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Sample records for waste glass compositional

  1. 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

  2. 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

  3. 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

  4. 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)

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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)

  11. 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.

  12. 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

  13. 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

  14. 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

  15. Development of glass compositions with 9% waste content for the vitrification of high-level waste from LWR nuclear reactors

    International Nuclear Information System (INIS)

    Lakatos, T.

    1979-10-01

    Reduction of the contents of waste in glass from 20-25% to 9% causes a decrease of the leaching resistance of the glass. The addition of Zn0 reduces the leaching values by a factor of approximately 10. The crystallized glass ceramics have a lower coefficient of thermal expansion than glassy waste bodies. The separation of the phase which contains Mo occurs during heat treatment. The amount of separated Mo is lower for low alkali sac type (Si0 2 - A1 2 0 3 -Ca0 system) of glasses by a factor of approximately 50. All the glasses were prepared with simulated waste composition. (GBn.)

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

    Directory of Open Access Journals (Sweden)

    Rozenstrauha, I.

    2013-04-01

    Full Text Available 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 aluminiumborosilicate glass fibre with relatively high softening temperature (> 600 ºC. 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 ºC in different treatment conditions. Materials produced in temperature range 1090–1100 ºC with the most optimal properties - porosity 38-52%, water absorption 39–47% and bulk density 1.35–1.67 g/cm3 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/cm3 - 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.Durante la obtención de ciertas fibras de vidrio se generan dos subproductos o residuos principalmente: Lodo de arcilla montmorillonítica capaz de adsorber el 50 % de materia orgánica y un vidrio silicato alumínico con temperatura de reblandecimiento relativamente alta (> 600 ºC. Con el fin de elaborar nuevos

  17. 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

  18. 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

  19. Formulation of portland composite cement using waste glass as a supplementary cementitious material

    Science.gov (United States)

    Manullang, Ria Julyana; Samadhi, Tjokorde Walmiki; Purbasari, Aprilina

    2017-09-01

    Utilization of waste glass in cement is an attractive options because of its pozzolanic behaviour and the market of glass-composite cement is potentially available. The objective of this research is to evaluate the formulation of waste glass as supplementary cementitious material (SCM) by an extreme vertices mixture experiment, in which clinker, waste glass and gypsum proportions are chosen as experimental variables. The composite cements were synthesized by mixing all of powder materials in jar mill. The compressive strength of the composite cement mortars after being cured for 28 days ranges between 229 to 268 kg/cm2. Composite cement mortars exhibit lower compressive strength than ordinary Portland cement (OPC) mortars but is still capable of meeting the SNI 15-7064-2004 standards. The highest compressive strength is obtained by shifting the cement blend composition to the direction of increasing clinker and gypsum proportions as well as reducing glass proportion. The lower compressive strength of composite cement is caused by expansion due to ettringite and ASR gel. Based on the experimental result, the composite cement containing 80% clinker, 15% glass and 5% gypsum has the highest compressive strength. As such, the preliminary technical feasibility of reuse of waste glass as SCM has been confirmed.

  20. 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.

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

    International Nuclear Information System (INIS)

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

    1981-09-01

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

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

    International Nuclear Information System (INIS)

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

    1994-11-01

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

  3. 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

  4. 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.

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  6. 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

  7. 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

  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. 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.

  10. 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 %.

  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. 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.

  14. 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)

  15. 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

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

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1991-01-01

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

  17. Industrial waste as a source for fabrication of composite ceramics-glass with a controlled porosity

    Directory of Open Access Journals (Sweden)

    Adziski R.

    2008-01-01

    Full Text Available Metallurgical slag with granulation (-0.125+0.063mm and 20 wt% waste TV glass were used for obtaining a glass ceramic composite with a controlled porosity. This material obtained by sintering at 950oC/2h possessed thermal stability, integral porosity of 43.6% and E-modulus and bending strength of 12 GPa and 39 MPa, respectively. The composite was characterized with a permeability of 0.47 Da and generation of air bubbles with size of 1-4 mm in a water medium.

  18. Property-Composition-Temperature Modeling of Waste Glass Melt Data Subject to a Randomization Restriction

    International Nuclear Information System (INIS)

    Piepel, Gregory F.; Heredia-Langner, Alejandro; Cooley, Scott K.

    2008-01-01

    Properties such as viscosity and electrical conductivity of glass melts are functions of melt temperature as well as glass composition. When measuring such a property for several glasses, the property is typically measured at several temperatures for one glass, then at several temperatures for the next glass, and so on. This data-collection process involves a restriction on randomization, which is referred to as split-plot experiment. The split-plot data structure must be accounted for in developing property-composition-temperature models and the corresponding uncertainty equations for model predictions. Instead of ordinary least squares (OLS) regression methods, generalized least squares (GLS) regression methods using restricted maximum likelihood (REML) estimation must be used. This article describes the methodology for developing property-composition-temperature models and corresponding prediction uncertainty equations using the GLS/REML regression approach. Viscosity data collected on 197 simulated nuclear waste glasses are used to illustrate the GLS/REML methods for developing a viscosity-composition-temperature model and corresponding equations for model prediction uncertainties. The correct results using GLS/REML regression are compared to the incorrect results obtained using OLS regression

  19. 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)

  20. 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

  1. MULTILAYER POROUS COMPOSITE FROM WASTE GLASS FOR WATER FILTRATION

    Directory of Open Access Journals (Sweden)

    M. P. Aji

    2015-07-01

    Full Text Available Multilayer porous composite have been produced through the heating process at temperature T=700oC for 2.5 h. Single layered porous composite was made with a varied mass percentage of from PEG polymer  1% to 10%. Double-layered porous composite were made by the arrangement of porosity (4:3%, (4:2% and (3:2%, while the three-layers porous composite have an arrangement (4:3:2%. Performance of multilayer porous composite for water filtration with pollutants of methylene blue 100 ppm was estimated from the absorbance spectrum. Rejection of methylene blue pollutants from single layered porous composite increases when the fraction of PEG polymer tend to be smaller in the matrix. Meanwhile, the double layered porous composite has a degradation of methylene blue pollutants are better than one layer. Triple layered porous composite have good performance for the water filtration where all the pollutants of methylene blue be able to be filtered.   Komposit pori berlapis telah dihasilkan dengan proses pemanasan pada temperatur T=700oC selama 2.5 jam. Komposit pori satu lapis dibuat dengan variasi persen massa polimer PEG 1% hingga 10%. Komposit pori dua lapis dibuat dengan susunan porositas (4:3%, (4:2% dan (3:2%, sedangkan komposit pori tiga lapis memiliki susunan porositas (4:3:2%. Kinerja komposit pori berlapis untuk filter air dengan polutan methylene blue 100 ppm diestimasi dari spektrum absorbansi. Rejeksi polutan methylene blue dari komposit pori satu lapis meningkat saat fraksi polimer PEG cenderung lebih kecil dalam matrik komposit. Sedangkan, komposit pori dua lapis memiliki kemampuan untuk degradasi polutan methylene blue yang lebih baik dari satu lapis. Komposit pori tiga lapis memiliki kinerja yang baik untuk filter air dimana seluruh polutan methylene blue mampu disaring. 

  2. Sintered glass ceramic composites from vitrified municipal solid waste bottom ashes

    International Nuclear Information System (INIS)

    Aloisi, Mirko; Karamanov, Alexander; Taglieri, Giuliana; Ferrante, Fabiola; Pelino, Mario

    2006-01-01

    A glass ceramic composite was obtained by sinter-crystallisation of vitrified municipal solid waste bottom ashes with the addition of various percentages of alumina waste. The sintering was investigated by differential dilatometry and the crystallisation of the glass particles by differential thermal analysis. The crystalline phases produced by the thermal treatment were identified by X-ray diffraction analysis. The sintering process was found to be affected by the alumina addition and inhibited by the beginning of the crystal-phase precipitation. Scanning electron microscopy was performed on the fractured sintered samples to observe the effect of the sintering. Young's modulus and the mechanical strength of the sintered glass ceramic and composites were determined at different heating rates. The application of high heating rate and the addition of alumina powder improved the mechanical properties. Compared to the sintered glass ceramic without additives, the bending strength and the Young's modulus obtained at 20 deg. C/min, increased by about 20% and 30%, respectively

  3. 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.

  4. 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

  5. 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

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

    International Nuclear Information System (INIS)

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

    1992-04-01

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

  7. 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

  8. Glass compositions

    Energy Technology Data Exchange (ETDEWEB)

    France, P W

    1985-05-30

    A fluoride glass for use in the production of optical fibres has an enhanced D/H ratio, preferably such that OD:OH is at least 9:1. In the example, such a glass is prepared by treating with D/sub 2/O a melt comprising 51.53 mole per cent ZrF/sub 4/, 20.47 mole per cent BaF/sub 2/, 5.27 mole per cent LaF/sub 3/, 3.24 mole per cent AlF/sub 3/, and 19.49 mole per cent LiF.

  9. MASBAL: A computer program for predicting the composition of nuclear waste glass produced by a slurry-fed ceramic melter

    International Nuclear Information System (INIS)

    Reimus, P.W.

    1987-07-01

    This report is a user's manual for the MASBAL computer program. MASBAL's objectives are to predict the composition of nuclear waste glass produced by a slurry-fed ceramic melter based on a knowledge of process conditions; to generate simulated data that can be used to estimate the uncertainty in the predicted glass composition as a function of process uncertainties; and to generate simulated data that can be used to provide a measure of the inherent variability in the glass composition as a function of the inherent variability in the feed composition. These three capabilities are important to nuclear waste glass producers because there are constraints on the range of compositions that can be processed in a ceramic melter and on the range of compositions that will be acceptable for disposal in a geologic repository. MASBAL was developed specifically to simulate the operation of the West Valley Component Test system, a commercial-scale ceramic melter system that will process high-level nuclear wastes currently stored in underground tanks at the site of the Western New York Nuclear Services Center (near West Valley, New York). The program is flexible enough, however, to simulate any slurry-fed ceramic melter system. 4 refs., 16 figs., 5 tabs

  10. 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

  11. 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

  12. 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

  13. 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)

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

    International Nuclear Information System (INIS)

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

    1996-03-01

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

  15. Chemical composition analysis of simulated waste glass T10-G-16A

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-08-01

    In this report, SRNL provides chemical composition analyses of a simulated LAW glass designated T10-G-16A.The measured chemical composition data are reported and compared with the targeted values for each component. No issues were identified in reviewing the analytical data.

  16. Modeling relations between the composition and properties of French light water reactor waste containment glass

    International Nuclear Information System (INIS)

    Ghaleb, D.; Dussossoy, J.L.; Fillet, C.; Pacaud, F.; Jacquet-Francillon, N.

