corrosion are identified, measured, and modeled as functions of glass composition. Second-order polynomial models have been used to optimize glass with respect to a set of requirements on glass properties, such as viscosity and outcomes of standard corrosion tests. Extensive databases exist for the 7-day Product Consistency Test and the 28-day Materials Characterization Center tests, which have been used for nuclear waste glasses in the United States. Models based on these tests are reviewed and discussed to demonstrate the compositional effects on the extent of corrosion under specified conditions. However, modeling the rate of corrosion is potentially more useful for predicting long-term behavior than modeling the extent of corrosion measured by standard tests. Based on an experimental study of two glasses, it is shown that the rate of corrosion can be characterized by simple functions with physically meaningful coefficients. (author)

  10. Matt waste from glass separated collection: an eco-sustainable addition for new building materials.

    Science.gov (United States)

    Bignozzi, M C; Saccani, A; Sandrolini, F

    2009-01-01

    Matt waste (MW), a by-product of purification processes of cullet derived from separated glass waste collection, has been studied as filler for self-compacting concrete and as an addition for newly blended cement. Properties of self-compacting concrete compared to reference samples are reported. They include characteristics at the fresh and hardened states, and the compressive strength and porosity of mortar samples that were formulated with increasing amounts of MW to be used as cement replacement (up to 50wt.%). The effects of matt waste are discussed with respect to the mechanical and microstructural characteristics of the resulting new materials.

  11. Compositional Dependence of Solubility/Retention of Molybdenum Oxides in Aluminoborosilicate-Based Model Nuclear Waste Glasses.

    Science.gov (United States)

    Brehault, Antoine; Patil, Deepak; Kamat, Hrishikesh; Youngman, Randall E; Thirion, Lynn M; Mauro, John C; Corkhill, Claire L; McCloy, John S; Goel, Ashutosh

    2018-02-08

    Molybdenum oxides are an integral component of the high-level waste streams being generated from the nuclear reactors in several countries. Although borosilicate glass has been chosen as the baseline waste form by most of the countries to immobilize these waste streams, molybdate oxyanions (MoO 4 2- ) exhibit very low solubility (∼1 mol %) in these glass matrices. In the past three to four decades, several studies describing the compositional and structural dependence of molybdate anions in borosilicate and aluminoborosilicate glasses have been reported in the literature, providing a basis for our understanding of fundamental science that governs the solubility and retention of these species in the nuclear waste glasses. However, there are still several open questions that need to be answered to gain an in-depth understanding of the mechanisms that control the solubility and retention of these oxyanions in glassy waste forms. This article is focused on finding answers to two such questions: (1) What are the solubility and retention limits of MoO 3 in aluminoborosilicate glasses as a function of chemical composition? (2) Why is there a considerable increase in the solubility of MoO 3 with incorporation of rare-earth oxides (for example, Nd 2 O 3 ) in aluminoborosilicate glasses? Accordingly, three different series of aluminoborosilicate glasses (compositional complexity being added in a tiered approach) with varying MoO 3 concentrations have been synthesized and characterized for their ability to accommodate molybdate ions in their structure (solubility) and as a glass-ceramic (retention). The contradictory viewpoints (between different research groups) pertaining to the impact of rare-earth cations on the structure of aluminoborosilicate glasses are discussed, and their implications on the solubility of MoO 3 in these glasses are evaluated. A novel hypothesis explaining the mechanism governing the solubility of MoO 3 in rare-earth containing aluminoborosilicate

  12. Recycling of waste automotive laminated glass and valorization of polyvinyl butyral through mechanochemical separation

    International Nuclear Information System (INIS)

    Swain, Basudev; Ryang Park, Jae; Yoon Shin, Dong; Park, Kyung-Soo; Hwan Hong, Myung; Gi Lee, Chan

    2015-01-01

    Due to strong binding, optical clarity, adhesion to many surfaces, toughness and flexibility polyvinyl butyral (PVB) resin films are commonly used in the automotive and architectural application as a protective interlayer in the laminated glass. Worldwide million tons of PVB waste generated from end-of-life automotive associated with various environmental issues. Stringent environmental directive, higher land cost eliminates land filling option, needs a study, we have developed a mechanochemical separation process to separate PVB resins from glass and characterized the separated PVB through various techniques, i.e., scanning electron microscope (SEM), energy-dispersive X-ray spectroscopy (EDS), infrared spectroscopy (IR) and nuclear magnetic resonance spectroscopy (NMR). Commercial nonionic surfactants D201 used for the mechanochemical separation purpose. Through parameter optimization following conditions are considered to be the optimum condition; 30 vol% D201, stirring speed of 400 rpm, 35 °C temperature, operation time 1 h, and dilute D201 volume to waste automotive laminated glass weight ratio of ≈25. The technology developed in our laboratory is sustainable, environmentally friendly, techno-economical feasible process, capable of mass production (recycling). - Highlights: • Waste automotive laminated glass and polyvinyl butyral mechanochemically separated. • An economical total recovery and environment-friendly process has been developed. • It is a global problem rather than regional environmental issue has been addressed. • Without using hazardous chemical wastes are being converted to a wealth.

  13. Glass as a matrix for SRP high-level defense waste

    International Nuclear Information System (INIS)

    Wiley, J.R.; Bibler, N.E.; Dukes, M.D.; Plodinec, M.J.

    1980-01-01

    Work done at Savannah River Laboratory and elsewhere that has led to development of glass as a candidate for solidifying Savannah River Plant waste is summarized. Areas of development described are glass formulation and fabrication, and leaching and radiation effects

  14. Structure and properties of rare earth-rich glassed for nuclear waste immobilisation

    International Nuclear Information System (INIS)

    Bardez, I.

    2004-11-01

    A new nuclear glass composition, able to immobilize highly radioactive liquid wastes from high burn-up UO 2 fuel, was established and its structure studied. The composition of the selected rare earth-rich glass is (molar %): 61.79 SiO 2 - 8.94 B 2 O 3 - 3.05 Al 2 O 3 - 14.41 Na 2 O - 6.32 CaO - 1.89 ZrO 2 - 3.60 RE 2 O 3 (with RE = La, Ce, Pr and Nd). The aim of this study was to determine the local environment of the rare earth in this glass and also to glean information about the effect of glass composition on the rare earth neighbouring (influence of Si, B, Al, Na and Ca contents). To this end, several series of glasses, prepared from the baseline glass, were studied by different characterisation methods such as EXAFS spectroscopy at the neodymium LIII-edge, optical absorption spectroscopy, Raman spectroscopy and 29 Si, 27 Al and 11 B MAS-NMR. By coupling all the results obtained, several hypotheses about the nature of the rare earth neighbouring in the glass were proposed. (author)

  15. 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.

  16. Analytical electron microscopy study of surface layers formed on the French SON68 nuclear waste glass during vapor hydration at 200 C

    International Nuclear Information System (INIS)

    Gong, W.L.; Wang, L.M.; Ewing, R.C.; Bates, J.K.; Ebert, W.L.

    1998-01-01

    Extensive solid-state characterization (AEM/SEM/HRTEM) was completed on six SON68 (inactive R7T7) waste glasses which were altered in the presence of saturated water vapor (200 C) for 22, 91, 241, 908, 1000, 1013, and 1021 days. The samples were examined by AEM in cross-section (lattice-fringe imaging, micro-diffraction, and quantitative thin-film EDS analysis). The glass monoliths were invariably covered by a thin altered rind, and the surface layer thickness increased with increasing time of reaction, ranging from 0.5 to 30 μm in thickness. Six distinctive zones, based on phase chemistry and microstructure, were distinguished within the well-developed surface layers. Numerous crystalline phases such as analcime, gyrolite, tobermorite, apatite, and weeksite were identified on the surfaces of the reacted glasses as precipitates. The majority of the surface layer volume was composed of two basic structures that are morphologically and chemically distinct: The A-domain consisted of well-crystallized fibrous smectite aggregates; and the B-domain consisted of poorly-crystallized regions containing smectite, possibly montmorillonite, crystallites and a ZrO 2 -rich amorphous silica matrix. The retention of the rare-earth elements, Mo, and Zr mostly occurred within the B-domain; while transition metal elements, such as Zn, Cr, Ni, Mn, and Fe, were retained in the A-domain. The element partitioning among A-domains and B-domains and recrystallization of the earlier-formed B-domains into the A-domain smectites were the basic processes which have controlled the chemical and structural evolution of the surface layer. The mechanism of surface layer formation during vapor hydration are discussed based on these cross-sectional AEM results. (orig.)

  17. Evaluation of final waste forms and recommendations for baseline alternatives to group and glass

    Energy Technology Data Exchange (ETDEWEB)

    Bleier, A.

    1997-09-01

    An assessment of final waste forms was made as part of the Federal Facilities Compliance Agreement/Development, Demonstration, Testing, and Evaluation (FFCA/DDT&E) Program because supplemental waste-form technologies are needed for the hazardous, radioactive, and mixed wastes of concern to the Department of Energy and the problematic wastes on the Oak Ridge Reservation. The principal objective was to identify a primary waste-form candidate as an alternative to grout (cement) and glass. The effort principally comprised a literature search, the goal of which was to establish a knowledge base regarding four areas: (1) the waste-form technologies based on grout and glass, (2) candidate alternatives, (3) the wastes that need to be immobilized, and (4) the technical and regulatory constraints on the waste-from technologies. This report serves, in part, to meet this goal. Six families of materials emerged as relevant; inorganic, organic, vitrified, devitrified, ceramic, and metallic matrices. Multiple members of each family were assessed, emphasizing the materials-oriented factors and accounting for the fact that the two most prevalent types of wastes for the FFCA/DDT&E Program are aqueous liquids and inorganic sludges and solids. Presently, no individual matrix is sufficiently developed to permit its immediate implementation as a baseline alternative. Three thermoplastic materials, sulfur-polymer cement (inorganic), bitumen (organic), and polyethylene (organic), are the most technologically developed candidates. Each warrants further study, emphasizing the engineering and economic factors, but each also has limitations that regulate it to a status of short-term alternative. The crystallinity and flexible processing of sulfur provide sulfur-polymer cement with the highest potential for short-term success via encapsulation. Long-term immobilization demands chemical stabilization, which the thermoplastic matrices do not offer. Among the properties of the remaining

  18. Evaluation of final waste forms and recommendations for baseline alternatives to grout and glass

    International Nuclear Information System (INIS)

    Bleier, A.