    1994-01-01

    Models have been developed to calculate the density, molten-state viscosity and initial corrosion rate according to the chemical composition of glass formulations used to vitrify high-level fission product solutions from reprocessed light water reactor fuel. Developed from other published work, these models have been adapted to allow for the effects of platinoid (Ru, Pd, Rh) inclusions on the molten glass rheology. (authors). 15 refs., 10 figs., 1 tab

  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. Preparation and Characterization of UPR/ LNR/ Glass Fiber Composite by using Unsaturated Polyester Resin (PET) from PET Wastes

    International Nuclear Information System (INIS)

    Siti Farhana Hisham; Ishak Ahmad; Rusli Daik

    2011-01-01

    UPR/ LNR/ glass fibre composite had been prepared by using unsaturated polyester resin (UPR) based from recycled PET product. PET waste was recycled by glycolysis process and the glycides product was then reacted with maleic anhydride to produce unsaturated polyester resin. The preparation of UPR/ LNR blends were conducted by varying the amount of LNR addition to the resin ranging from 0-7.5 % (wt). The composition of UPR/LNR blend with good mechanical properties had been selected as a matrix of the glass fiber reinforced composite. Glass fibre was also treated by (3-Amino propil)triethoxysilane as a coupling agent. From the result, the addition of 2.5 % LNR in UPR had showed the optimum mechanical and morphological properties where the elastomer particle's were well dispersed in the matrix with smaller size. The silane treatment on the glass fiber increased the tensile and impact strength values of the UPR/ LNR/ GF composite compared to untreated fiber reinforcement. (author)

  20. 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

  1. 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

  2. Using physical properties of molten glass to estimate glass composition

    International Nuclear Information System (INIS)

    Choi, Kwan Sik; Yang, Kyoung Hwa; Park, Jong Kil

    1997-01-01

    A vitrification process is under development in KEPRI for the treatment of low-and medium-level radioactive waste. Although the project is for developing and building Vitrification Pilot Plant in Korea, one of KEPRI's concerns is the quality control of the vitrified glass. This paper discusses a methodology for the estimation of glass composition by on-line measurement of molten glass properties, which could be applied to the plant for real-time quality control of the glass product. By remotely measuring viscosity and density of the molten glass, the glass characteristics such as composition can be estimated and eventually controlled. For this purpose, using the database of glass composition vs. physical properties in isothermal three-component system of SiO 2 -Na 2 O-B 2 O 3 , a software TERNARY has been developed which determines the glass composition by using two known physical properties (e.g. density and viscosity)

  3. 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

  4. 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.

  5. 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.

  6. 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

  7. 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.

  8. 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

  9. The Effect of Composition on Spinel Crystals Equilibrium in Low-Silica High-Level Waste Glasses

    International Nuclear Information System (INIS)

    Jiricka, Milos; Hrma, Pavel R.; Vienna, John D.

    2003-01-01

    The liquidus temperature (TL) and the equilibrium mass fraction of spinel were measured in the regions of low-silica (less than 42 mass% SiO2) high-level waste borosilicate glasses within the spinel primary phase field as functions of glass composition. The components that varied, one at a time, were Al2O3, B2O3, Cr2O3, Fe2O3, Li2O, MnO, Na2O, NiO, SiO2, and ZrO2. The effects of Al2O3, B2O3, Fe2O3, NiO, SiO2, and ZrO2 on the TL in this region and in glasses with 42 to 56 mass% SiO2 were similar. However, in the low-silica region, Cr2O3 increased the TL substantially less, and Li2O and Na2O decreased the TL significantly less than in the region with 42 to 56 mass% SiO2. The effect of MnO on the TL of the higher SiO2 glasses is not yet understood with sufficient accuracy. The temperature at which the equilibrium mass fraction of spinel was 1 mass% was 25C to 64C below the TL

  10. Effect of composition and radiation on the Hertzian indentation behavior of nuclear waste glasses

    International Nuclear Information System (INIS)

    Matzke, H.; Kahl, L.; Routbort, J.L.; Saidl, J.

    1983-01-01

    The Hertzian indentation technique has been used to determine the fracture toughness, K/sub Ic/ of two borosilicate glasses developed to contain high-level nuclear waste. For the product VG 98/12, adding selected groups of fission products leaves K/sub Ic/ unchanged, but addition of Pb lowers K/sub Ic/ by approx. 20%. Radiation with 77 MeV α-particles to a dose of approx. 10 15 α/cm 2 increases K/sub Ic/ by approx. 75%. For the product SM 58 LW 11, the fracture toughness was measured on pieces taken from different parts of a large cylinder to investigate the effects of segregation phenomena and of partial crystallization and formation of small cristobalite inclusions which decrease K/sub Ic/ by approx. 25%

  11. 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)

  12. Effect of composition and temperature on viscosity and electrical conductivity of borosilicate glasses for Hanford nuclear waste immobilization

    International Nuclear Information System (INIS)

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

    1993-04-01

    Viscosity and electrical conductivity of 79 simulated borosilicate glasses in the expected range of compositions to be produced in the Hanford Waste Vitrification Plant were measured within the temperature span from 950 to 1250 degree C. The nine major oxide components were SiO 2 , B 2 O 3 , Li 2 O, Na 2 O, CaO, MgO, Fe 2 O 3 , Al 2 O 3 , and ZrO 2 . The test compositions were generated statistically. The data were fitted by Fulcher and Arrhenius equations with temperature coefficients being multilinear functions of the mass fractions of the oxide components. Mixture models were also developed for the natural logarithm of viscosity and that of electrical conductivity at 1150 degree C. Least squares regression was used to obtain component coefficients for all the models

  13. 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

  14. 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.

  15. Characterization of waste of soda-lime glass generated from lapping process to reuse as filler in composite materials as thermal insulation

    Directory of Open Access Journals (Sweden)

    A. C. P. Galvão

    2015-09-01

    Full Text Available AbstractThe beneficiation plate process by soda-lime glass lapping in the glass industry generates, an untapped residue (waste. The waste of this material is sent to landfills, causing impact on the environment. This work aimed to characterize and evaluate the waste of soda-lime glass (GP lapping. After its acquisition, the GP was processed by grinding and sieving and further characterized by the chemical/mineralogical analysis (XRF, EDS and XRD, SEM morphology, particle size by laser diffraction, thermogravimetric analyses (TGA and DSC and thermophysical analyses. It was observed that the GP particles are irregular and micrometric with the predominant presence of Na, Si and Ca elements characteristic of amorphous soda-lime glass. The assessment of the chemical/mineralogical, morphological, thermophysical and thermal gravimetric characteristics of GP suggest its reuse as reinforcing fillers or filler in composite materials to obtain thermal insulation.

  16. 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

  17. 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

  18. Property/composition relationships for Hanford high-level waste glasses melting at 115 degrees C volume 1: Chapters 1-11

    International Nuclear Information System (INIS)

    Hrma, P.R.; Piepel, G.F.

    1994-12-01

    A Composition Variation study (CVS) is being performed within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) project in support of a future high-level nuclear waste vitrification plant at the Hanford site in Washington. From 1989 to 1994, over 120 nonradioactive glasses were melted and properties measured in five statistically-designed experimental phases. Glass composition is represented by the 10 components SiO 2 , B 2 O 3 , Al 2 O 3 , Fe 2 O 3 , ZrO 2 , Na 2 O, Li 2 O, CaO, MgO, and Others (all remaining components). The properties measured include viscosity (η), electrical conductivity (ε), glass transition temperature (T g ), thermal expansion of solid glass (α s ) and molten glass (α m ), crystallinity (quenched and canister centerline cooled glasses), liquidus temperature (T L ), durability based on normalized elemental releases from the Materials Characterization Center-1 28-day dissolution test (MCC-1, r mi ) and the 7-day Product Consistency Test (PCT, r pi ), and solution pHs from MCC-1 and PCT. Amorphous phase separation was also evaluated. Empirical first- and second-order mixture models were fit using the CVS data to relate the various properties to glass composition. Equations for calculating the uncertainty associated with property values predicted by the models were also developed. The models were validated using both internal and external data. Other modeling approaches (e.g., non-bridging oxygen, free energy of hydration, phase-equilibria T L ) were investigated for specific properties. A preliminary Qualified Composition Region was developed to identify glass compositions with high confidence of being processable in a melter and meeting waste form acceptance criteria

  19. Property/composition relationships for Hanford high-level waste glasses melting at 1150 degrees C volume 2: Chapters 12-16 and appendices A-K

    International Nuclear Information System (INIS)

    Hrma, P.R.; Piepel, G.F.

    1994-12-01

    A Composition Variation Study (CVS) is being performed within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) project in support of a future high-level nuclear waste vitrification plant at the Hanford site in Washington. From 1989 to 1994, over 120 nonradioactive glasses were melted and properties measured in five statistically-designed experimental phases. Glass composition is represented by the 10 components SiO 2 , B 2 O 3 , ZrO 2 , Na 2 O, Li 2 O, CaO, MgO, and Others (all remaining components). The properties measured include viscosity (η), electrical conductivity (ε), glass transition temperature (T g ), thermal expansion of solid glass (α s ) and molten glass (α m ), crystallinity (quenched and canister centerline cooled glasses), liquidus temperature (T L ), durability based on normalized elemental releases from the Materials Characterization Center-1 28-day dissolution test (MCC-1, r mi ) and the 7-day Product Consistency Test (PCT, r pi ), and solution pHs from MCC-1 and PCT. Amorphous phase separation was also evaluated. Empirical first- and second-order mixture models were fit using the CVS data to relate the various properties to glass composition. Equations for calculating the uncertainty associated with property values predicted by the models were also developed. The models were validated using both internal and external data. Other modeling approaches (e.g., non-bridging oxygen, free energy of hydration, phase-equilibria T L ) were investigated for specific properties. A preliminary Qualified Composition Region was developed to identify glass compositions with high confidence of being processable in a melter and meeting waste form acceptance criteria

  20. Glass composition development for plasma processing of Hanford high sodium content low-level radioactive liquid waste

    International Nuclear Information System (INIS)

    Marra, J.C.

    1995-02-01

    To assess the acceptability of prospective compositions, response criteria based on durability, homogeneity, viscosity and volatility were defined. Response variables were weighted: durability 35%, homogeneity 25%, viscosity 25%, volatility 15%. A Plackett-Burman experimental design was used to define the first twelve glass formulations. Glass former additives included Al2O3, B2O3, CaO, Li2O, ZrO2 and SiO2. Lithia was added to facilitate fritting of the additives. The additives were normalized to silica content to ease experimental matrix definition and glass formulation. Preset high and low values of these ratios were determined for the initial twelve melts. Based on rankings of initial compositions, new formulations for testing were developed based on a simplex algorithm. Rating and ranking of subsequent compositions continued until no apparent improvement in glass quality was achieved in newly developed formulations. An optimized composition was determined by averaging the additive component values of the final best performing compositions. The glass former contents to form the optimized glass were: 16.1 wt % Al2O3, 12.3 wt % B2O3, 5.5 wt % CaO, 1.7 wt % Li2O, 3.3 wt % ZrO2, 61.1 wt % SiO2. An optimized composition resulted after only 25 trials despite studying six glass additives. A vitrification campaign was completed using a small-scale Joule heated melter. 80 lbs of glass was produced over 96 hours of continuous operation. Several salt compounds formed and deposited on melter components during the run and likely caused the failure of several pour chamber heaters. In an attempt to minimize sodium volatility, several low or no boron glasses were formulated. One composition containing no boron produced a homogeneous glass worthy of additional testing

  1. DEFENSE HIGH LEVEL WASTE GLASS DEGRADATION

    International Nuclear Information System (INIS)

    Ebert, W.

    2001-01-01

    provide models and parameter values that can be used to calculate the dissolution rates for the different modes of water contact. The analyses were conducted to identify key aspects of the mechanistic model for glass dissolution to be included in the abstracted models used for PA calculations, evaluate how the models can be used to calculate bounding values of the glass dissolution rates under anticipated water contact modes in the disposal. system, and determine model parameter values for the range of potential waste glass compositions and anticipated environmental conditions. The analysis of a bounding rate also considered the effects of the buildup of glass corrosion products in the solution contacting the glass and potential effects of alteration phase formation. Note that application of the models and model parameter values is constrained to the anticipated range of HLW glass compositions and environmental conditions. The effects of processes inherent to exposure to humid air and dripping water were not modeled explicitly. Instead, the impacts of these processes on the degradation rate were taken into account by using empirically measured parameter values. These include the rates at which water sorbs onto the glass, drips onto the glass, and drips off of the glass. The dissolution rates of glasses that were exposed to humid air and dripping water measured in laboratory tests are used to estimate model parameter values for contact by humid air and dripping water in the disposal system

  2. 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

  3. 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.

  4. 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)

  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. 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.

  7. 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.

  8. Bulk metallic glass matrix composites

    International Nuclear Information System (INIS)

    Choi-Yim, H.; Johnson, W.L.

    1997-01-01

    Composites with a bulk metallic glass matrix were synthesized and characterized. This was made possible by the recent development of bulk metallic glasses that exhibit high resistance to crystallization in the undercooled liquid state. In this letter, experimental methods for processing metallic glass composites are introduced. Three different bulk metallic glass forming alloys were used as the matrix materials. Both ceramics and metals were introduced as reinforcement into the metallic glass. The metallic glass matrix remained amorphous after adding up to a 30 vol% fraction of particles or short wires. X-ray diffraction patterns of the composites show only peaks from the second phase particles superimposed on the broad diffuse maxima from the amorphous phase. Optical micrographs reveal uniformly distributed particles in the matrix. The glass transition of the amorphous matrix and the crystallization behavior of the composites were studied by calorimetric methods. copyright 1997 American Institute of Physics

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  10. 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

  11. GLASS-FIBRE REINFORCED COMPOSITES: THE EFFECT OF ...

    African Journals Online (AJOL)

    HOD

    mechanical and corrosion wear behaviour of any reinforced composites. In other ..... physical properties of glass fibre reinforced epoxy resin and the following .... waste in concrete and cement composites," Journal of Cleaner Production, vol.

  12. Development of a sintering process for recycling oil shale fly ash and municipal solid waste incineration bottom ash into glass ceramic composite

    International Nuclear Information System (INIS)

    Zhang, Zhikun; Zhang, Lei; Li, Aimin

    2015-01-01

    Highlights: • Glass ceramic composite is prepared from oil shale fly ash and MSWI bottom ash. • A novel method for the production of glass ceramic composite is presented. • It provides simple route and lower energy consumption in terms of recycling waste. • The vitrified slag can promote the sintering densification process of glass ceramic. • The performances of products decrease with the increase of oil shale fly ash content. - Abstract: Oil shale fly ash and municipal solid waste incineration bottom ash are industrial and municipal by-products that require further treatment before disposal to avoid polluting the environment. In the study, they were mixed and vitrified into the slag by the melt-quench process. The obtained vitrified slag was then mixed with various percentages of oil shale fly ash and converted into glass ceramic composites by the subsequent sintering process. Differential thermal analysis was used to study the thermal characteristics and determine the sintering temperatures. X-ray diffraction analysis was used to analyze the crystalline phase compositions. Sintering shrinkage, weight loss on ignition, density and compressive strength were tested to determine the optimum preparation condition and study the co-sintering mechanism of vitrified amorphous slag and oil shale fly ash. The results showed the product performances increased with the increase of sintering temperatures and the proportion of vitrified slag to oil shale fly ash. Glass ceramic composite (vitrified slag content of 80%, oil shale fly ash content of 20%, sintering temperature of 1000 °C and sintering time of 2 h) showed the properties of density of 1.92 ± 0.05 g/cm 3 , weight loss on ignition of 6.14 ± 0.18%, sintering shrinkage of 22.06 ± 0.6% and compressive strength of 67 ± 14 MPa. The results indicated that it was a comparable waste-based material compared to previous researches. In particular, the energy consumption in the production process was reduced

  13. Development of a sintering process for recycling oil shale fly ash and municipal solid waste incineration bottom ash into glass ceramic composite

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Zhikun; Zhang, Lei; Li, Aimin, E-mail: leeam@dlut.edu.cn

    2015-04-15

    Highlights: • Glass ceramic composite is prepared from oil shale fly ash and MSWI bottom ash. • A novel method for the production of glass ceramic composite is presented. • It provides simple route and lower energy consumption in terms of recycling waste. • The vitrified slag can promote the sintering densification process of glass ceramic. • The performances of products decrease with the increase of oil shale fly ash content. - Abstract: Oil shale fly ash and municipal solid waste incineration bottom ash are industrial and municipal by-products that require further treatment before disposal to avoid polluting the environment. In the study, they were mixed and vitrified into the slag by the melt-quench process. The obtained vitrified slag was then mixed with various percentages of oil shale fly ash and converted into glass ceramic composites by the subsequent sintering process. Differential thermal analysis was used to study the thermal characteristics and determine the sintering temperatures. X-ray diffraction analysis was used to analyze the crystalline phase compositions. Sintering shrinkage, weight loss on ignition, density and compressive strength were tested to determine the optimum preparation condition and study the co-sintering mechanism of vitrified amorphous slag and oil shale fly ash. The results showed the product performances increased with the increase of sintering temperatures and the proportion of vitrified slag to oil shale fly ash. Glass ceramic composite (vitrified slag content of 80%, oil shale fly ash content of 20%, sintering temperature of 1000 °C and sintering time of 2 h) showed the properties of density of 1.92 ± 0.05 g/cm{sup 3}, weight loss on ignition of 6.14 ± 0.18%, sintering shrinkage of 22.06 ± 0.6% and compressive strength of 67 ± 14 MPa. The results indicated that it was a comparable waste-based material compared to previous researches. In particular, the energy consumption in the production process was reduced

  14. 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

  15. 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.

  16. 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.

  17. 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

  18. 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...

  19. 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

  20. Kinetic Rate Law Parameter Measurements on a Borosilicate Waste Glass: Effect of Temperature, pH, and Solution Composition on Alkali Ion Exchange

    International Nuclear Information System (INIS)

    Pierce, Eric M.; McGrail, B PETER.; Icenhower, J P.; Rodriguez, Elsa A.; Steele, Jackie L.; Baum, Steven R.

    2004-01-01

    The reaction kinetics of glass is controlled by matrix dissolution and ion exchange (IEX). Dissolution of an alkali-rich simulated borosilicate waste glass was investigated using single-pass flow-through (SPFT) experiments. Experiments were conducted as a function of temperature, pH, and solution composition by varying the SiO 2 (aq) activity in the influent solution. Results showed that under dilute conditions matrix dissolution increased with increasing pH and temperature, and decreased with increasing SiO 2 (aq) activity. IEX rates decreased with increasing pH and temperature, and increased with increasing SiO 2 (aq) activity. Over the solution composition range interrogated in this study the dominant dissolution mechanism changed from matrix dissolution to IEX. These results suggest that ''secondary'' reactions may become dominant under certain environmental conditions and emphasize the need to incorporate these reactions into dissolution rate models

  1. Thermal conductivity of glass copper-composite

    International Nuclear Information System (INIS)

    Kinoshita, Makoto; Terai, Ryohei; Haidai, Haruki

    1980-01-01

    Glass-metal composites are to be one of the answers for promoting thermal conduction in the glassy solids containing high-level radioactive wastes. In order to investigate the effect of metal addition on thermal conductivity of glasses, glass-copper composites were selected, and the conductivities of the composites were measured and discussed in regards to copper content and microstructure. Fully densified composites were successfully prepared by pressure sintering of the powder mixtures of glass and copper at temperatures above the yield points of the constituent glasses if the copper content was not so much. The conductivity was measured by means of a comparative method, in which the thermal gradient of the specimen was compared with that of quartz glass as standard under thermally steady state. Measurements were carried out at around 50 0 C. The thermal conductivity increased with increasing content of copper depending on the kind of copper powder used. The conductivities of the composites of the same copper content differed considerably each another. Fine copper powder was effective on increasing conductivity, and the conductivity became about threefold of that of glass by mixing the fine copper powder about 10 vol%. For the composites containing the fine copper powder less than 5 vol%, the conductivity obeyed so-called logarithmic rule, one of the mixture rules of conductivity, whereas for composites containing more than 5 vol%, the conductivity remarkably increased apart from the rule. This fact suggests that copper becomes continuous in the composite when the copper content increased beyond 5 vol%. For the composites containing coarse copper powder, the conductivity was increased not significantly, and obeyed an equation derived from the model in which conductive material dispersed in less conductive one. (author)

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 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.

  7. 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.)

  8. 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

  9. Glass-Graphite Composite Materials

    International Nuclear Information System (INIS)

    Mayzan, M.Z.H.; Lloyd, J.W.; Heath, P.G.; Stennett, M.C.; Hyatt, N.C.; Hand, R.J.

    2016-01-01

    A summary is presented of investigations into the potential of producing glass-composite materials for the immobilisation of graphite or other carbonaceous materials arising from nuclear power generation. The methods are primarily based on the production of base glasses which are subsequently sintered with powdered graphite or simulant TRISO particles. Consideration is also given to the direct preparation of glass-graphite composite materials using microwave technology. Production of dense composite wasteforms with TRISO particles was more successful than with powdered graphite, as wasteforms containing larger amounts of graphite were resistant to densification and the glasses tried did not penetrate the pores under the pressureless conditions used. Based on the results obtained it is concluded that the production of dense glassgraphite composite wasteforms will require the application of pressure. (author)

  10. 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

  11. Development of a sintering process for recycling oil shale fly ash and municipal solid waste incineration bottom ash into glass ceramic composite.