    1997-09-01

    An assessment of final waste forms was made as part of the Federal Facilities Compliance Agreement/Development, Demonstration, Testing, and Evaluation (FFCA/DDT ampersand E) Program because supplemental waste-form technologies are needed for the hazardous, radioactive, and mixed wastes of concern to the Department of Energy and the problematic wastes on the Oak Ridge Reservation. The principal objective was to identify a primary waste-form candidate as an alternative to grout (cement) and glass. The effort principally comprised a literature search, the goal of which was to establish a knowledge base regarding four areas: (1) the waste-form technologies based on grout and glass, (2) candidate alternatives, (3) the wastes that need to be immobilized, and (4) the technical and regulatory constraints on the waste-from technologies. This report serves, in part, to meet this goal. Six families of materials emerged as relevant; inorganic, organic, vitrified, devitrified, ceramic, and metallic matrices. Multiple members of each family were assessed, emphasizing the materials-oriented factors and accounting for the fact that the two most prevalent types of wastes for the FFCA/DDT ampersand E Program are aqueous liquids and inorganic sludges and solids. Presently, no individual matrix is sufficiently developed to permit its immediate implementation as a baseline alternative. Three thermoplastic materials, sulfur-polymer cement (inorganic), bitumen (organic), and polyethylene (organic), are the most technologically developed candidates. Each warrants further study, emphasizing the engineering and economic factors, but each also has limitations that regulate it to a status of short-term alternative. The crystallinity and flexible processing of sulfur provide sulfur-polymer cement with the highest potential for short-term success via encapsulation. Long-term immobilization demands chemical stabilization, which the thermoplastic matrices do not offer. Among the properties of the

  19. 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)

  20. Glass melter assembly for the Hanford Waste Vitrification Plant

    International Nuclear Information System (INIS)

    Chen, A.E.; Russell, A.; Shah, K.R.; Kalia, J.

    1993-01-01

    The Hanford Waste Vitrification Plant (HWVP) is designed to solidify high level radioactive waste by converting it into stable borosilicate after mixing with glass frit and water. The heart of this conversion process takes place in the glass melter. The life span of the existing melter is limited by the possible premature failure of the heater assembly, which is not remotely replaceable, in the riser and pour spout. A goal of HWVP Project is to design remotely replaceable riser and pour spout heaters so that the useful life of the melter can be prolonged. The riser pour spout area is accessible only by the canyon crane and impact wrench. It is also congested with supporting frame members, service piping, electrode terminals, canister positioning arm and other various melter components. The visibility is low and the accessibility is limited. The problem is further compounded by the extreme high temperature in the riser core and the electrical conductive nature of the molten glass that flows through it

  1. Nucleation and crystal growth behavior of nepheline in simulated high-level waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Amoroso, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Mcclane, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-09-26

    The Savannah River National Laboratory (SRNL) has been tasked with supporting glass formulation development and process control strategies in key technical areas, relevant to the Department of Energy’s Office of River Protection (DOE-ORP) and related to high-level waste (HLW) vitrification at the Waste Treatment and Immobilization Plant (WTP). Of specific interest is the development of predictive models for crystallization of nepheline (NaAlSiO4) in HLW glasses formulated at high alumina concentrations. This report summarizes recent progress by researchers at SRNL towards developing a predicative tool for quantifying nepheline crystallization in HLW glass canisters using laboratory experiments. In this work, differential scanning calorimetry (DSC) was used to obtain the temperature regions over which nucleation and growth of nepheline occur in three simulated HLW glasses - two glasses representative of WTP projections and one glass representative of the Defense Waste Processing Facility (DWPF) product. The DWPF glass, which has been studied previously, was chosen as a reference composition and for comparison purposes. Complementary quantitative X-ray diffraction (XRD) and optical microscopy confirmed the validity of the methodology to determine nucleation and growth behavior as a function of temperature. The nepheline crystallization growth region was determined to generally extend from ~ 500 to >850 °C, with the maximum growth rates occurring between 600 and 700 °C. For select WTP glass compositions (high Al2O3 and B2O3), the nucleation range extended from ~ 450 to 600 °C, with the maximum nucleation rates occurring at ~ 530 °C. For the DWPF glass composition, the nucleation range extended from ~ 450 to 750 °C with the maximum nucleation rate occurring at ~ 640 °C. The nepheline growth at the peak temperature, as determined by XRD, was between 35 - 75 wt.% /hour. A maximum nepheline growth rate of ~ 0.1 mm/hour at 700 °C was measured for the DWPF

  2. Nucleation and crystal growth behavior of nepheline in simulated high-level waste glasses

    International Nuclear Information System (INIS)

    Fox, K.; Amoroso, J.; Mcclane, D.

    2017-01-01

    The Savannah River National Laboratory (SRNL) has been tasked with supporting glass formulation development and process control strategies in key technical areas, relevant to the Department of Energy's Office of River Protection (DOE-ORP) and related to high-level waste (HLW) vitrification at the Waste Treatment and Immobilization Plant (WTP). Of specific interest is the development of predictive models for crystallization of nepheline (NaAlSiO4) in HLW glasses formulated at high alumina concentrations. This report summarizes recent progress by researchers at SRNL towards developing a predicative tool for quantifying nepheline crystallization in HLW glass canisters using laboratory experiments. In this work, differential scanning calorimetry (DSC) was used to obtain the temperature regions over which nucleation and growth of nepheline occur in three simulated HLW glasses - two glasses representative of WTP projections and one glass representative of the Defense Waste Processing Facility (DWPF) product. The DWPF glass, which has been studied previously, was chosen as a reference composition and for comparison purposes. Complementary quantitative X-ray diffraction (XRD) and optical microscopy confirmed the validity of the methodology to determine nucleation and growth behavior as a function of temperature. The nepheline crystallization growth region was determined to generally extend from ~ 500 to >850 °C, with the maximum growth rates occurring between 600 and 700 °C. For select WTP glass compositions (high Al2O3 and B2O3), the nucleation range extended from ~ 450 to 600 °C, with the maximum nucleation rates occurring at ~ 530 °C. For the DWPF glass composition, the nucleation range extended from ~ 450 to 750 °C with the maximum nucleation rate occurring at ~ 640 °C. The nepheline growth at the peak temperature, as determined by XRD, was between 35 - 75 wt.% /hour. A maximum nepheline growth rate of ~ 0.1 mm/hour at 700 °C was measured for the DWPF

  3. 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

  4. Structural and crystallisation study of a rare earth alumino borosilicate glass designed for nuclear waste confinement; Etude de la structure et du comportement en cristallisation d'un verre nucleaire d'aluminoborosilicate de terre rare

    Energy Technology Data Exchange (ETDEWEB)

    Quintas, A

    2007-09-15

    This work is devoted to the study of a rare earth alumino borosilicate glass, which molar composition is 61,81 SiO{sub 2} - 3,05 Al{sub 2}O{sub 3} - 8,94 B{sub 2}O{sub 3} - 14,41 Na{sub 2}O - 6,33 CaO - 1,90 ZrO{sub 2} - 3,56 Nd{sub 2}O{sub 3}, and envisaged for the immobilization of nuclear wastes originating from the reprocessing of high discharge burn up spent fuel. From a structural viewpoint, we investigated the role of the modifier cations on the arrangement of the glass network through different modifications of the glass composition: variation of the Na/Ca ratio and modification of the nature of the alkali and alkaline earth cations. The NMR and Raman spectroscopic techniques were useful to determine the distribution of modifier cations among the glass network and also to cast light on the competition phenomena occurring between alkali and alkaline earth cations for charge compensation of [AlO{sub 4}]{sup -} and [BO{sub 4}]{sup -} species. The neodymium local environment could be probed by optical absorption and EXAFS spectroscopies which enabled to better understand the insertion mode of Nd{sup 3+} ions among the silicate domains of the glass network. Concerning the crystallization behavior we were interested in how the glass composition may influence the crystallization processes and especially the formation of the apatite phase of composition Ca{sub 2}Nd{sub 8}(SiO{sub 4}){sub 6}O{sub 2}. In particular, this work underlined the important role of both alkaline earth and rare earth cations on the crystallization of the apatite phase. (author)

  5. Radionuclide decay effects on waste glass corrosion and weathering

    International Nuclear Information System (INIS)

    Wronkiewicz, D.J.

    1993-01-01

    The release of glass components into solution, including radionuclides, may be influenced by the presence of radiolytically produced nitric acid, carboxylic acid, and transient water dissociation products such as ·OH and O 2 - . Under batch test conditions, glass corrosion has been shown to increase up to a maximum of three-to five-fold in irradiated tests relative to nonirradiated tests, while in other studies the presence of radiolytic products has actually decreased glass corrosion rates. Bicarbonate groundwaters will buffer against pH decreases and changes in corrosion rates. Under high surface area-to-solution volume (S/V) conditions, the bicarbonate buffering reservoir may be quickly overwhelmed by radiolytic acids that are concentrated in the thin films of water contacting the samples. Glass reaction rates have been shown to increase up to 10-to-15-fold due to radiation exposure under high S/V conditions. Radiation damage to solid glass materials results in bond damage and atomic displacements. This type of damage has been shown to increase the release rates of glass components up to four-fold during subsequent corrosion tests, although under actual disposal conditions, glass annealing processes may negate the solid radiation damage effects

  6. Development and radiation stability of glasses for highly radioactive wastes

    International Nuclear Information System (INIS)

    Hall, A.R.; Dalton, J.T.; Hudson, B.; Marples, J.A.C.

    1976-01-01

    The variation of formation temperature, crystallizing behaviour and leach resistance with composition changes for sodium-lithium borosilicate glasses suitable for vitrifying Magnox waste are discussed. Viscosities have been measured between 400 and 1050 0 C. The principal crystal phases which occur have been identified as magnesium silicate, magnesium borate and ceria. The leach rate of polished discs in pure water at 100 0 C does not decrease with time if account is taken of the fragile siliceous layer that is observed to occur. The effect of 100 years' equivalent α- and β-irradiation on glass properties is discussed. Stored energy release experiments demonstrated that energy is released over a wide temperature range so that it cannot be triggered catastrophically. Temperatures required to release energy are dependent upon the original storage temperature. Helium release is by Fick's diffusion law up to at least 30% of the total inventory, with diffusion coefficients similar to those for comparable borosilicate glasses. Leach rates were not measurably affected by α-radiation. β-radiation in a Van de Graaff accelerator did not change physical properties, but irradiation in an electron microscope caused minute bubbles in lithium-containing glasses above 200 0 C. (author)

  7. 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

  8. 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

  9. Wastes vitrification by plasma torch: study of a glass formulation compatible with a wide range of B wastes; Vitrification des dechets par torche a plasma: recherche d'une formulation de verre compatible avec un large eventail de dechets B

    Energy Technology Data Exchange (ETDEWEB)

    Poitou, S.; Richaud, D.; Fiquet, O.; Gramondi, P.; Massit, H. [CEA Cadarache, Dept. d' Entreposage et de Stockage des Dechets, 13 - Saint-Paul-lez-Durance (France)

    2001-07-01

    Within the context of radioactive waste management, CEA has equipped itself with a 'PLASMARC' device. The central element of this device is a plasma torch treatment furnace. It has been implemented and validated for the vitrification of low level radioactive wastes. Meanwhile, the plasma torch presents potentially interests for immobilizing under an inert form in vitreous matrices, B wastes which are generally divided and of complex chemical composition. The application of this process to this type of wastes has been studied here. The obtained results show that with the plasma torch it is possible to make glasses with a high amount of silicon and aluminium oxide and which are adapted to the treatment / packaging of the B wastes. (O.M.)

  10. Study of glass alteration mechanisms in cement waters

    International Nuclear Information System (INIS)

    Depierre, S.

    2012-01-01

    In the French deep geological repository concept, intermediate-level vitrified waste packages could be disposed of concrete medium. Chemical composition and pH of the interstitial leaching water are expected to influence the chemical durability of glass. Investigations have thus been carried out to study glass dissolution mechanisms and kinetics in contact with cement waters. Three cement pore waters were studied: the first two correspond to two stages of the Portland cement aging and the third corresponds to equilibrium with a low pH concrete. The S/V ratio (glass-surface-area-to-solution-volume ratio) and the chemistry of cement waters are the two main parameters that control glass alteration mechanisms. If the leaching flow from the glass allows a degree of supersaturation to be reached and maintained which leads to nucleation of secondary phases, then precipitation of these phases drives glass dissolution. At a very low S/V ratio, the calcium uptake into the alteration layer increases its passivating properties. Conversely, at a high S/V ratio, the calcium precipitates as cementitious phases consuming elements which form the alteration layer. The glass dissolution is maintained at a high rate. This study contributes to highlighting the beneficial role of low pH cement in glass alteration, and is a first step towards understanding the mechanisms between the glass and the cement medium. (author) [fr

  11. 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.

  12. 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

  13. 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

  14. Application of glass-nonmetals of waste printed circuit boards to produce phenolic moulding compound

    International Nuclear Information System (INIS)

    Guo Jie; Rao Qunli; Xu Zhenming

    2008-01-01

    The aim of this study was to investigate the feasibility of using glass-nonmetals, a byproduct of recycling waste printed circuit boards (PCBs), to replace wood flour in production of phenolic moulding compound (PMC). Glass-nonmetals were attained by two-step crushing and corona electrostatic separating processes. Glass-nonmetals with particle size shorter than 0.07 mm were in the form of single fibers and resin powder, with the biggest portion (up to 34.6 wt%). Properties of PMC with glass-nonmetals (PMCGN) were compared with reference PMC and the national standard of PMC (PF2C3). When the adding content of glass-nonmetals was 40 wt%, PMCGN exhibited flexural strength of 82 MPa, notched impact strength of 2.4 kJ/m 2 , heat deflection temperature of 175 deg. C, and dielectric strength of 4.8 MV/m, all of which met the national standard. Scanning electron microscopy (SEM) showed strong interfacial bonding between glass fibers and the phenolic resin. All the results showed that the use of glass-nonmetals as filler in PMC represented a promising method for resolving the environmental pollutions and reducing the cost of PMC, thus attaining both environmental and economic benefits

  15. 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

  16. Magnetic properties of glasses from geothite industrial wastes recycling (FeOOH)

    International Nuclear Information System (INIS)

    Romero, M.; Rincon, J.M.; Esparza, M.; Gonzalez-Oliver, C.

    1997-01-01

    It has been carried out the magnetic properties determination for high iron oxide content glasses series obtained from a geothite red mud waste from the zinc hydrometallurgy and dolomite and glass cullet as main raw materials. It has been determined the magnetic susceptibility and magnetization values for the glasses here investigated. The results suggest that the magnetic behaviour are depending on the glass chemical composition, so that glasses can be differently classified like ferrimagnetic, ferromagnetic, superparamagnetic and paramagnetic. (Author) 6 refs

  17. Combined Waste Form Cost Trade Study

    International Nuclear Information System (INIS)

    Gombert, Dirk; Piet, Steve; Trickel, Timothy; Carter, Joe; Vienna, John; Ebert, Bill; Matthern, Gretchen

    2008-01-01

    A new generation of aqueous nuclear fuel reprocessing, now in development under the auspices of the DOE Office of Nuclear Energy (NE), separates fuel into several fractions, thereby partitioning the wastes into groups of common chemistry. This technology advance enables development of waste management strategies that were not conceivable with simple PUREX reprocessing. Conventional wisdom suggests minimizing high level waste (HLW) volume is desirable, but logical extrapolation of this concept suggests that at some point the cost of reducing volume further will reach a point of diminishing return and may cease to be cost-effective. This report summarizes an evaluation considering three groupings of wastes in terms of cost-benefit for the reprocessing system. Internationally, the typical waste form for HLW from the PUREX process is borosilicate glass containing waste elements as oxides. Unfortunately several fission products (primarily Mo and the noble metals Ru, Rh, Pd) have limited solubility in glass, yielding relatively low waste loading, producing more glass, and greater disposal costs. Advanced separations allow matching the waste form to waste stream chemistry, allowing the disposal system to achieve more optimum waste loading with improved performance. Metals can be segregated from oxides and each can be stabilized in forms to minimize the HLW volume for repository disposal. Thus, a more efficient waste management system making the most effective use of advanced waste forms and disposal design for each waste is enabled by advanced separations and how the waste streams are combined. This trade-study was designed to juxtapose a combined waste form baseline waste treatment scheme with two options and to evaluate the cost-benefit using available data from the conceptual design studies supported by DOE-NE

  18. 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.

  19. Fabrication and characterization of bioactive glass-ceramic using soda-lime-silica waste glass.

    Science.gov (United States)

    Abbasi, Mojtaba; Hashemi, Babak

    2014-04-01

    Soda-lime-silica waste glass was used to synthesize a bioactive glass-ceramic through solid-state reactions. In comparison with the conventional route, that is, the melt-quenching and subsequent heat treatment, the present work is an economical technique. Structural and thermal properties of the samples were examined by X-ray diffraction (XRD) and differential thermal analysis (DTA). The in vitro test was utilized to assess the bioactivity level of the samples by Hanks' solution as simulated body fluid (SBF). Bioactivity assessment by atomic absorption spectroscopy (AAS) and scanning electron microscopy (SEM) was revealed that the samples with smaller amount of crystalline phase had a higher level of bioactivity. Copyright © 2014 Elsevier B.V. All rights reserved.

  20. 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.

  1. Use of wasted foundry sand (WFS) as a partial substitute for silica in a soda lime glass

    International Nuclear Information System (INIS)

    Martin, A.C.; Ueno, O.K.; Folgueras, M.V.

    2016-01-01

    The waste foundry sand (WFS) is the main waste generates in foundry industries. Studies in the literature suggest the use of WFS in different materials, such as concrete, brick or asphalt. This work aims to partially replace the silica of a soda-lime glass by the WFS. The waste foundry sand has in its composition elements such as iron and aluminum that can affect the glass quality, which justifies the residue processing to reduce the impurity content. The treatments, that included mechanical agitation and thermal treatment, resulted in a slight decrease in the percent of iron with consequent increase of the silica content. After treatment, some sands were incorporated into the glass, that showed green color but with lower absorption intensity for the sand with less iron content. It was observed that it's possible to obtain glasses using WFS, however, there is difficulty in color controlling. (author

  2. 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

  3. Performance of surrogate high-level waste glass in the presence of iron corrosion products

    International Nuclear Information System (INIS)

    Jain, V.; Pan, Y.M.

    2004-01-01

    Radionuclide release from a waste package (WP) is a series of processes that depend upon the composition and flux of groundwater contacting the waste-forms (WF); the corrosion rate of WP containers and internal components made of Alloy 22, 316L SS, 304L SS and carbon steel; the dissolution rate of high-level radioactive waste (HLW) glass and spent nuclear fuel (SNF); the solubility of radionuclides; and the retention of radionuclides in secondary mineral phases. In this study, forward reaction rate measurements were made on a surrogate HLW glass in the presence of FeCl 3 species. Results indicate that the forward reaction rate increases with an increase in the FeCl 3 concentration. The addition of FeCl 3 causes the drop in the pH due to hydrolysis of Fe 3+ ions in the solution. Results based on the radionuclide concentrations and dissolution rates for HLW glass and SNF indicate that the contribution from glass is similar to SNF at 75 deg C. (authors)

  4. 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

  5. Projected radionuclide inventories of DWPF glass from current waste at time of production

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1993-01-01

    The Waste Acceptance Preliminary Specifications (WAPS) require that the DWPF estimate the inventory of long-lived radionuclides present in the waste glass, and report the values in the Waste Form Qualification Report. In this report, conservative (biased high) estimates of the radionuclide inventory of glass produced from waste currently in the Tank Farm are provided. In most cases, these calculated values compare favorably with actual data. In those cases where the agreement is not good, the values reported here are conservative

  6. Influence of Waste Glass Powder Addition on the Pore Structure and Service Properties of Cement Mortars

    Directory of Open Access Journals (Sweden)

    José Marcos Ortega

    2018-03-01

    Full Text Available At present, reusing waste constitutes an important challenge in order to reach a more sustainable environment. The cement industry is an important pollutant industrial sector. Therefore, the reduction of its CO2 emissions is now a popular topic of study. One way to lessen those emissions is partially replacing clinker with other materials. In this regard, the reuse of waste glass powder as a clinker replacement could be possible. This is a non-biodegradable residue that permanently occupies a large amount of space in dumping sites. The aim of this work is to study the long-term effects (400 days of the addition of waste glass powder on the microstructure and service properties of mortars that incorporate up to 20% of this addition as clinker replacement. The microstructure has been characterised using the non-destructive impedance spectroscopy technique and mercury intrusion porosimetry. Furthermore, differential thermal analysis was also performed. Compressive strength and both steady-state and non-steady-state chloride diffusion coefficients have also been determined. Considering the obtained results, mortars with 10% and 20% waste glass powder showed good service properties until 400 days, similar to or even better than those made with ordinary Portland cement without additions, with the added value of contributing to sustainability.

  7. Long-term leaching behavior of simulated Savannah River Plant waste glass: Part 1, MCC-1 leachability results, four-year leaching data

    International Nuclear Information System (INIS)

    Wicks, G.G.; Stone, J.A.; Chandler, G.T.; Williams, S.