    Science.gov (United States)

    Zhang, Zhikun; Zhang, Lei; Li, Aimin

    2015-04-01

    Oil shale fly ash and municipal solid waste incineration bottom ash are industrial and municipal by-products that require further treatment before disposal to avoid polluting the environment. In the study, they were mixed and vitrified into the slag by the melt-quench process. The obtained vitrified slag was then mixed with various percentages of oil shale fly ash and converted into glass ceramic composites by the subsequent sintering process. Differential thermal analysis was used to study the thermal characteristics and determine the sintering temperatures. X-ray diffraction analysis was used to analyze the crystalline phase compositions. Sintering shrinkage, weight loss on ignition, density and compressive strength were tested to determine the optimum preparation condition and study the co-sintering mechanism of vitrified amorphous slag and oil shale fly ash. The results showed the product performances increased with the increase of sintering temperatures and the proportion of vitrified slag to oil shale fly ash. Glass ceramic composite (vitrified slag content of 80%, oil shale fly ash content of 20%, sintering temperature of 1000 °C and sintering time of 2h) showed the properties of density of 1.92 ± 0.05 g/cm(3), weight loss on ignition of 6.14 ± 0.18%, sintering shrinkage of 22.06 ± 0.6% and compressive strength of 67 ± 14 MPa. The results indicated that it was a comparable waste-based material compared to previous researches. In particular, the energy consumption in the production process was reduced compared to conventional vitrification and sintering method. Chemical resistance and heavy metals leaching results of glass ceramic composites further confirmed the possibility of its engineering applications. Copyright © 2015 Elsevier Ltd. All rights reserved.

  12. 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

  13. Devitrification of defense nuclear waste glasses: role of melt insolubles

    International Nuclear Information System (INIS)

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

    1985-01-01

    Time-temperature-transformation (TTT) curves have been determined for simulated nuclear waste glasses bounding the compositional range in the Defense Waste Processing Facility (DWPF). Formulations include all of the minor chemical elements such as ruthenium and chromium which have limited solubility in borosilicate glasses. Heterogeneous nucleation of spinel on ruthenium dioxide, and subsequent nucleation of acmite on spinel is the major devitrification path. Heterogeneous nucleation on melt insolubles causes more rapid growth of crystalline devitrification phases, than in glass free of melt insolubles. These studies point out the importance of simulating waste glass composition and processing as accurately as possible to obtain reliable estimates of glass performance. 11 refs., 8 figs., 1 tab

  14. Characterization of Savannah River Plant waste glass

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1985-01-01

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

  15. Low melting high lithia glass compositions and methods

    Science.gov (United States)

    Jantzen, Carol M.; Pickett, John B.; Cicero-Herman, Connie A.; Marra, James C.

    2000-01-01

    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.

  16. 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.

  17. 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.

  18. 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)

  19. Chemical Composition Measurements of LAWA44 Glass Samples

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

    DOE is building the Hanford Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is temporarily stored in 177 underground tanks. Both low-activity and high-level wastes will then be vitrified into borosilicate glass using Joule-heated ceramic melters. Efforts are being made to increase the loading of Hanford tank wastes in the glass. One area of work is enhancing waste glass composition/property models and broadening the compositional regions over which those models are applicable. In this report, the Savannah River National Laboratory provides chemical analysis results for several samples of a simulated low-activity waste glass, LAWA44, provided by the 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 97.9 to 102.6 wt %, indicating acceptable recovery of the glass 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 %. It was noted that the measured B2O3 concentrations are somewhat above the targeted values for the study glasses. No obvious trends were observed with regard to the multiple melting steps used to prepare the study glasses, indicating that any potential effects of volatility were below measurable thresholds.

  20. 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.)

  1. 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

  2. Production of an impermeable composite of irradiated graphite and glass by hot isostatic pressing as a long term leach resistant waste form

    Energy Technology Data Exchange (ETDEWEB)

    Fachinger, Johannes; Muller, Walter [FNAG ZU Hanau, Hanau (Germany); Marsat, Eric [FNAG SAS Le Pont de Claix (France); Grosse, Karl-Heinz; Seemann, Richard [ALD Hanau (Germany); Scales, Charlie; Easton, Michael Mark [NNL, Workington (United Kingdom); Anthony Banford [NNL, Warrington (United Kingdom); University of Manchester, Manchester (United Kingdom)

    2013-07-01

    Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. The most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or

  3. Production of an impermeable composite of irradiated graphite and glass by hot isostatic pressing as a long term leach resistant waste form

    International Nuclear Information System (INIS)

    Fachinger, Johannes; Muller, Walter; Marsat, Eric; Grosse, Karl-Heinz; Seemann, Richard; Scales, Charlie; Easton, Michael Mark; Anthony Banford

    2013-01-01

    Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. The most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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%.

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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)

  19. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1982-01-01

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

  20. 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

  1. 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)

  2. 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.

  3. 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.

  4. 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

  5. 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

  6. 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

  7. Glass composition and solution speciation effects on stage III dissolution

    Energy Technology Data Exchange (ETDEWEB)

    Trivelpiece, Cory L. [Pennsylvania State Univ., University Park, PA (United States); Rice, Jarret A. [Pennsylvania State Univ., University Park, PA (United States); Pantano, Carlo G. [Pennsylvania State Univ., University Park, PA (United States)

    2017-10-03

    To understand and mitigate the onset of Stage III corrosion of multicomponent oxides waste glasses. Stage III refers to a resumption of the high initial rate of glass dissolution in some glass samples that have otherwise exhibited dissolution at the much lower residual rate for a long time (Stage II). Although the onset of Stage III is known to occur concurrently with the precipitation of particular alteration products, the root cause of the transition is still unknown. Certain glass compositions (notably AFCI) and high pH environmental conditions are also associated with this observed transition.

  8. Glass composition and solution speciation effects on stage III dissolution

    International Nuclear Information System (INIS)

    Trivelpiece, Cory L.; Rice, Jarret A.; Pantano, Carlo G.

    2017-01-01

    To understand and mitigate the onset of Stage III corrosion of multicomponent oxides waste glasses. Stage III refers to a resumption of the high initial rate of glass dissolution in some glass samples that have otherwise exhibited dissolution at the much lower residual rate for a long time (Stage II). Although the onset of Stage III is known to occur concurrently with the precipitation of particular alteration products, the root cause of the transition is still unknown. Certain glass compositions (notably AFCI) and high pH environmental conditions are also associated with this observed transition.

  9. 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

  10. 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

  11. 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

  12. 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.

  13. 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)

  14. 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

  15. 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

  16. 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.

  17. 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

  18. 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

  19. Preliminary results of durability testing with borosilicate glass compositions

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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.

  11. Identifying glass compositions in fly ash

    Directory of Open Access Journals (Sweden)

    Katherine eAughenbaugh

    2016-01-01

    Full Text Available In this study, four Class F fly ashes were studied with a scanning electron microscope; the glassy phases were identified and their compositions quantified using point compositional analysis with k-means clustering and multispectral image analysis. The results showed that while the bulk oxide contents of the fly ashes were different, the four fly ashes had somewhat similar glassy phase compositions. Aluminosilicate glasses (AS, calcium aluminosilicate glasses (CAS, a mixed glass, and, in one case, a high iron glass were identified in the fly ashes. Quartz and iron crystalline phases were identified in each fly ash as well. The compositions of the three main glasses identified, AS, CAS, and mixed glass, were relatively similar in each ash. The amounts of each glass were varied by fly ash, with the highest calcium fly ash containing the most of calcium-containing glass. Some of the glasses were identified as intermixed in individual particles, particularly the calcium-containing glasses. Finally, the smallest particles in the fly ashes, with the most surface area available to react in alkaline solution, such as when mixed with portland cement or in alkali-activated fly ash, were not different in composition than the large particles, with each of the glasses represented. The method used in the study may be applied to a fly ash of interest for use as a cementing material in order to understand its potential for reactivity.

  12. 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

  13. 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.)

  14. Defining the Glass Composition Limits for SRS Contaminated Soils

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  15. 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.

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

    International Nuclear Information System (INIS)

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

    1988-06-01

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

  17. 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

  18. 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.

  19. 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

  20. Mineralogical textural and compositional data on the alteration of basaltic glass from Kilauea, Hawaii to 300 degrees C: Insights to the corrosion of a borosilicate glass waste-form

    International Nuclear Information System (INIS)

    Smith, D.K.

    1990-01-01

    Mineralogical, textural and compositional data accompanying greenschist facies metamorphism (to 300 degrees C) of basalts of the East Rift Zone (ERZ), Kilauea, Hawaii may be evaluated relative to published and experimental results for the surface corrosion of borosilicate glass. The ERZ alteration sequence is dominated by intermittent palagonite, interlayered smectite-chlorite, chlorite, and actinolite-epidote-anhydrite. Alteration is best developed in fractures and vesicles where surface reaction layers root on the glass matrix forming rinds in excess of 100 microns thick. Fractures control fluid circulation and the alteration sequence. Proximal to the glass surface, palagonite, Fe-Ti oxides and clays replace fresh glass as the surface reaction layer migrates inwards; away from the surface, amphibole, anhydrite, quartz and calcite crystallize from hydrothermal fluids in contact with the glass. The texture and composition of basaltic glass surfaces are similar to those of a SRL-165 glass leached statically for sixty days at 150 degrees C. While the ERZ reservoir is a complex open system, conservative comparisons between the alteration of ERZ and synthetic borosilicate glass are warranted. 31 refs., 2 figs

  1. 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.

  2. 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

  3. 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)

  4. Glasses, ceramics, and composites from lunar materials

    Science.gov (United States)

    Beall, George H.

    1992-01-01

    A variety of useful silicate materials can be synthesized from lunar rocks and soils. The simplest to manufacture are glasses and glass-ceramics. Glass fibers can be drawn from a variety of basaltic glasses. Glass articles formed from titania-rich basalts are capable of fine-grained internal crystallization, with resulting strength and abrasion resistance allowing their wide application in construction. Specialty glass-ceramics and fiber-reinforced composites would rely on chemical separation of magnesium silicates and aluminosilicates as well as oxides titania and alumina. Polycrystalline enstatite with induced lamellar twinning has high fracture toughness, while cordierite glass-ceramics combine excellent thermal shock resistance with high flexural strengths. If sapphire or rutile whiskers can be made, composites of even better mechanical properties are envisioned.

  5. 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.

  6. 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

  7. 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

  8. 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)

  9. 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)

  10. 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...

  11. 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

  12. 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.

  13. Composite waste analysis system

    International Nuclear Information System (INIS)

    Wachter, J.R.; Hagan, R.C.; Bonner, C.A.; Malcom, J.E.; Camp, K.L.

    1993-01-01

    Nondestructive analysis (NDA) of radioactive waste forms an integral component of nuclear materials accountability programs and waste characterization acceptance criterion. However, waste measurements are often complicated by unknown isotopic compositions and the potential for concealment of special nuclear materials in a manner that is transparent to gamma-ray measurement instruments. To overcome these complications, a new NDA measurement system has been developed to assay special nuclear material in both transuranic and low level waste from the same measurement platform. The system incorporates a NaI detector and customized commercial software routines to measure small quantities of radioactive material in low level waste. Transuranic waste analysis is performed with a coaxial HPGE detector and uses upgraded PC-based segmented gamma scanner software to assay containers up to 55 gal. in volume. Gamma-Ray isotopics analysis of both waste forms is also performed with this detector. Finally, a small neutron counter using specialized software is attached to the measurement platform to satisfy safeguards concerns related to nuclear materials that are not sensed by the gamma-ray instruments. This report describes important features and capabilities of the system and presents a series of test measurements that are to be performed to define system parameters

  14. 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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  16. 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

  17. 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.

  18. 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

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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

  7. Celsian Glass-Ceramic Matrix Composites

    Science.gov (United States)

    Bansal, Narottam P.; Dicarlo, James A.

    1996-01-01

    Glass-ceramic matrix reinforced fiber composite materials developed for use in low dielectric applications, such as radomes. Materials strong and tough, exhibit low dielectric properties, and endure high temperatures.

  8. 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)

  9. Hot isostatic pressing of glass-zeolite composites

    International Nuclear Information System (INIS)

    Hash, M.C.; Pereira, C.; Lewis, M.A.

    1996-01-01

    Glass-zeolite waste forms are being developed for immobilizing the chloride waste salt generated from the electrometallurgical treatment of spent fuel. Glass-zeolite composites with high densities were made using hot isostatic pressing (HIP) techniques. Processing parameters were investigated to yield desirable structural ceramic properties such as mechanical, chemical, and thermal stability. Limits for these parameters were determined by differential thermal and thermogravimetric analysis. The resulting ceramic properties such as bulk density, open or apparent porosity, and leach resistance were determined. In addition, phase equilibria and particle-size distribution were observed by optical light and electron microscopy. Pre-HIP processing techniques were also studied to ensure intimate mixing of the glass and zeolite powders. Particle size distributions resulting from dry blending procedure are appropriate for needed flow and packing characteristics

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-30

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

  11. 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.

  12. 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)

  13. 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...

  14. 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.

  15. 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.

  16. 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

  17. 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

  18. 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), ...

  19. 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

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

    International Nuclear Information System (INIS)

    Kruger, A.A.; Jain, V.

    2009-01-01

    A computational method for formulating Hanford HLW glasses was developed that is based on empirical glass composition-property models, accounts for all associated uncertainties, and can be solved in Excel R in minutes. Calculations for all waste form processing and compliance requirements included. Limited experimental validation performed.

  1. GLASS FORMULATION FOR THE HANFORD TANK WASTE TREATMENT AND IMMOBILIZATION PLANT (WTP)

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; VIENNA JD; KIM DS; JAIN V

    2009-05-27

    A computational method for formulating Hanford HLW glasses was developed that is based on empirical glass composition-property models, accounts for all associated uncertainties, and can be solved in Excel{sup R} in minutes. Calculations for all waste form processing and compliance requirements included. Limited experimental validation performed.

  2. 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

  3. 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

  4. 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.

  5. 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

  6. Molybdenum sealing glass-ceramic composition

    International Nuclear Information System (INIS)

    Eagan, R.J.

    1976-01-01

    A glass-ceramic composition is described having low hydrogen and helium permeability properties, along with high fracture strength, and a thermal coefficient of expansion similar to that of molybdenum. The composition is adaptable for hermetically sealing to molybdenum at temperatures between 900 and about 950 0 C to form a hermetically sealed insulator body

  7. Preparation of fullerene/glass composites

    Science.gov (United States)

    Mattes, Benjamin R.; McBranch, Duncan W.; Robinson, Jeanne M.; Koskelo, Aaron C.; Love, Steven P.

    1995-01-01

    Synthesis of fullerene/glass composites. A direct method for preparing solid solutions of C.sub.60 in silicon dioxide (SiO.sub.2) glass matrices by means of sol-gel chemistry is described. In order to produce highly concentrated fullerene-sol-gel-composites it is necessary to increase the solubility of these "guests" in a delivery solvent which is compatible with the starter sol (receiving solvent). Sonication results in aggregate disruption by treatment with high frequency sound waves, thereby accelerating the rate of hydrolysis of the alkoxide precursor, and the solution process for the C.sub.60. Depending upon the preparative procedure, C.sub.60 dispersed within the glass matrix as microcrystalline domains, or dispersed as true molecular solutions of C.sub.60 in a solid glass matrix, is generated by the present method.

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

    International Nuclear Information System (INIS)

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

    1987-12-01

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

  9. 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.

  10. 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

  11. 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.

  12. 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)

  13. 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.

  14. 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.)

  15. 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...

  16. 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

  17. Composition and property measurements for PHA Phase 4 glasses

    International Nuclear Information System (INIS)

    Edwards, T.B.

    2000-01-01

    The results presented in this report are for nine Precipitate Hydrolysis Aqueous (PHA) Phase 4 glasses. Three of the glasses contained HM sludge at 22, 26, and 30 wt% respectively, 10 wt% PHA and 1.25 wt% monosodium titanate (MST), all on an oxide basis. The remaining six glasses were selected from the Phase 1 and Phase 2 studies (Purex sludge) but with an increased amount of MST. The high-end target for MST of 2.5 wt% oxide was missed in Phases 1 and 2 due to ∼30 wt% water content of the MST. A goal of this Phase 4 study was to determine whether this increase in titanium concentration from the MST had any impact on glass quality or processibility. Two of the glasses, pha14c and pha15c, were rebatched and melted due to apparent batching errors with pha14 and pha15. The models currently in the Defense Waste Processing Facility's (DWPF) Product Composition Control System (PCCS) were used to predict durability, homogeneity, liquidus, and viscosity for these nine glasses. All of the HM glasses and half of the Purex glasses were predicted to be phase separated, and consequently prediction of glass durability is precluded with the cument models for those glasses that failed the homogeneity constraint. If one may ignore the homogeneity constraint, the measured durabilities were within the 95% prediction limits of the model. Further efforts will be required to resolve this issue on phase separation (inhomogeneity). The liquidus model predicted unacceptable liquidus temperatures for four of the nine glasses. The approximate, bounding liquidus temperatures measured for all had upper limits of 1,000 C or less. Given the fact that liquidus temperatures were only approximated, the 30 wt% loading of Purex may be near or at the edge of acceptability for liquidus. The measured viscosities were close to the predictions of the model. For the Purex glasses, pha12c and pha15c, the measured viscosities of 28 and 23 poise, respectively, indicate that DWPF processing may be compromised

  18. Solid oxide fuel cell having a glass composite seal

    Science.gov (United States)

    De Rose, Anthony J.; Mukerjee, Subhasish; Haltiner, Jr., Karl Jacob

    2013-04-16

    A solid oxide fuel cell stack having a plurality of cassettes and a glass composite seal disposed between the sealing surfaces of adjacent cassettes, thereby joining the cassettes and providing a hermetic seal therebetween. The glass composite seal includes an alkaline earth aluminosilicate (AEAS) glass disposed about a viscous glass such that the AEAS glass retains the viscous glass in a predetermined position between the first and second sealing surfaces. The AEAS glass provides geometric stability to the glass composite seal to maintain the proper distance between the adjacent cassettes while the viscous glass provides for a compliant and self-healing seal. The glass composite seal may include fibers, powders, and/or beads of zirconium oxide, aluminum oxide, yttria-stabilized zirconia (YSZ), or mixtures thereof, to enhance the desirable properties of the glass composite seal.

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-18

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  2. 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

  3. 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.

  4. 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.

  5. Effect of small glass composition changes on flue gas emissions of glass furnaces

    NARCIS (Netherlands)

    Limpt, J.A.C. van; Beerkens, R.G.C.; Kersbergen, M.J. van

    2008-01-01

    Relatively small changes in glass composition might have drastic consequences on the evaporation rates of volatile glass components in glass melting furnaces. Transpiration evaporation tests have been applied to measure the impact of minor glass composition changes on the evaporation rates of

  6. 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

  7. 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

  8. 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)

  9. Phase 2B experimental design for the INEEL glass composition variation study

    International Nuclear Information System (INIS)

    Peeler, D.

    2000-01-01

    The U.S. Department of Energy's (DOE's) Offices of Science and Technology (through the Tanks Focus Area [TFA]) and Waste Management are sponsoring a partnership among Idaho National Engineering and Environmental Laboratory (INEEL), Pacific Northwest National Laboratory (PNNL), and the Savannah River Technology Center (SRTC) for a collaborative glass composition variation study (CVS). The purpose of the CVS is to investigate property - composition relationships within a glass-composition region compatible with the expected range of Idaho Nuclear Technology and Engineering Center (INTEC) high-activity wastes (HAWs). The CVS has been conducted in phases to allow INEEL, PNNL, and SRTC researchers to adjust the glass composition region of interest as flowsheet options are refined and/or waste-stream compositions become more defined

  10. 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

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

    Czech Academy of Sciences Publication Activity Database

    Harris, W.H.; Guillen, D.P.; Kloužek, Jaroslav; Pokorný, P.; Yano, T.; Lee, S.; Schweiger, M. J.; Hrma, P.

    2017-01-01

    Roč. 100, č. 9 (2017), s. 3883-3894 ISSN 0002-7820 Institutional support: RVO:67985891 Keywords : borosilicate glass * computed tomography * glass melting * morphology * nuclear waste * X-ray Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass OBOR OECD: Ceramics Impact factor: 2.841, year: 2016

  12. 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

  13. 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

  14. Composite quarterly technical report: long-term high-level waste technology, October-December 1980

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-04-01

    The technical information in this report summarizes work performed at participating sites to immobilize high-level radioactive wastes. The areas reported are in: program management and support; waste preparation; waste fixation; and final handling. Majority of the studies were in the area of waste fixation, some of which are: leaching tests of ceramic forms, high silica glass, graphite powder and other carbon preparations; viscosity measurements for a range of waste-glass compositions from references borosilicate glass to high-alumina glasses; neutron activation analysis for measuring leach rates; preparation of SYNROC D spheres; formulations for preparing ceramics from defense waste composition; development of a pilot-scale glass melter, and kinetic studies of slag formation in glass melters

  15. 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.%

  16. 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.

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  18. 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...

  19. 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.

  20. 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

  1. 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

  2. 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.

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  8. Impacts of Process and Prediction Uncertainties on Projected Hanford Waste Glass Amount

    Energy Technology Data Exchange (ETDEWEB)

    Gervasio, Vivianaluxa [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vienna, John D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kim, Dong-Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kruger, Albert A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2018-02-19

    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 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 IHLW glass when no uncertainties were taken into accound. Accounting for prediction and composition/process uncertainties resulted in 5.01 relative percent increase in estimated glass mass 24,531 MT. Roughly equal impacts were found for prediction uncertainties (2.58 RPD) and composition/process uncertainties (2.43 RPD). ILAW mass was predicted to be 282,350 MT without uncertainty and with weaste 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 MTG. 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.

  9. 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.