    1986-08-01

    Long-term leaching data were obtained on SRP 131/TDS waste glass using MCC-1 or slightly modified MCC-1 standard leaching tests. Experiments were conducted out to four years at 40 0 C and 3-1/2 years at 90 0 C. These experiments have produced the longest standardized leaching data currently available in the waste management community. Long-term leaching data provide important input to modeling of waste glass behavior and ultimate prediction of waste glass performance. In this study, the leaching behavior of SRP waste glass was found to be excellent; leachates based on a variety of elements were not only very low, but also improved with increasing time. In addition to these data, results are also reported from another independent Savannah River study. Leaching behavior at 40 0 C and 90 0 C was assessed not only for a similar SRP 131 waste glass composition, but also for extreme waste glass compositions involving high-iron and high-aluminum waste. In addition, these experiments were performed using not only a standard deionized water leachant, but also simplified brine and silicate groundwater simulations. These two large data bases will be summarized and correlated along with some of the more interesting results recently reported in another study, a two-year leaching program performed on a similar SRP waste glass composition at Battelle Pacific Northwest Laboratories

  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. Studies and implementation of glass-crystal materials as the ecological shells and coatings for storage and disposal of radioactive and toxic wastes

    International Nuclear Information System (INIS)

    Iskakov, T.; Ibraev, Yu.; Garifulin, V.

    1996-01-01

    The ground space and missile systems designed to launch the modern rockets have the launching constructions. These construction are equipped with protection systems for construction units, which are undergoing by high temperature gas dynamic impact. Protective materials are featuring with a high strength, and stability regarding to high temperatures and heat impact. Such materials were obtained in glass-crystal class. Institute for Special Material Science Research has synthesized the glass-crystal materials with high thermal stability, strength, which are also stable regarding to chemical, biological, radioactive, gas dynamic, etc. impacts. The developed materials in form of plates were successfully tested in launching constructions at Baikonur space system. This material is valid for heat protective coatings, constructions of rocket-space system, plates used in airstrips and grounds for aircraft with vertical take-off and landing. These materials were studied to obtain the super strength, wearing-proof, chemically stable and radiation stable materials to be designed for ecological shells on their base. It was established that some types of glass-crystal materials have the properties, which are valid for application as ecological shells

  10. Corrosion of synthesized glasses and glazes as analogs for nuclear waste glass degradation

    International Nuclear Information System (INIS)

    Vandiver, P.B.

    1994-01-01

    Synthesized glasses provide an opportunity to study natural corrosion processes which are intermediate in time span between geological examples of natural glasses, such as obsidians and tektites, and relatively short term laboratory tests lasting a few hours to several decades. In addition, synthesized glasses can usually be tracked to particular archaeological find sites with known dates of production and often burial. Environmental conditions are routinely measured at archaeological sites as a part of the excavation-process, such that information is available on the yearly cycling of temperature and relative humidity, sometimes at the depth at which the artifact was found. Whether the artifacts were excavated in an air enclosure, such as a tomb, or in the soil can also be reconstructed, such that one can determine whether aqueous or atmospheric corrosion was involved in the degradation process. For instance, so-called open-quotes Roman glassclose quotes may span a time period of production of 800 years and a geographical range from Germany to North Africa and from Britain to Afghanistan. One example is the storage during World War II of glass from the British Museum in underground metro stations. Some of these glasses have been in collections for over 100 years. Thus, populations of glasses can be chosen for experimentation which compare variations in bulk composition, dopants, microstructure, heat treatment, ground vs. fire polished surfaces, aqueous vs. atmospheric corrosion, geographic, geological as well as recent storage conditions. Glasses in museums are generally considered to have had their corrosion arrested and be stable because changes in visual appearance are not obvious. However, if we attempt to measure the range of surface water content in these glasses using Fourier transform infrared analysis, a considerable variability is found, as shown

  11. Leach testing of waste glasses under near-saturation conditions

    International Nuclear Information System (INIS)

    Strachan, D.M.; Grambow, B.

    1983-11-01

    Two waste glasses, MCC 76 to 68 and C31 to 3, were leached in deionized water and 0.001 M MgCl 2 for periods up to 158 days. At 57 days the gel layer was removed from some of the specimens and leaching continued for up to 100 days. Results from leaching in deionized water showed that the gel layer was not protective. Results from leaching in 0.001 M MgCl 2 are in good agreement with the predicted results obtained from the use of the PHREEQE geochemical code and with sepiolite [Mg 2 Si 3 O 6 (OH) 4 ] as the Mg-bearing precipitate. Both B and Si were predicted and observed to increase with increasing glass dissolution while maintaining sepiolite solubility. Both MCC 76 to 68 and C31 to 3 glasses showed increased leaching in 0.001 M MgCl 2 upon removal of the layer. This suggests a leaching mechanism whereby leaching is driven by the formation of an alteration product

  12. Na, Mg, Ni and Cs distribution and speciation after long-term alteration of a simulated nuclear waste glass: A micro-XAS/XRF/XRD and wet chemical study

    Science.gov (United States)

    Curti, Enzo; Dähn, Rainer; Farges, François; Vespa, Marika

    2009-04-01

    Microscopic distribution and speciation of Na, Mg, Ni and Cs in a simulated (inactive) nuclear waste glass were studied using micro X-ray fluorescence (μ-XRF) and micro X-ray absorption spectroscopy (μ-XAS), after aqueous leaching during 12 years at 90 °C. Na and Mg are major constituents of the glass that can be used to determine the progress of the glass corrosion process and the nature of secondary alteration phases. Ni and Cs represent dose determining long-lived radionuclides ( 59Ni, 135Cs) in vitrified nuclear waste. The Na-Mg μ-XRF maps revealed that the core regions of the glass fragments are apparently unaltered and compositionally homogeneous, whereas rims and interstitial spaces are enriched with Mg-rich smectite formed during the leaching process. The micro X-ray absorption near edge structure (μ-XANES) spectra collected at the Mg K-edge in the altered zones show three sharp resonances typical for crystalline Mg-silicates. These resonances are distinctive of Mg occupying undistorted octahedral positions. In contrast, the μ-XANES spectra collected in the core zones of the glass fragments lack this resonance pattern and are identical to the spectra measured on the pristine (unleached) MW glass. Micro extended X-ray absorption fine structure (μ-EXAFS) and μ-XANES analyses at the Ni K-edge revealed three distinct Ni(II) species: (a) Ni uniformly distributed in the glass matrix, (b) micro-inclusions with high Ni concentrations and (c) Ni associated to the Mg-clay. The comparison with reference spectra of unleached MW and other Ni-bearing silicate glasses indicated that species (a) represents the original coordinative environment of Ni in the glass. The μ-EXAFS analyses revealed that species (b) is structural Ni in trevorite (NiFe 2O 4), which probably formed through unmixing processes during the cooling of the glass melt. The μ-EXAFS of species (c) could be successfully modeled assuming specific adsorption or incorporation of Ni into the lattice of

  13. Using of Glass Wastes as a Fine Aggregate in Concrete Mixture

    Directory of Open Access Journals (Sweden)

    Mohammad F. Al-Deen

    2013-04-01

    Full Text Available In this study, the waste glass (WG is considered as a fine aggregate in the concretemixture. WG is used after grinding to size according to Iraqi sand specificationsNo.45. The waste glass has been used instead of sand in different proportions whichare 0%, 33%, 66% and 100%. The effects of WG on compressive strength of theconcrete and unit weight are analysed. As results of this study, WG is determined tohave a significant effect upon the reduction of its compressive strength and there is asignificant decreasing of its unit weight. As for cost analysis, it was determined tolower the cost of concrete production. This study was an environmental one inconsideration of the fact that WG could be used in the concrete as fine aggreagateswithout the need for a high cost or rigorous energy.

  14. Glass ceramic obtained by tailings and tin mine waste reprocessing from Llallagua, Bolivia

    Science.gov (United States)

    Arancibia, Jony Roger Hans; Villarino, Cecilia; Alfonso, Pura; Garcia-Valles, Maite; Martinez, Salvador; Parcerisa, David

    2014-05-01

    In Bolivia Sn mining activity produces large tailings of SiO2-rich residues. These tailings contain potentially toxic elements that can be removed into the surface water and produce a high environmental pollution. This study determines the thermal behaviour and the viability of the manufacture of glass-ceramics from glass. The glass has been obtained from raw materials representative of the Sn mining activities from Llallagua (Bolivia). Temperatures of maximum nucleation rate (Tn) and crystallization (Tcr) were calculated from the differential thermal analyses. The final mineral phases were determined by X-ray diffraction and textures were observed by scanning electron microscopy. Crystalline phases are nefeline occurring with wollastonite or plagioclase. Tn for nepheline is between 680 ºC and 700 ºC, for wollastonite, 730 ºC and for plagioclase, 740 ºC. Tcr for nefeline is between 837 and 965 ºC; for wollastonite, 807 ºC and for plagioclase, 977 ºC. In order to establish the mechanical characteristics and efficiency of the vitrification process in the fixation of potentially toxic elements the resistance to leaching and micro-hardness were determined. The obtained contents of the elements leached from the glass ceramic are well below the limits established by the European legislation. So, these analyses confirm that potentially toxic elements remain fixed in the structure of mineral phases formed in the glass-ceramic process. Regarding the values of micro-hardness results show that they are above those of a commercial glass. The manufacture of glass-ceramics from mining waste reduces the volume of tailings produced for the mining industry and, in turn enhances the waste, transforming it into a product with industrial application. Acknowledgements: This work was partly financed by the project AECID: A3/042750/11, and the SGR 2009SGR-00444.

  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. SRL in-situ tests in the United Kingdom: Part 2, Surface analyses of SRS waste glass buried for one and two years in limestone at Ballidon, UK

    International Nuclear Information System (INIS)

    Namboodri, C.G. Jr.; Wicks, G.G.

    1991-01-01

    A multiphase experimental program to assess and understand waste glass behavior under a wide range of conditions has been in progress at the Savannah River Laboratory (SRL) for over a decade. An important part of this experimental effort is to assess the effects of repository relevant conditions on performance of SRS waste glass, in both controlled laboratory tests, as well as in actual field experiments. In laboratory test, SRS waste glass, simulated and in many cases also fully radioactive, has been tested in environments of salt, basalt, shale, granite, clay and tuff. In field experiments, there are four joint international programs being conducted in four different countries, involving burial of SRS simulated waste glass in granite, limestone, clay and salt geologies. This report discusses the SRS waste glass studies in limestone at Ballidon, UK

  17. 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.

  18. 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

  19. 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.

  20. Alteration of rhyolitic (volcanic) glasses in natural Bolivian salt lakes. - Natural analogue for the behavior of radioactive waste glasses in rock salt repositories

    International Nuclear Information System (INIS)

    Abdelouas, A.

    1996-06-01

    Alteration experiments with the R7T7 glass in three salt brines, saturated respectively in MgCl 2 , MgCl 2 -CaCl 2 and NaCl, showed that the solubilities of most radionuclides are controlled by the secondary phases. Nd, La, and Pr are trapped in powellite, Ce in cerianite, U in coffinite, and Sr is partially immobilized in barite. There is a good similarity between the secondary phases formed experimentally on volcanic glasses and the R7T7 glass altered in MgCl 2 CaCl 2 -saturated brine (formation of hydrotalcite and chlorite-serpentine at short-term and saponite at long-term). These results support the use of volcanic glasses alteration patterns in Mg-rich solutions (seawater, brines) to understand the long-term behavior of nuclear waste glasses and to evaluate the stability of the secondary phases. The study of the sediments of Uyuni (Bolivia) showed that the corrosion rate of the rhyolitic glass in brines at 10 C is 12 to 30 time lower than those of rhyolitic glasses altered in high dilute conditions. The neoformed phases in the sediments are: Smectite, alunite, pyrite, barite, celestite and cerianite. The low alteration rate of rhyolitic glasses in brines and the formation of secondary phases such as smectite, barite and cerianite (also formed during the experimental alteration of the R7T7 glass), permit us to expect the low alteration of nuclear waste glasses at long-term in brines and the trapping of certain radionuclides in secondary phases. (orig.) [de

  1. 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.

  2. 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

  3. Kinetic model for quartz and spinel dissolution during melting of high-level-waste glass batch

    International Nuclear Information System (INIS)

    Pokorny, Richard; Rice, Jarrett A.; Crum, Jarrod V.; Schweiger, Michael J.; Hrma, Pavel

    2013-01-01

    The dissolution of quartz particles and the growth and dissolution of crystalline phases during the conversion of batch to glass potentially affects both the glass melting process and product quality. Crystals of spinel exiting the cold cap to molten glass below can be troublesome during the vitrification of iron-containing high-level wastes. To estimate the distribution of quartz and spinel fractions within the cold cap, we used kinetic models that relate fractions of these phases to temperature and heating rate. Fitting the model equations to data showed that the heating rate, apart from affecting quartz and spinel behavior directly, also affects them indirectly via concurrent processes, such as the formation and motion of bubbles. Because of these indirect effects, it was necessary to allow one kinetic parameter (the pre-exponential factor) to vary with the heating rate. The resulting kinetic equations are sufficiently simple for the detailed modeling of batch-to-glass conversion as it occurs in glass melters. The estimated fractions and sizes of quartz and spinel particles as they leave the cold cap, determined in this study, will provide the source terms needed for modeling the behavior of these solid particles within the flow of molten glass in the melter

  4. 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

  5. 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

    using a high lead content borate glass, or other low melting glass is also recommended for further evaluation and development. Additional laboratory studies of phase behavior and chemical durability of low-temperature glasses is also recommended to provide risk mitigation if one of the primary development paths proves infeasible. This report is a deliverable for the task “Candidate Low-T Glass Waste Forms for EMF Bottoms On-Site Disposition Alternative Option.”

  6. Role of lead as modifier on the properties of lead iron phosphate nuclear waste glasses

    International Nuclear Information System (INIS)

    Hazra, G.; Mitra, P.; Das, T.

    2011-01-01

    Lead-iron phosphate glasses are a promising new waste form for the safe immobilization of both high level defence and high level commercial radioactive waste for long term disposal. Lead iron phosphate glasses have several advantages such as lower aqueous corrosion rate, lower processing temperature etc. (author)

  7. 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

  8. 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

  9. 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

  10. Startup and operation of a plant-scale continuous glass melter for vitrification of Savannah River Plant simulated waste

    International Nuclear Information System (INIS)

    Willis, T.A.

    1980-01-01

    The reference process for disposal of radioactive waste from the Savannah River Plant is vitrification of the waste in borosilicate glass in a continuous glass melter. Design, startup, and operation of a plant-scale developmental melter system are discussed

  11. Partial replacement of the feldspar waste of flat glass ceramics for masses in white

    International Nuclear Information System (INIS)

    Porto, V.S.; Araujo, A.M.B.; Morais, C.R.S.; Cavalcanti, M.S.L.

    2012-01-01

    In all the industrial production process requires the consumption of raw materials exaggerated the traditional scarcity of incurring the same. To reverse this situation, one of the possible actions is the search for alternative technologies that aim to replace these materials by waste that exhibit similar characteristics. This study aims to verify the possibility of partially replacing feldspar by waste flat glass ceramic mass for white, since this type of waste, when subjected to high temperatures can act as a flux. For this research, initially the raw materials were characterized using the techniques of energy dispersive spectroscopy X-ray (EDX) and X-ray diffraction. Then, test pieces were prepared to be burned at temperatures between 1000 and 1250 ° C, which were submitted to tests of porosity to absorb water. The results are within the standards required by the standards established for ceramic products, which confirms the feasibility of such waste to act as a flux in ceramic white masses. (author)

  12. Characterization of high level nuclear waste glass samples following extended melter idling

    Energy Technology Data Exchange (ETDEWEB)

    Fox, Kevin M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Peeler, David K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kruger, Albert A. [USDOE Office of River Protection, Richland, WA (United States)

    2015-06-16

    The Savannah River Site Defense Waste Processing Facility (DWPF) melter was recently idled with glass remaining in the melt pool and riser for approximately three months. This situation presented a unique opportunity to collect and analyze glass samples since outages of this duration are uncommon. The objective of this study was to obtain insight into the potential for crystal formation in the glass resulting from an extended idling period. The results will be used to support development of a crystal-tolerant approach for operation of the high-level waste melter at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Two glass pour stream samples were collected from DWPF when the melter was restarted after idling for three months. The samples did not contain crystallization that was detectible by X-ray diffraction. Electron microscopy identified occasional spinel and noble metal crystals of no practical significance. Occasional platinum particles were observed by microscopy as an artifact of the sample collection method. Reduction/oxidation measurements showed that the pour stream glasses were fully oxidized, which was expected after the extended idling period. Chemical analysis of the pour stream glasses revealed slight differences in the concentrations of some oxides relative to analyses of the melter feed composition prior to the idling period. While these differences may be within the analytical error of the laboratories, the trends indicate that there may have been some amount of volatility associated with some of the glass components, and that there may have been interaction of the glass with the refractory components of the melter. These changes in composition, although small, can be attributed to the idling of the melter for an extended period. The changes in glass composition resulted in a 70-100 °C increase in the predicted spinel liquidus temperature (TL) for the pour stream glass samples relative to the analysis of the melter feed prior to

  13. Volume reduction and solidification of radioactive waste incineration ash with waste glass

    International Nuclear Information System (INIS)

    Koyama, Hidemi; Kobayashi, Masayuki

    2007-01-01

    The low-level radioactive waste generated from research institutions and hospitals etc. is packed into a container and is kept. The volume reduced state or the unprocessed state by incineration or compression processing are used because neither landfill sites nor disposal methods have been fixed. Especially, because the bulk density is low, and it is easy to disperse, the low-level radioactive waste incineration ash incinerated for the volume reduction is a big issue in security, safety, stability in the inventory location. A safe and appropriate disposal processing method is desired. When the low temperature sintering method in the use of the glass bottle cullet was examined, volume reduction and stabilization of low-level radioactive waste incineration ash were verified. The proposed method is useful for the easy treatment of the low-level radioactive waste incineration ash. (author)

  14. Characteristics of borosilicate waste glass form for high-level radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Soo; Chun, Kwan Sik; Choi, Jong Won; Kang, Chul Hyung

    2001-03-01

    Basic data, required for the design and the performance assessment of a repository of HLW, suchas the chemical composition and the characteristics of the borosilicate waste glass have been identified according to the burn-ups of spent PWR fuels. The diemnsion of waste canister is 430mm in diameter and 1135mm in length, and the canister should hold less than 2kwatts of heat from their decay of radionuclides contained in the HLW. Based on the reprocessing of 5 years-cooled spent fuel, one canister could hold about 11.5wt.% and 10.8wt.% of oxidized HLW corresponding to their burn-ups of 45,000MWD/MTU and 55,000MWD/MTU, respectively. These waste forms have been recommanded as the reference waste forms of HLW. The characteristics of these wastes as a function of decay time been evaluated. However, after a specific waste form and a specific site for the disposal would be selected, the characteristics of the waste should be reevaluated under the consideration of solidification period, loaded waste, storage condition and duration, site circumstances for the repository system and its performance assessment.

  15. 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

  16. 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

  17. Cold-cap reactions in vitrification of nuclear waste glass: experiments and modeling

    International Nuclear Information System (INIS)

    Chun, Jaehun; Pierce, David A.; Pokorny, Richard; Hrma, Pavel R.

    2013-01-01

    Cold-cap reactions are multiple overlapping reactions that occur in the waste-glass melter during the vitrification process when the melter feed is being converted to molten glass. In this study, we used differential scanning calorimetry (DSC) to investigate cold-cap reactions in a high-alumina high-level waste melter feed. To separate the reaction heat from both sensible heat and experimental instability, we employed the run/rerun method, which enabled us to define the degree of conversion based on the reaction heat and to estimate the heat capacity of the reacting feed. Assuming that the reactions are nearly independent and can be approximated by the nth order kinetics, we obtained the kinetic parameters using the Kissinger method combined with least squares analysis. The resulting mathematical simulation of the cold-cap reactions provides a key element for the development of an advanced cold-cap model

  18. Silicate glasses

    International Nuclear Information System (INIS)

    Lutze, W.

    1988-01-01

    Vitrification of liquid high-level radioactive wastes has received the greatest attention, world-wide, compared to any other HLW solidification process. The waste form is a borosilicate-based glass. The production of phosphate-based glass has been abandoned in the western world. Only in the Soviet Union are phosphate-based glasses still being developed. Vitrification techniques, equipment and processes and their remote operation have been developed and studied for almost thirty years and have reached a high degree of technical maturity. Industrial demonstration of the vitrification process has been in progress since 1978. 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 glasses

  19. Reuse of the red brick waste and dust waste of blasting chamber (glass micro spheres) in the red ceramic industry

    International Nuclear Information System (INIS)

    Rodrigues, R.A.; Felippe, C.E.C.; Guimaraes, C.S.; Almeida, V.C.