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  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. Evaluation the microwave heating of spinel crystals in high-level waste glass

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Ecology Lab. (SREL); Washington, A. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Ecology Lab. (SREL)

    2015-08-18

    In this report, the microwave heating of a crystal-free and a partially (24 wt%) trevorite-crystallized waste glass simulant were evaluated. The results show that a 500 mg piece of partially crystallized waste glass can be heated from room-temperature to above 1600 °C (as measured by infrared radiometry) within 2 minutes using a single mode, highly focused, 2.45 GHz microwave, operating at 300 W. X-ray diffraction measurements show that the partially crystallized glass experiences an 87 % reduction in trevorite following irradiation and thermal quenching. When a crystal-free analogue of the same waste glass simulant composition is exposed to the same microwave radiation it could not be heated above 450 °C regardless of the heating time.

  13. 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

  14. Long-term modeling of glass waste in portland cement- and clay-based matrices

    International Nuclear Information System (INIS)

    Stockman, H.W.; Nagy, K.L.; Morris, C.E.

    1995-12-01

    A set of ''templates'' was developed for modeling waste glass interactions with cement-based and clay-based matrices. The templates consist of a modified thermodynamic database, and input files for the EQ3/6 reaction path code, containing embedded rate models and compositions for waste glass, cement, and several pozzolanic materials. Significant modifications were made in the thermodynamic data for Th, Pb, Ra, Ba, cement phases, and aqueous silica species. It was found that the cement-containing matrices could increase glass corrosion rates by several orders of magnitude (over matrixless or clay matrix systems), but they also offered the lowest overall solubility for Pb, Ra, Th and U. Addition of pozzolans to cement decreased calculated glass corrosion rates by up to a factor of 30. It is shown that with current modeling capabilities, the ''affinity effect'' cannot be trusted to passivate glass if nuclei are available for precipitation of secondary phases that reduce silica activity

  15. 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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  17. DM100 AND DM1200 MELTER TESTING WITH HIGH WASTE LOADING GLASS FORMULATIONS FOR HANFORD HIGH-ALUMINUM HLW STREAMS

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS; KOT WK; PEGG IL; JOSEPH I

    2009-12-30

    This Test Plan describes work to support the development and testing of high waste loading glass formulations that achieve high glass melting rates for Hanford high aluminum high level waste (HLW). In particular, the present testing is designed to evaluate the effect of using low activity waste (LAW) waste streams as a source of sodium in place ofchemical additives, sugar or cellulose as a reductant, boehmite as an aluminum source, and further enhancements to waste processing rate while meeting all processing and product quality requirements. The work will include preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM 100) tests designed to examine the effects of enhanced glass formulations, glass processing temperature, incorporation of the LAW waste stream as a sodium source, type of organic reductant, and feed solids content on waste processing rate and product quality. Also included is a confirmatory test on the HLW Pilot Melter (DM1200) with a composition selected from those tested on the DM100. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy's (DOE's) Office of River Protection (ORP) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same waste composition. This Test Plan is 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 Hanford Tank Waste Treatment and Immobilization Plant (WTP) is about 12,500. This estimate is based upon the inventory ofthe tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat

  18. Effect of composition on peraluminous glass properties: An application to HLW containment

    Science.gov (United States)

    Piovesan, V.; Bardez-Giboire, I.; Perret, D.; Montouillout, V.; Pellerin, N.

    2017-01-01

    Part of the Research and Development program concerning high level nuclear waste (HLW) glasses aims to assess new glass formulations able to incorporate a high waste content with enhanced properties in terms of thermal stability, chemical durability, and process ability. This study focuses on peraluminous glasses of the SiO2 - Al2O3 - B2O3 - Na2O - Li2O - CaO - La2O3 system, defined by an excess of aluminum ions Al3+ in comparison with modifier elements such as Na+, Li+ or Ca2+. To understand the effect of composition on physical properties of glasses (viscosity, density, Tg), a Design Of Experiments (DOE) approach was applied to investigate the peraluminous glass domain. The influence of each oxide was quantified to build predictive models for each property. Lanthanum and lithium oxides appear to be the most influential factors on peraluminous glass properties.

  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. Ceramic nuclear waste forms. II. A ceramic-waste composite prepared by hot pressing. Progress report and preprint

    International Nuclear Information System (INIS)

    McCarthy, G.J.

    1975-01-01

    A feasibility study was conducted to determine whether nuclear waste calcine and a crystalline ceramic matrix can be fabricated by hot pressing into a composite waste form with suitable leaching resistance and thermal stability. It was found that a hard, dense composite could be formed using the typical commercial waste formulation PW-4b and a matrix of α-quartz with a small amount of a lead borosilicate glass added as a consolidation aide. Its density, waste loading, and leaching resistance are comparable to the glasses currently being considered for fixation of nuclear wastes. The hot pressed composite offers a closer approach to thermodynamic stability and improved thermal stability (in monolithic form) compared to glass waste forms. Recommendations for further optimization of the hot pressed waste form are given. (U.S.)

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  2. Fabrication of slag-glass composite with controlled porosity

    Directory of Open Access Journals (Sweden)

    Ranko Adziski

    2008-06-01

    Full Text Available The preparation and performance of porous ceramics made from waste materials were investigated. Slag from thermal electrical plant Kakanj (Bosnia and Herzegovina with defined granulations: (0.500÷0.250 mm; (0.250÷0.125 mm; (0.125÷0.063 mm; (0.063÷0.045 mm and 20/10 wt.% of the waste TV screen glass with a granulation <0.063 mm were used for obtaining slag-glass composites with controlled porosity. The one produced from the slag powder fraction (0.125÷0.063 mm and 20 wt.% TV screen glass, sintered at 950°C/2h, was considered as the optimal. This system possesses open porosity of 26.8±1.0%, and interconnected pores with the size of 250–400 μm. The values of E-modulus and bending strength of this composite were 10.6±0.6 GPa and 45.7±0.7 MPa, respectively. The coefficient of thermal expansion was 8.47·10-6/°C. The mass loss in 0.1M HCl solution after 30 days was 1.2 wt.%. The permeability and the form coefficient of the porous composite were K0=0.12 Da and C0=4.53·105 m-1, respectively. The porous composite shows great potential to be used as filters, diffusers for water aeration, dust collectors, acoustic absorbers, etc.

  3. 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.

  4. 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)

  5. Glass as a medium for the ultimate disposal of highly radioactive waste

    International Nuclear Information System (INIS)

    Sombret, C.

    1983-09-01

    The conversion of high level radioactive liquid wastes into glass is now considered in every nuclear country. The glass composition must take into account the components of the solutions and be formulated in order to meet certain requirements, mainly those necessary for safe further disposal. The compositions of these glasses, all borosilicates, are consequently unusual. Heat due to β γ decay generates some devitrification but it has not yet been demonstrated that this is detrimental. β irradiation has minor effects on the glass structure but the effect of α emitters is not presently totally investigated. If stored energy consequenses are negligible, further experiments must be carried out to ascertain the effect of helium build up or the behaviour of the mechanical properties. Processes of industrial interest have been developped and a plant has already produced radioactive glass blocks for 5 years

  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. 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

  8. 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

  9. Effect of Na{sub 2}O on aqueous dissolution of nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Farooqi, Rahmat Ullah, E-mail: rufarooqi@live.com [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, 77 Cheongam-Ro, Nam-Gu, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Hrma, Pavel [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, 77 Cheongam-Ro, Nam-Gu, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Pacific Northwest National Laboratory, Richland, WA (United States)

    2017-04-15

    Sodium oxide is present in the majority of commercial and waste glasses as a viscosity-reducing component. In some nuclear waste glasses, its source is the waste itself. As such, it can limit the waste loading because of its deleterious effect on the resistance of the glass to attack by aqueous media. The maximum tolerable content of Na{sub 2}O in glass depends on the presence and concentration of components that interact with it. To assess the acceptability limits of Na{sub 2}O in the composition region of nuclear waste glasses, we formulated 11 baseline compositions by varying the content of oxides of Si, B, Al, Ca, Zr, and Li. In each of these compositions, we varied the Na{sub 2}O fraction from 8–16 mass% to 23–30 mass%. To each of 146 glasses thus formulated, we applied the seven-day Product Consistency Test (PCT) to determine normalized B and Na releases (r{sub i}, where i ≡ B or Na). Fitting approximation functions ln(r{sub i}/gm{sup −2}) = Σb{sub ij}g{sub j} to r{sub i} data (g{sub j} is the j-th component mass fraction and b{sub ij} the corresponding component coefficient), we showed that the r{sub B} (and, consequently, the initial glass alteration rate) was proportional to the glass component mass fractions in the order Al{sub 2}O{sub 3}glass structure would fall apart or beyond which a continuous nondurable phase would be separated. Specific examples are given to demonstrate restrictions imposed on the boundary of the composition region of acceptable glasses by the maximum allowable r{sub B} and by the melt viscosity required for glass melter operation. Finally, the role that PCT data may play in understanding the evolution of the glass alteration process is discussed.

  10. 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

  11. Improving subcritical crack growth resistance for alumina glass dental composite

    NARCIS (Netherlands)

    Zhu, Q.; With, de G.

    2005-01-01

    The improvement of subcritical crack growth (SCG) resistance for alumina glass dental composites was explored in this study. The addition of nitrogen to the glass phases in the composite was found to increase the SCG resistance, where the SCG exponent n increases from 22 for the oxide glass

  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. 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.

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

    Directory of Open Access Journals (Sweden)

    V. W. Francis Thoo

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Wronkiewicz, D.J.

    1993-12-01

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

  16. Laboratory stabilization/solidification of surrogate and actual mixed-waste sludge in glass and grout

    International Nuclear Information System (INIS)

    Spence, R.D.; Gilliam, T.M.; Mattus, C.H.; Mattus, A.J.

    1998-01-01

    Grouting and vitrification are currently the most likely stabilization/solidification technologies for mixed wastes. Grouting has been used to stabilize and solidify hazardous and low-level waste for decades. Vitrification has long been developed as a high-level-waste alternative and has been under development recently as an alternative treatment technology for low-level mixed waste. Laboratory testing has been performed to develop grout and vitrification formulas for mixed-waste sludges currently stored in underground tanks at Oak Ridge National Laboratory (ORNL) and to compare these waste forms. Envelopes, or operating windows, for both grout and soda-lime-silica glass formulations for a surrogate sludge were developed. One formulation within each envelope was selected for testing the sensitivity of performance to variations (±10 wt%) in the waste form composition and variations in the surrogate sludge composition over the range previously characterized in the sludges. In addition, one sludge sample of an actual mixed-waste tank was obtained, a surrogate was developed for this sludge sample, and grout and glass samples were prepared and tested in the laboratory using both surrogate and the actual sludge. The sensitivity testing of a surrogate tank sludge in selected glass and grout formulations is discussed in this paper, along with the hot-cell testing of an actual tank sludge sample

  17. 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

  18. Glass Transition Temperature- and Specific Volume- Composition Models for Tellurite Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Vienna, John D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2017-09-01

    This report provides models for predicting composition-properties for tellurite glasses, namely specific gravity and glass transition temperature. Included are the partial specific coefficients for each model, the component validity ranges, and model fit parameters.

  19. 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

  20. 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

  1. 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

  2. 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

  3. Reuse of waste of glass wool in the production of mortar

    International Nuclear Information System (INIS)

    Vieira, P.L.C.; Santos, N.A.; Louzada, D.M.; Araujo, G.S.; Della, V.P.

    2014-01-01

    In recent years, the incorporation of alternative materials, especially waste, in mortars and concretes has become a common practice in the building industry. Against this background, this paper seeks to examine the possibility of using waste glass wool resulting from the steel industry in mortars in partial replacement the thin fraction of sand. To the knowledge of their chemical and mineralogical composition, the waste was subjected to x-ray fluorescence and diffraction assays. Mortars with different percentages of incorporation of waste were produced and performed flow test assays, Water Absorption by capillarity, compressive strength and compressive flexural strength. The results were compared with a reference mortar without residue. (author)

  4. High level waste containing granules coated and embedded in metal as an alternative to HLW glasses

    International Nuclear Information System (INIS)

    Neumann, W.

    1980-01-01

    Simulated high level waste containing granules were overcoated with pyrocarbon or nickel respectively. The coatings were performed by the use of chemical vapour deposition in a fluidized bed. The coated granules were embedded in an aluminium-silicon-alloy to improve the dissipation of radiation induced heat. The metal-granules-composites obtained were of improved product stability related to the high level waste containing glasses. (orig.) [de

  5. TIME-TEMPERATURE-TRANSFORMATION (TTT) DIAGRAMS FOR FUTURE WASTE COMPOSITIONS

    International Nuclear Information System (INIS)

    Billings, A.; Edwards, T.

    2010-01-01

    As a part of the Waste Acceptance Product Specifications (WAPS) for Vitrified High-Level Waste Forms defined by the Department of Energy - Office of Environmental Management, the waste form stability must be determined for each of the projected high-level waste (HLW) types at the Savannah River Site (SRS). Specifically, WAPS 1.4.1 requires the glass transition temperature (T g ) to be defined and time-temperature-transformation (TTT) diagrams to be developed. The T g of a glass is an indicator of the approximate temperature where the supercooled liquid converts to a solid on cooling or conversely, where the solid begins to behave as a viscoelastic solid on heating. A TTT diagram identifies the crystalline phases that can form as a function of time and temperature for a given waste type or more specifically, the borosilicate glass waste form. In order to assess durability, the Product Consistency Test (PCT) was used and the durability results compared to the Environmental Assessment (EA) glass. The measurement of glass transition temperature and the development of TTT diagrams have already been performed for the seven Defense Waste Processing Facility (DWPF) projected compositions as defined in the Waste Form Compliance Plan (WCP) and in SRNL-STI-2009-00025. Additional phase transformation information exists for other projected compositions, but overall these compositions did not cover composition regions estimated for future waste processing. To develop TTT diagrams for future waste types, the Savannah River National Laboratory (SRNL) fabricated two caches of glass from reagent grade oxides to simulate glass compositions which would be likely processed with and without Al dissolution. These were used for glass transition temperature measurement and TTT diagram development. The glass transition temperatures of both glasses were measured using differential scanning calorimetry (DSC) and were recorded to be 448 C and 452 C. Using the previous TTT diagrams as guidance

  6. Effect of composition on peraluminous glass properties: An application to HLW containment

    Energy Technology Data Exchange (ETDEWEB)

    Piovesan, V. [CEA, DEN, DTCD, SECM, LDMC – Marcoule, F-30207 Bagnols sur Cèze (France); CNRS, CEMHTI UPR3079, Univ. Orléans, F-45071 Orléans (France); Bardez-Giboire, I., E-mail: isabelle.giboire@cea.fr [CEA, DEN, DTCD, SECM, LDMC – Marcoule, F-30207 Bagnols sur Cèze (France); Perret, D. [CEA, DEN, DTCD, SECM, LDMC – Marcoule, F-30207 Bagnols sur Cèze (France); Montouillout, V.; Pellerin, N. [CNRS, CEMHTI UPR3079, Univ. Orléans, F-45071 Orléans (France)

    2017-01-15

    Part of the Research and Development program concerning high level nuclear waste (HLW) glasses aims to assess new glass formulations able to incorporate a high waste content with enhanced properties in terms of thermal stability, chemical durability, and process ability. This study focuses on peraluminous glasses of the SiO{sub 2} – Al{sub 2}O{sub 3} – B{sub 2}O{sub 3} – Na{sub 2}O – Li{sub 2}O – CaO – La{sub 2}O{sub 3} system, defined by an excess of aluminum ions Al{sup 3+} in comparison with modifier elements such as Na{sup +}, Li{sup +} or Ca{sup 2+}. To understand the effect of composition on physical properties of glasses (viscosity, density, T{sub g}), a Design Of Experiments (DOE) approach was applied to investigate the peraluminous glass domain. The influence of each oxide was quantified to build predictive models for each property. Lanthanum and lithium oxides appear to be the most influential factors on peraluminous glass properties. - Highlights: • A Design of Experiment approach to link composition and glass properties. • Adding alkali decreases glass transition temperature. • Adding La{sub 2}O{sub 3} strongly decreases glass melt viscosity. • Adding La{sub 2}O{sub 3} increases density.

  7. 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

  8. Fracture behaviour of brittle (glass) matrix composites

    Czech Academy of Sciences Publication Activity Database

    Dlouhý, Ivo; Chlup, Zdeněk; Boccaccini, A. R.

    2005-01-01

    Roč. 482, - (2005), s. 115-122 ISSN 0255-5476. [International Conference on Materials Structure and Micromechanics of Fracture /4./. Brno, 23.06.2004-25.06.2004] R&D Projects: GA AV ČR(CZ) IAA2041003; GA ČR(CZ) GA101/02/0683 Institutional research plan: CEZ:AV0Z2041904 Keywords : Ceramic matrix composites * fracture toughness * toughening effects Subject RIV: JH - Ceramic s, Fire-Resistant Materials and Glass Impact factor: 0.399, year: 2005

  9. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-01

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

  11. Predicting the glass transition temperature of bioactive glasses from their molecular chemical composition.

    Science.gov (United States)

    Hill, Robert G; Brauer, Delia S

    2011-10-01

    A recently published paper (M.D. O'Donnell, Acta Biomaterialia 7 (2011) 2264-2269) suggests that it is possible to correlate the glass transition temperature (T(g)) of bioactive glasses with their molar composition, based on iterative least-squares fitting of published T(g) data. However, we show that the glass structure is an important parameter in determining T(g). Phase separation, local structural effects and components (intermediate oxides) which can switch their structural role in the glass network need to be taken into consideration, as they are likely to influence the glass transition temperature of bioactive glasses. Although the model suggested by O'Donnell works reasonably well for glasses within the composition range presented, it is oversimplified and fails for glasses outside certain compositional boundaries. Copyright © 2011 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

  12. 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.

  13. Specialty glass development for radiation shielding windows and nuclear waste immobilization

    International Nuclear Information System (INIS)

    Mandal, S.; Ghorui, S.; Roy Chowdhury, A.; Sen, R.; Chakraborty, A.K.; Sen, S.; Maiti, H.S.

    2015-01-01

    The technology of two important varieties of specialty glasses, namely high density Radiation Shielding Window (RSW) glass and specialty glass beads of borosilicate composition have been successfully developed in CGCRI with an aim to meet the countries requirement. Radiation Shielding Windows used in nuclear installations, are viewing devices, which allow direct viewing into radioactive areas while still providing adequate protection to the operating personnel. The glass blocks are stabilized against damage from radiation by introducing cerium in definite proportions. Considering the essentially of developing an indigenous technology to make the country self-sufficient for this critical item, CGCRI has taken up a major programme to develop high lead containing glasses required for RSWs under a MoD with BARC. On the other hand, the specialty glass bead of specific composition and properties is a critical material required for management of radioactive waste in a closed nuclear fuel cycle that is followed by India. During reprocessing of the spent nuclear fuel, high level radio-active liquid waste (HLW) is produced containing unwanted radio isotopes some of which remain radioactive for thousands of years. The need is to immobilize them within a molecular structure so that they will not come out and be released to the ambience and thereby needs to be resolved if nuclear power is to make a significant contribution to the country's power requirement. Borosilicate glass has emerged as the material of choice for immobilization due to its unique random network structure

  14. Secondary phases formed during nuclear waste glass-water interactions: Thermodynamic and derived properties

    International Nuclear Information System (INIS)

    McKenzie, W.F.

    1992-08-01

    The thermodynamic properties of secondary phases observed to form during nuclear waste glass-water interactions are of particular interest as it is with the application of these properties together with the thermodynamic properties of other solid phases, fluid phases, and aqueous species that one may predict the environmental consequences of introducing radionuclides contained in the glass into groundwater at a high-level nuclear waste repository. The validation of these predicted consequences can be obtained from laboratory experiments and field observations at natural analogue sites. The purpose of this report is to update and expand the previous compilation (McKenzie, 1991) of thermodynamic data retrieved from the literature and/or estimated for secondary phases observed to form (and candidate phases from observed chemical compositions) during nuclear waste glass-water interactions. In addition, this report includes provisionally recommended thermodynamic data of secondary phases

  15. 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

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

    International Nuclear Information System (INIS)

    Kawamoto, Takamichi; Terai, Ryohei; Hara, Shigeo

    1978-01-01

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

  17. TSA waste stream and final waste form composition

    International Nuclear Information System (INIS)

    Grandy, J.D.; Eddy, T.L.; Anderson, G.L.

    1993-01-01

    A final vitrified waste form composition, based upon the chemical compositions of the input waste streams, is recommended for the transuranic-contaminated waste stored at the Transuranic Storage Area of the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The quantities of waste are large with a considerable uncertainty in the distribution of various waste materials. It is therefore impractical to mix the input waste streams into an ''average'' transuranic-contaminated waste. As a result, waste stream input to a melter could vary widely in composition, with the potential of affecting the composition and properties of the final waste form. This work examines the extent of the variation in the input waste streams, as well as the final waste form under conditions of adding different amounts of soil. Five prominent Rocky Flats Plant 740 waste streams are considered, as well as nonspecial metals and the ''average'' transuranic-contaminated waste streams. The metals waste stream is the most extreme variation and results indicate that if an average of approximately 60 wt% of the mixture is soil, the final waste form will be predominantly silica, alumina, alkaline earth oxides, and iron oxide. This composition will have consistent properties in the final waste form, including high leach resistance, irrespective of the variation in waste stream. For other waste streams, much less or no soil could be required to yield a leach resistant waste form but with varying properties

  18. 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)

  19. 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)

  20. Control of DWPF [Defense Waste Processing Facility] melter feed composition

    International Nuclear Information System (INIS)

    Edwards, R.E. Jr.; Brown, K.G.; Postles, R.L.