    2010-01-01

    The search for alternative environmentally less aggressive disposal of solid waste has been adopted to reverse the negative scenario established by the improper disposal of these materials. The aim of this study was to evaluate the reuse of waste: leftover red brick from the civil construction and glass micro spheres, obtained from the blasting chamber, aiming to develop a ceramic product. Mixtures containing various amounts of waste were prepared. The ceramic pieces were burned at 1000 and 1200 deg C being tested for water absorption and tensile strength and characterized by X-ray diffraction. The analysis of volatile organic compounds released during the burning process was performed. The results indicate that the ceramic material produced has a high resistance although the analysis of gases from the burning point to a negative environmental impact. (author)

  20. Influence of phosphate glass recrystallization on the stability of a waste matrix to leaching

    Science.gov (United States)

    Yudintsev, S. V.; Pervukhina, A. M.; Mokhov, A. V.; Malkovsky, V. I.; Stefanovsky, S. V.

    2017-04-01

    In Russia, highly radioactive liquid wastes from recycling of spent fuel of nuclear reactors are solidified into Na-Al-P glass for underground storage. The properties of the matrix including the radionuclide fixation will change with time due to crystallization. This is supported by the results of study of the interaction between glassy matrices, products of their crystallization, and water. The concentration of Cs in a solution at the contact of a recrystallized sample increased by three orders of magnitude in comparison with an experiment with glass. This difference is nearly one order of magnitude for Sr, Ce, and Nd (simulators of actinides) and U due to their incorporation into phases with low solubility in water. Based on data on the compositional change of solutions after passing through filters of various diameters, it is concluded that Cs occurs in the dissolved state in runs with a glass and recrystallized matrix. At the same time, Sr, lanthanides, and U occur in the dissolved state and in the composition of colloids in runs with glass, and mostly in colloid particles after contact with the recrystallized sample. These results should be regarded for substantiation of safety for geological waste storage.

  1. 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)

  2. 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

  3. 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

  4. Leaching characteristics of actinides from simulated reactor waste glass

    International Nuclear Information System (INIS)

    Weed, H.C.; Coles, D.G.; Bradley, D.J.; Mensing, R.W.; Schweiger, J.S.

    1979-01-01

    Two methods for measuring the leach rates of simulated high level waste glass are compared. One is a modification of the standard IAEA method and the other is a one-pass method in which fresh leachant solution is pumped over the sample at a controlled flow rate and temperature. For times up to 3 days, there is close agreement between results from the two methods at 25.0 0 C. Leach rates from the one-pass method show a correlation with flow rate only on day 1 at 25.0 0 C, whereas they show a correlation with flow rate for all three days at 75.0 0 C. 237 Np rates at 75.0 0 C are greater than those at 25.0 0 C, but 239 Pu rates at 75.0 0 C are less than or equal to those at 25.0 0 C

  5. 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.

  6. HLW Glass Studies: Development of Crystal-Tolerant HLW Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Matyas, Josef; Huckleberry, Adam R.; Rodriguez, Carmen P.; Lang, Jesse B.; Owen, Antionette T.; Kruger, Albert A.

    2012-04-02

    In our study, a series of lab-scale crucible tests were performed on designed glasses of different compositions to further investigate and simulate the effect of Cr, Ni, Fe, Al, Li, and RuO2 on the accumulation rate of spinel crystals in the glass discharge riser of the HLW melter. The experimental data were used to expand the compositional region covered by an empirical model developed previously (Matyáš et al. 2010b), improving its predictive performance. We also investigated the mechanism for agglomeration of particles and impact of agglomerates on accumulation rate. In addition, the TL was measured as a function of temperature and composition.

  7. 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

  8. 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.)

  9. 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.

  10. Immobilization of high-level wastes into sintered glass: 2

    International Nuclear Information System (INIS)

    Bevilacqua, A.M.; Russo, D.O.; Messi de Bernasconi, N.; Audero, M.A.

    1987-01-01

    High level radioactive wastes are immobilized into borosilicate glasses. Experiences with the variety VG 98/12 (SiO 2 , TiO 2 , Al 2 O 3 , B 2 O 3 , MgO, CaO, Na 2 O) are described. The pressing was performed in a matrix of 12.7 mm diameter, the walls of which were lubricated with sterotex dissolved in Cl 4 C. The sintering was made in an horizontal electric furnace in air atmosphere at temperatures between 500 and 600 deg C. It was observed that the maximum density occurs at 605 deg C. Comparing both the hot and the cold pressing process, it is concluded that: 1) In spite of requiring more complex equipment the hot pressing process has the advantage that lower pressures are applied, with the consequent obtainment of waste blocks with larger diameters, and 2) it is advisable that pressing process should be performed in the definitive can. (M.E.L.) [es

  11. Chemical composition analysis and product consistency tests to support enhanced Hanford waste glass models: Results for the January, March, and April 2015 LAW glasses

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. 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); Riley, W. T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Best, D. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-03

    In this report, the Savannah River National Laboratory provides chemical analyses and Product Consistency Test (PCT) results for several simulated low activity waste (LAW) glasses (designated as the January, March, and April 2015 LAW glasses) fabricated by the Pacific Northwest National Laboratory. The results of these analyses will be used as part of efforts to revise or extend the validation regions of the current Hanford Waste Treatment and Immobilization Plant glass property models to cover a broader span of waste compositions.

  12. Chemical composition analysis and product consistency tests to support Enhanced Hanford Waste Glass Models. Results for the Augusta and October 2014 LAW Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. 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); Best, D. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-07

    In this report, the Savannah River National Laboratory provides chemical analyses and Product Consistency Test (PCT) results for several simulated low activity waste (LAW) glasses (designated as the August and October 2014 LAW glasses) fabricated by the Pacific Northwest National Laboratory. The results of these analyses will be used as part of efforts to revise or extend the validation regions of the current Hanford Waste Treatment and Immobilization Plant glass property models to cover a broader span of waste compositions.

  13. Preliminary flowsheet for the conversion of Hanford high-level waste to glass

    International Nuclear Information System (INIS)

    Beary, M.M.; Chick, L.A.; Ely, P.C.; Gott, S.A.

    1977-06-01

    The flowsheets describe a process for converting waste removed from the Hanford underground waste tanks to more immobile form. The process involves a chemical separation of the radionuclides from industrial chemicals, and then making glass from the resulting small volume of highly radioactive waste. Removal of Sr, actinides, cesium, and technetium is discussed

  14. Platinoids and molybdenum in nuclear waste containment glasses: a structural study; Les platinoides et le molybdene dans des verres d'interet nucleaires: etude structurale

    Energy Technology Data Exchange (ETDEWEB)

    Le Grand, M. [CEA/VALRHO - site de Marcoule, Dept. de Recherche en Retraitement et en Vitrification (DRRV), 30 - Marcoule (France)]|[Paris-7 Univ., 75 (France)

    2000-07-01

    This work deals with the structure of borosilicate nuclear glasses and with some relationships between structure and macroscopic properties. Two types of elements which may disturb the industrial process - platinoids (Ru and Pd) and molybdenum - are central to this work. Platinoids induce weak modifications on the structure of the glass, causing a depolymerization of the glassy network, an increase of the {sup [3]}B/{sup [4]}B ratio and a modification of the medium range order around Si between 3.3 and 4.5 angstrom. The modifications of viscosity and density induced by platinoids in the glass are not due to the structural effect of the platinoids. The increase of viscosity is attributed to needle shaped RuO{sub 2}. It can be moderated by imposing reducing conditions during the elaboration of the glass. The slight difference between experimental and calculated densities is due to the increase of the volume percentage of bubbles in the glass with increasing platinoid content. Mo is either present in the glass as molybdic groupings, or mobilized in chemically complex molybdic crystalline phases. The chemical composition and mineralogy of these phases has been obtained using electronic microprobe data and XRD with Rietveld analysis. The distribution of the different elements between the crystalline phases and the glass is strongly influenced by the structural role of the various cations in the glass. The Mo present in the glass appears as MoO{sub 4} tetrahedra, independent of the borosilicate network. The formation of the crystalline phases can be explained by the existence of a precursor in which the MoO{sub 4} tetrahedra are concentrated in rich alkali and earth-alkali bearing areas of the glass. (author)

  15. Platinoids and molybdenum in nuclear waste containment glasses: a structural study; Les platinoides et le molybdene dans des verres d'interet nucleaires: etude structurale

    Energy Technology Data Exchange (ETDEWEB)

    Le Grand, M [CEA/VALRHO - site de Marcoule, Dept. de Recherche en Retraitement et en Vitrification (DRRV), 30 - Marcoule (France); [Paris-7 Univ., 75 (France)

    2000-07-01

    This work deals with the structure of borosilicate nuclear glasses and with some relationships between structure and macroscopic properties. Two types of elements which may disturb the industrial process - platinoids (Ru and Pd) and molybdenum - are central to this work. Platinoids induce weak modifications on the structure of the glass, causing a depolymerization of the glassy network, an increase of the {sup [3]}B/{sup [4]}B ratio and a modification of the medium range order around Si between 3.3 and 4.5 angstrom. The modifications of viscosity and density induced by platinoids in the glass are not due to the structural effect of the platinoids. The increase of viscosity is attributed to needle shaped RuO{sub 2}. It can be moderated by imposing reducing conditions during the elaboration of the glass. The slight difference between experimental and calculated densities is due to the increase of the volume percentage of bubbles in the glass with increasing platinoid content. Mo is either present in the glass as molybdic groupings, or mobilized in chemically complex molybdic crystalline phases. The chemical composition and mineralogy of these phases has been obtained using electronic microprobe data and XRD with Rietveld analysis. The distribution of the different elements between the crystalline phases and the glass is strongly influenced by the structural role of the various cations in the glass. The Mo present in the glass appears as MoO{sub 4} tetrahedra, independent of the borosilicate network. The formation of the crystalline phases can be explained by the existence of a precursor in which the MoO{sub 4} tetrahedra are concentrated in rich alkali and earth-alkali bearing areas of the glass. (author)

  16. 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

  17. A relationship between leach rate of nuclear waste glass and residual amount of sodium on the glass surface

    International Nuclear Information System (INIS)

    Kamizono, Hiroshi; Banba, Tsunetaka

    1984-12-01

    Leach tests of simulated high-level waste glass were carried out in order to examine the quantitative relationship between the amount of elements on the sample surface and that in the leachate. An experimental equation was obtained expressing the relationship between the amount of Na on the sample surface and that in the leachate. This shows that it is possible in some cases to estimate the amount of Na in the leachate by measuring the amount of Na on the sample surface. One example of such an estimation was observed with the simulated high-level waste glass leached at 100 0 C in the presence of a backfill material. (author)

  18. Radioactive waste repository study

    International Nuclear Information System (INIS)

    1978-11-01

    This is the second part of a report of a preliminary study for AECL. It considers the requirements for an underground waste repository for the disposal of wastes produced by the Canadian Nuclear Fuel Program. The following topics are discussed with reference to the repository: 1) geotechnical assessment, 2) hydrogeology and waste containment, 3) thermal loading and 4) rock mechanics. (author)

  19. Infrared and Raman investigation of rare-earth phosphate glasses for potential use as radioactive waste forms

    International Nuclear Information System (INIS)

    Morgan, S.H.

    1989-01-01

    This project was designed to investigate the properties of the rare-earth phosphate glass systems CeO 2 -P 2 O 5 and Pr 2 O 3 -P 2 O 5 for potential use as radioactive waste glasses. The glass-forming region and optimum processing parameters of these glass systems were investigated. The structure of the host glasses and glassed loaded with simulated waste elements was investigated using Raman and infrared spectroscopy. Because of the radical differences in the spectra of the molybdenum-loaded glasses, the structure of the MoO 3 -P 2 O 5 glass system was also investigated. 29 refs., 8 figs., 2 tabs

  20. Spectroscopic study of biologically active glasses

    Science.gov (United States)

    Szumera, M.; Wacławska, I.; Mozgawa, W.; Sitarz, M.

    2005-06-01

    It is known that the chemical activity phenomenon is characteristic for some inorganic glasses and they are able to participate in biological processes of living organisms (plants, animals and human bodies). An example here is the selective removal of silicate-phosphate glass components under the influence of biological solutions, which has been applied in designing glasses acting as ecological fertilizers of controlled release rate of the nutrients for plants. The structure of model silicate-phosphate glasses containing the different amounts of the glass network formers, i.e. Ca 2+ and Mg 2+, as a binding components were studied. These elements besides other are indispensable of the normal growth of plants. In order to establish the function and position occupied by the particular components in the glass structure, the glasses were examined by FTIR spectroscopy (with spectra decomposition) and XRD methods. It has been found that the increasing amount of MgO in the structure of silicate-phosphate glasses causes the formation of domains the structure of which changes systematically from a structure of the cristobalite type to a structure corresponding to forsterite type. Whilst the increasing content of CaO in the structure of silicate-phosphate glasses causes the formation of domains the structure of which changes from a structure typical for cristobalite through one similar to the structure of calcium orthophosphate, to a structure corresponding to calcium silicates. The changing character of domains structure is the reason of different chemical activity of glasses.

  1. Copper solubility in DWPF, Batch 1 waste glass: Update report

    International Nuclear Information System (INIS)

    Schumacker, R.F.

    1992-01-01

    The ''Late Washing'' Step in the processing of precipitate will require the use of additional copper formate in the Precipitate Reactor to catalyze the hydrolysis reaction. The increased copper concentration in the melter feed increases the potential for metal precipitation during the vitrification of the melter feed. This report describes recent results with a conservative glass selected from the DWPF acceptable region in the Batch 1 Variability Study

  2. 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

  3. 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

  4. Leaching of the simulated borosilicate waste glasses and spent nuclear fuel under a repository condition

    International Nuclear Information System (INIS)

    Kim, Seung Soo; Chun, Kwan Sik; Kang, Chul Hyung; Suh, Hang Suk

    2002-12-01

    Leaching behaviors of simulated waste glass and spent fuel, contacted on bentonite blocks, in synthetic granitic groundwater were investigated in this study. The leach rate of boron from borosilicate waste glass between the compacted bentonite blocks reached about 0.03 gm-2day-1 at 1500 days, like as that of molybdenum. However, the concentration of uranium in leachate pass through bentonite blocks was less than their detection limits of 2 μg/L and whose yellow amorphous compound was found on the surface of glass contacted with the bentonite blocks. The leaching mechanism of waste glasses differed with their composition. The release rate of cesium from PWR spent fuel in the simulated granitic water without bentonite was leas than $1.0x10 -5 fraction/day after 300 days. The retardation factor of cesium by a 10 -mm thickness of bentonite block was more than 100 for 4-years leaching time. The cumulative release fraction of uranium for 954 days was 0.016% (1.7x10 -7 fraction/day) in granitic water without bentonite. The gap inventory of cesium for spent fuel G23-J11 was 0.15∼0.2%. However, the release of cesium from C15-I08 was 0.9% until 60 days and has being continued after that. Gap inventories of strontium and iodine in G23-J11 were 0.033% and below 0.2%, respectively. The sum of fraction of cesium in gap and grain boundary of G23-J11 was suggested below 3% and less

  5. 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.

  6. 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

  7. Isothermal crystallization kinetics in simulated high-level nuclear waste glass

    International Nuclear Information System (INIS)

    Vienna, J.D.; Hrma, P.; Smith, D.E.

    1997-01-01

    Crystallization kinetics of a simulated high-level waste (HLW) glass were measured and modelled. Kinetics of acmite growth in the standard HW39-4 glass were measured using the isothermal method. A time-temperature-transformation (TTT) diagram was generated from these data. Classical glass-crystal transformation kinetic models were empirically applied to the crystallization data. These models adequately describe the kinetics of crystallization in complex HLW glasses (i.e., RSquared = 0.908). An approach to measurement, fitting, and use of TTT diagrams for prediction of crystallinity in a HLW glass canister is proposed

  8. Solid waste study

    International Nuclear Information System (INIS)

    Ortiz, Paul G.

    1995-01-01

    The purpose of this document is to study the solid waste issues brought about by a Type C Investigation; ''Disposal of Inappropriate Material in the Los Alamos County Landfill'' (May 28, 1993). The study was completed in August 1995 by Coleman Research Corporation, under subcontract number 405810005-Y for Los Alamos National Laboratory (LANL). The study confirmed the issues identified in the Type C investigation, and also ascertained further issues or problems. During the course of this study two incidents involving hazardous waste resulted in the inappropriate disposal of the waste. An accidental spill, on June 8, 1995, at one of Laboratory buildings was not handled correctly, and ended up in the LAC Landfill. Hazardous waste was disposed of in a solid waste container and sent to the Los Alamos County Landfill. An attempt to locate the hazardous waste at the LAC Landfill was not successful. The second incident involving hazardous waste was discovered by the FSS-8, during a random dumpster surveillance. An interim dumpster program managed by FSS-8 discovered hazardous waste and copper chips in the solid waste, on August 9, 1995. The hazardous waste and copper chips would have been transported to the LAC Landfill if the audit team had not brought the problem to the awareness of the facility waste management personnel

  9. SOLID WASTE STUDY

    Energy Technology Data Exchange (ETDEWEB)

    PAUL G. ORTIZ - COLEMAN RESEARCH CORP/COMPA INDUSTRIES

    1995-08-01

    The purpose of this document is to study the solid waste issues brought about by a Type C Investigation; ``Disposal of Inappropriate Material in the Los Alamos County Landfill'' (May 28, 1993). The study was completed in August 1995 by Coleman Research Corporation, under subcontract number 405810005-Y for Los Alamos National Laboratory (LANL). The study confirmed the issues identified in the Type C investigation, and also ascertained further issues or problems. During the course of this study two incidents involving hazardous waste resulted in the inappropriate disposal of the waste. An accidental spill, on June 8, 1995, at one of Laboratory buildings was not handled correctly, and ended up in the LAC Landfill. Hazardous waste was disposed of in a solid waste container and sent to the Los Alamos County Landfill. An attempt to locate the hazardous waste at the LAC Landfill was not successful. The second incident involving hazardous waste was discovered by the FSS-8, during a random dumpster surveillance. An interim dumpster program managed by FSS-8 discovered hazardous waste and copper chips in the solid waste, on August 9, 1995. The hazardous waste and copper chips would have been transported to the LAC Landfill if the audit team had not brought the problem to the awareness of the facility waste management personnel.

  10. Studies of waste-canister compatibility

    International Nuclear Information System (INIS)

    McCoy, H.E.

    1983-01-01

    Compatibility studies were conducted between 7 waste forms and 15 potential canister structural materials. The waste forms were Al-Si and Pb-Sn matrix alloys, FUETAP, glass, Synroc D, and waste particles coated with carbon or carbon plus silicon carbide. The canister materials included carbon steel (bare and with chromium or nickel coatings), copper, Monel, Cu-35% Ni, titanium (grades 2 and 12), several Inconels, aluminum alloy 5052, and two stainless steels. Tests of either 6888 or 8821 h were conducted at 100 and 300 0 C, which bracket the low and high limits expected during storage. Glass and FUETAP evolved sulfur, which reacted preferentially with copper, nickel, and alloys of these metals. The Pb-Sn matrix alloy stuck to all samples and the carbon-coated particles to most samples at 300 0 C, but the extent of chemical reaction was not determined. Testing for 0.5 h at 800 0 C was included because it is representative of a transportation accident and is required of casks containing nuclear materials. During these tests (1) glass and FUETAP evolved sulfur, (2) FUETAP evolved large amounts of gas, (3) Synroc stuck to titanium alloys, (4) glass was molten, and (5) both matrix alloys were molten with considerable chemical interactions with many of the canister samples. If this test condition were imposed on waste canisters, it would be design limiting in many waste storage concepts

  11. Evaluation of the potential of waste fondant glass in formulations of ceramic pasta

    International Nuclear Information System (INIS)

    Soares Filho, J.E.; Santos, L.L. dos; Nascimento, R.M. do; Feitosa, A.O.; Dutra, R.P.S.

    2014-01-01

    An increasing amount of waste generated and deposited on the environment, many unspecified decomposition with time, as is the case of the glass. Thinking about it, the purpose of this study is to evaluate the power of the flux residue on glass formulations porcelains, as a flux to feldspar replacement. This study was performed in comparison with a standard formulation. The raw materials were characterized in the diffraction X-ray fluorescence and X-ray thermal differential analysis, and determination of the technological properties of water absorption, linear contraction, ignition loss, apparent porosity and apparent specific gravity in the formulation standard and replacement of feldspar in different percentages of waste and processing conditions. Specimens of the formulations were subjected to assay of three points. Results indicate that the residue glass has the potential of being used as a flux material in the composition of the ceramic body reduces the apparent porosity and according to the technology of water absorption property. The ceramic mass standard was classified as semi-stoneware, the BIIa group, and after the addition of the residue in any of the three percentages evaluated was classified as sandstone, belonging to the group BIb.(author)

  12. 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

  13. P2O5-doping in waste glasses: evolution of viscosity and crystallization processes

    Science.gov (United States)

    Tarrago, Mariona; Espuñes, Alex; Garcia-Valles, Maite; Martinez, Salvador

    2015-04-01

    Current concern for environmental preservation is the main motive for the study of new, more sustainable materials. Increasing amounts of sewage sludge are produced in wastewater treatment plants over the world every day. This fact represents a major problem for the municipalities and industries due to the volume of waste and also to the contaminant elements it may bear, which require expensive conditions for disposal in landfills. Vitrification is an established technique in the inertization of different types of toxic wastes (such as nuclear wastes and contaminated soils) that has been used successfully for sewage sludge. Glasses of basaltic composition (43.48SiO2-14.00Al2O3-12.86Fe2O3-10.00CaO-9.94MgO-3.27Na2O-1.96K2O-0.17MnO-0.55P2O5-2.48TiO2) are used as a laboratory analogous of wastes such as sewage sludge and galvanic sludge to study the properties of the inertization matrix. This basaltic matrix is doped by adding 1%, 2%, 3%, 4% and 20% of P5O5 in order to cover the compositional range of phosphate in sewage sludge encountered in the literature. In this study, the focus has been placed in the effect of the concentration of phosphate (P2O5) in glass stability, thermal properties and evolution of viscosity with temperature. The dependence of viscosity on temperature and the thermal behaviour of these glasses are critical parameters in the design of their production process. Regarding the compositional limits of the mixture, it has been observed that melt reactivity is much increased when P2O5 content is over 4%, hindering the glass conformation process. Moreover, stanfieldite (calcium and magnesium phosphate) crystallized during glass making when phosphate concentration approached 20%, hence establishing the upper limit for glass stability. Viscosity is also dramatically increased in this range, hence requiring production amends. Differential thermal analysis has provided nucleation and crystallization temperatures of the glasses around 915°C and 1050

  14. AN ALTERNATIVE HOST MATRIX BASED ON IRON PHOSPHATE GLASSES FOR THE VITRIFICATION OF SPECIALIZED WASTE FORMS

    International Nuclear Information System (INIS)

    Day, Delbert D.

    2000-01-01

    As mentioned above, the overall goal of this research project was to collect the scientific information essential to develop iron phosphate glass based nuclear wasteforms. The specific objectives of the project were: (1) Investigate the structure of binary iron phosphate glasses and it's dependence on the composition and melting atmosphere: Understand atomic arrangements and nature of the bonding. Establish structure-property relationships. Determine the compositions and melting conditions which optimize the critical properties of the base glass. (2) Understand the structure of iron phosphate wasteforms and it's dependence on the composition and melting atmosphere: Investigate how the waste elements are bonded and coordinated within the glass structure. Establish structure-property relationships for the waste glasses. Determine the compositions and melting atmosphere for which the critical properties of the waste forms would be optimum. (3) Determine the role(s) played by the valence states of iron ions and it's dependence on the composition and melting atmosphere: Understand the different roles of iron(II) and iron(III) ions in determining the critical properties of the base glass and the waste forms. Investigate how the iron valence and its significance depend on the composition and melting atmosphere. (4) Investigate glass forming and crystallization processes of the iron phosphate glasses and their waste forms: Understand the dependence of the glass forming and crystallization characteristics on overall glass composition and valence states of iron ions. Identify the products of devitrification and investigate the critical properties of these crystalline compounds which may adversely affect the chemical and physical properties of the waste forms

  15. Electrical power supply and controls for a remotely operated glass melter for nuclear waste

    International Nuclear Information System (INIS)

    Haideri, A.Q.

    1985-01-01

    An electrical power supply, controls and instruments used for a joule heated glass melter for nuclear waste are discussed. Remotely replaceable interconnection wiring assemblies for power, controls and instruments are also described

  16. 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.

  17. Viscosity and electrical conductivity of glass melts as a function of waste composition

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Wiley, J.R.

    1979-01-01

    Radioactive waste at the Savannah River Plant contains high concentrations of nonradioactive compounds of iron and aluminum. Simulated waste compositions containing varying ratios of iron to aluminum were added to glass melts to determine the effect on the melt properties. Waste containing high-aluminum increased the melt viscosity, but waste containing high-iron reduced the melt viscosity. Aluminum and iron both reduced the melt conductivity

  18. 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

  19. The role of ceramics, cement and glass in the immobilization of radioactive wastes

    International Nuclear Information System (INIS)

    Glasser, F.P.

    1985-01-01

    A brief account is given of the constitution and origin of nuclear waste. The immobilization of wastes is discussed: borosilicate glasses are considered as possible matrices; ceramic forms are dealt with in more detail. The principles of the use of ceramics are explained, with examples of different ceramic structures; cements are mentioned as being suitable for wet, medium- to low-active wastes. The effects of radiation on cement, ceramic and glass waste forms are indicated. The account concludes with 'summary and future progress'. (U.K.)

  20. 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.

  1. Thermal stability of the French nuclear waste glass - long term behavior modeling

    International Nuclear Information System (INIS)

    Orlhac, X.

    2000-01-01

    The thermal stability of the French nuclear waste glass was investigated experimentally and by modeling to predict its long-term evolution at low temperature. The crystallization mechanisms were analyzed by studying devitrification in the supercooled liquid. Three main crystalline phases were characterized (CaMoO 4 , CeCO 2 , ZnCr 2 O 4 ). Their crystallisation was TO 4.24 wt%, due to the low concentration of the constituent elements. The nucleation and growth curves showed that platinoid elements catalysed nucleation but did not affect growth, which was governed by volume diffusion. The criteria of classic nucleation theory were applied to determine the thermodynamic and diffusional activation energies. Viscosity measurements illustrate the analogy between the activation energy of viscous flow and diffusion, indicating control of crystallization by viscous flow phenomena. The combined action of nucleation and growth was assessed by TTT plots, revealing a crystallization equilibrium line that enables the crystallized fractions to be predicted over the long term. The authors show that hetero-genetics catalyze the transformation without modifying the maximum crystallized fraction. A kinetic model was developed to describe devitrification in the glass based on the nucleation and growth curves alone. The authors show that the low-temperature growth exhibits scale behavior (between time and temperature) similar to thermo-rheological simplicity. The analogy between the resulting activation energy and that of the viscosity was used to model growth on the basis of viscosity. After validation with a simplified (BaO 2 SiO 2 ) glass, the model was applied to the containment glass. The result indicated that the glass remained completely vitreous after a cooling scenario with the one measured at the glass core. Under isothermal conditions, several million years would be required to reach the maximum theoretical crystallization fraction. (author)

  2. Structural characterization of hog iron oxide content glasses obtained from zinc hydrometallurgy wastes

    International Nuclear Information System (INIS)

    Romero, M.; Rincon, J.M.; Musik, S.; Kozhujharov, W.

    1997-01-01

    It has been carried out the structural characterization of high oxide content glasses obtained by melting of a goethite industrial waste from the zinc hydrometallurgy with other raw materials as dolomite and glass cullet. The structural characterization has been carried out by X-ray Diffraction (XRD), X-Ray Diffraction by Amorphous Dispersion (RDF) and Mossbauer spectroscopy. It has been determined the interatomic distance, the oxidation state and the coordination of iron atoms in these glasses. (Author) 16 refs

  3. Effect of clayey groundwater on the dissolution rate of SON68 simulated nuclear waste glass at 70 °C

    Science.gov (United States)

    De Echave, T.; Tribet, M.; Jollivet, P.; Marques, C.; Gin, S.; Jégou, C.

    2018-05-01

    To predict the long-term behavior of high-level radioactive waste glass, it is necessary to study aqueous dissolution of the glass matrix under geological repository conditions. The present article focuses on SON68 (an inactive surrogate of the R7T7 glass) glass alteration in synthetic clayey groundwater at 70 °C. Experiments in deionized water as reference were also performed in the same conditions. Results are in agreement with those of previous studies showing that magnesium present in the solution is responsible for higher glass alteration. This effect is transient and pH-dependent: Once all the magnesium is consumed, the glass alteration rate diminishes. Precipitation of magnesium silicate of the smectite group seems to be the main factor for the increased glass alteration. A pH threshold of 7.5-7.8 was found, above which precipitation of these magnesium silicates at 70 °C is possible. TEM observations reveal that magnesium silicates grow at the expense of the passivating gel, which partly dissolves, forming large pores which increase mass transfer between the reacting glass surface and the bulk solution.

  4. NMR studies of the structure of glasses

    International Nuclear Information System (INIS)

    Bray, P.J.; Gravina, S.J.; Stallworth, P.E.; Szu, S.P.; Jianhui Zhong

    1988-01-01

    Earlier continuous wave (CW) NMR studies of chemical bonding and structure in glasses are summarized. Examples are given of this use of the quadrupolar interaction and chemical shift to obtain structural information. New NMR data and analyses are presented for alkali borate and gallate glasses. Extensions to other elements (e.g. molybdenum, lanthanum) are suggested. 44 refs. (author)

  5. Alkali activated slag cements using waste glass as alternative activators. Rheological behaviour

    Directory of Open Access Journals (Sweden)

    Manuel Torres-Carrasco

    2015-03-01

    The findings show that AAS paste behaviour of rheology when the activator was a commercial waterglass solution or NaOH/Na2CO3 with waste glass was similar, fit the Herschel-Bulkley model. The formation of primary C-S-H gel in both cases were confirmed. However, the rheological behaviour in standard cements fit the Bingham model. The use of the waste glass may be feasible from a rheological point of view in pastes can be used.

  6. Electrical service and controls for Joule heating of a defense waste experimental glass melter

    International Nuclear Information System (INIS)

    Erickson, C.J.; Haideri, A.Q.

    1983-01-01

    Vitrification of radioactive liquid waste in a glass matrix is a leading candidate for long-term storage of high-level waste. This paper describes the electrical service and control system for an experimental electrically heated, nonradioactive glass melter installed at Savannah River Laboratory. Data accumulated, and design/operating experience acquired in operating this melter, are being used to design a modified melter to be installed in a processing area for use with radioactive materials

  7. Waste glass as partial mineral precursor in alkali-activated slag/fly ash system

    NARCIS (Netherlands)

    Zhang, S.; Keulen, A.; Arbi, K.; Ye, G.

    2017-01-01

    The feasibility of a waste glass powder residue (GP) from glass recycling as partial mineral precursor to produce alkali-activated materials is investigated. GP served as powder coal fly ash (PCFA) replacement within a reference system composed of 50% PCFA and 50% ground granulated blast furnace

  8. 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

  9. Glass Formulation For The Hanford Tank Waste Treatment And Immobilization Plant (WTP)

    International Nuclear Information S