    1990-01-01

    The Defense Waste Processing Facility will be used to immobilize Savannah River Site high-level waste into a stable borosilicate glass for disposal in a geologic repository. Proper control of the melter feed composition in this facility is essential to the production of glass which meets product durability constraints dictated by repository regulations and facility processing constraints dictated by melter design. A technique has been developed which utilizes glass property models to determine acceptable processing regions based on the multiple constraints imposed on the glass product and to display these regions graphically. This system along with the batch simulation of the process is being used to form the basis for the statistical process control system for the facility. 13 refs., 3 figs., 1 tab

  1. A new method locating good glass-forming compositions

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Dechuan [Department of Materials Physics and Chemistry, Northeastern University, No.3-11, Wenhua Road, Shenyang, 110819 (China); Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, 72 Wenhua Road, Shenyang, 110016 (China); Geng, Yan [Department of Materials Physics and Chemistry, Northeastern University, No.3-11, Wenhua Road, Shenyang, 110819 (China); Li, Zhengkun [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, 72 Wenhua Road, Shenyang, 110016 (China); Liu, Dingming [Department of Materials Physics and Chemistry, Northeastern University, No.3-11, Wenhua Road, Shenyang, 110819 (China); Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, 72 Wenhua Road, Shenyang, 110016 (China); Fu, Huameng; Zhu, Zhengwang [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, 72 Wenhua Road, Shenyang, 110016 (China); Qi, Yang, E-mail: qiyang@imp.neu.edu.cn [Department of Materials Physics and Chemistry, Northeastern University, No.3-11, Wenhua Road, Shenyang, 110819 (China); Zhang, Haifeng, E-mail: hfzhang@imr.ac.cn [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, 72 Wenhua Road, Shenyang, 110016 (China)

    2015-10-15

    A new method was proposed to pinpoint the compositions with good glass forming ability (GFA) by combining atomic clusters and mixing entropy. The clusters were confirmed by analyzing competing crystalline phases. The method was applied to the Zr–Al–Ni–Cu–Ag alloy system. A series of glass formers with diameter up to 20 mm were quickly detected in this system. The good glass formers were located only after trying 5 compositions around the calculated composition. The method was also effective in other multi-component systems. This method might provide a new way to understand glass formation and to quickly pinpoint compositions with high GFA. - Highlights: • A new method was proposed to quickly design glass formers with high glass forming ability. • The method of designing pentabasic Zr–Al–Ni–Cu–Ag alloys was applied. • A series of new Zr-based bulk metallic glasses with critical diameter of 20 mm were discovered.

  2. A new method locating good glass-forming compositions

    International Nuclear Information System (INIS)

    Yu, Dechuan; Geng, Yan; Li, Zhengkun; Liu, Dingming; Fu, Huameng; Zhu, Zhengwang; Qi, Yang; Zhang, Haifeng

    2015-01-01

    A new method was proposed to pinpoint the compositions with good glass forming ability (GFA) by combining atomic clusters and mixing entropy. The clusters were confirmed by analyzing competing crystalline phases. The method was applied to the Zr–Al–Ni–Cu–Ag alloy system. A series of glass formers with diameter up to 20 mm were quickly detected in this system. The good glass formers were located only after trying 5 compositions around the calculated composition. The method was also effective in other multi-component systems. This method might provide a new way to understand glass formation and to quickly pinpoint compositions with high GFA. - Highlights: • A new method was proposed to quickly design glass formers with high glass forming ability. • The method of designing pentabasic Zr–Al–Ni–Cu–Ag alloys was applied. • A series of new Zr-based bulk metallic glasses with critical diameter of 20 mm were discovered

  3. 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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  5. 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

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. 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

  11. Glass Composition for the use as a Sealant

    DEFF Research Database (Denmark)

    2014-01-01

    The invention concerns a glass composition for the use as a sealant, particularly in a solid oxide fuel cell (SOFC) or in a solid oxide electrolyser cell (SOEC). The glass composition comprises 35-70 mol% CaO, 5-45 mol% ZnO, 5-50 mol% B2O3, 1-45 mol% SiO2, and 1 mol% or less of each element...... of the group, comprising Ba, Na and Sr, based on the total glass composition. Furthermore, the invention relates to an SOFC and an SOEC employing a sealant of said glass composition....

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

    Energy Technology Data Exchange (ETDEWEB)

    Delbert E. Ray; Chandra S. Ray

    2013-11-01

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

  13. 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

  14. Evolution of mechanical properties of silicate glasses: Impact of the chemical composition and effects of irradiation

    International Nuclear Information System (INIS)

    Barlet, Marina

    2014-01-01

    This thesis examines: (1) how the chemical composition changes the hardness, toughness, and stress corrosion cracking behavior in model pristine and (2) how external irradiation impact these properties. It is to be incorporated in the context of the storage of nuclear waste in borosilicate glass matrix, the structural integrity of which should be assessed. Eight simplified borosilicate glasses made of 3 oxides with modulated proportions (SiO 2 -B 2 O 3 -Na 2 O (SBN) have been selected and their hardness, toughness, and stress corrosion cracking behavior have been characterized prior and after irradiation. The comparative study of the non-irradiated SBN glasses provides the role played by the chemical composition. The sodium content is found to be the key parameter: As it increases, the glass plasticity increases, leading to changes in the mechanical response to strain. Hardness (Hv) and toughness (Kc) decrease since the flow under indenter increases. The analysis of the stress corrosion behavior evidences a clear shift of the SCC curves linked also to the glass plasticity. Four of the 8 simplified SBN glass systems highlight the influence of electron, light and heavy ions irradiations on the mechanical properties. Once again, the sodium content is a key parameter. It is found to inhibit the glass modification: Glasses with high sodium content are more stable. Ions irradiations highlight the predominant role of nuclear interaction in changing the glass properties. Finally, electronic interaction induced by helium and electron irradiation does not lead to the same structural/mechanical glasses variations. (author) [fr

  15. Glass and Glass-Ceramic Materials from Simulated Composition of Lunar and Martian Soils: Selected Properties and Potential Applications

    Science.gov (United States)

    Ray, C. S.; Sen, S.; Reis, S. T.; Kim, C. W.

    2005-01-01

    In-situ resource processing and utilization on planetary bodies is an important and integral part of NASA's space exploration program. Within this scope and context, our general effort is primarily aimed at developing glass and glass-ceramic type materials using lunar and martian soils, and exploring various applications of these materials for planetary surface operations. Our preliminary work to date have demonstrated that glasses can be successfully prepared from melts of the simulated composition of both lunar and martian soils, and the melts have a viscosity-temperature window appropriate for drawing continuous glass fibers. The glasses are shown to have the potential for immobilizing certain types of nuclear wastes without deteriorating their chemical durability and thermal stability. This has a direct impact on successfully and economically disposing nuclear waste generated from a nuclear power plant on a planetary surface. In addition, these materials display characteristics that can be manipulated using appropriate processing protocols to develop glassy or glass-ceramic magnets. Also discussed in this presentation are other potential applications along with a few selected thermal, chemical, and structural properties as evaluated up to this time for these materials.

  16. 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

  17. Thermoset composite recycling: Properties of recovered glass fiber

    DEFF Research Database (Denmark)

    Beauson, Justine; Fraisse, Anthony; Toncelli, C.

    2015-01-01

    Recycling of glass fiber thermoset polymer composite is a challenging topic and a process able to recover the glass fibers original properties in a limited cost is still under investigation. This paper focuses on the recycling technique separating the glass fiber from the matrix material. Four...

  18. Near net-shape fabrication of hydroxyapatite glass composites

    NARCIS (Netherlands)

    Zhu, Q.; With, de G.; Dortmans, L.; Feenstra, F.

    2004-01-01

    Near net-shape fabrication of hydroxyapatite (HA) glass composites has been attempted by infiltrating a glass into porous HA performs. Main efforts were put to develop glasses that are chemically compatible with HA at elevated temperatures. After extensive investigations in the phosphate and

  19. 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

  20. 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.

  1. 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)

  2. The influence of glass composition on crystalline phase stability in glass-ceramic wasteforms

    International Nuclear Information System (INIS)

    Maddrell, Ewan; Thornber, Stephanie; Hyatt, Neil C.

    2015-01-01

    Highlights: • Crystalline phase formation shown to depend on glass matrix composition. • Zirconolite forms as the sole crystalline phase only for most aluminous glasses. • Thermodynamics indicate that low silica activity glasses stabilise zirconolite. - Abstract: Zirconolite glass-ceramic wasteforms were prepared using a suite of Na 2 O–Al 2 O 3 –B 2 O 3 –SiO 2 glass matrices with variable Al:B ratios. Zirconolite was the dominant crystalline phase only for the most alumina rich glass compositions. As the Al:B ratio decreased zirconolite was replaced by sphene, zircon and rutile. Thermodynamic data were used to calculate a silica activity in the glass melt below which zirconolite is the favoured crystalline phase. The concept of the crystalline reference state of glass melts is then utilised to provide a physical basis for why silica activity varies with the Al:B ratio

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

    International Nuclear Information System (INIS)

    Grambow, B.; Strachan, D.M.

    1988-01-01

    A model selected for the licensing process must be based on a physical and chemical understanding of the glass corrosion mechanism. The purpose of this paper is to show that a dissolution/precipitation model can be used to better understand the effects of various system variables on glass dissolution. The application and validation of this model are also discussed. A dissolution/precipitation model developed appears applicable to experiments with a wide range of solution compositions as well as to more complex systems, such as the bentonite/glass/water system the steel corrosion product/glass/water system, or the dissolution of natural basalt glass in a geologic environment. This model is based on solution chemistry and transition state theory. The theoretical background of this model is discussed elsewhere and is used to describe the dissolution behavior of three nuclear waste glasses. These glasses were selected because they represent a wide range of behavior and, therefore, could be used to illustrate the capabilities of the dissolution/precipitation model. The effects of parameters, such as temperature and starting solution composition, on the dissolution behavior of glass are also discussed. 27 refs., 10 figs., 1 tab

  4. 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.

  5. 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)

  6. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

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

  8. 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.

  9. 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.

  10. The effect of clay on the dissolution of nuclear waste glass

    Science.gov (United States)

    Lemmens, K.

    2001-09-01

    In a nuclear waste repository, the waste glass can interact with metals, backfill materials (if present) and natural host rock. Of the various host rocks considered, clays are often reported to delay the onset of the apparent glass saturation, where the glass dissolution rate becomes very small. This effect is ascribed to the sorption of silica or other glass components on the clay. This can have two consequences: (1) the decrease of the silica concentration in solution increases the driving force for further dissolution of glass silica, and (2) the transfer of relatively insoluble glass components (mainly silica) from the glass surface to the clay makes the alteration layer less protective. In recent literature, the latter explanation has gained credibility. The impact of the environmental materials on the glass surface layers is however not well understood. Although the glass dissolution can initially be enhanced by clay, there are arguments to assume that it will decrease to very low values after a long time. Whether this will indeed be the case, depends on the fate of the released glass components in the clay. If they are sorbed on specific sites, it is likely that saturation of the clay will occur. If however the released glass components are removed by precipitation (growth of pre-existing or new secondary phases), saturation of the clay is less likely, and the process can continue until exhaustion of one of the system components. There are indications that the latter mechanism can occur for varying glass compositions in Boom Clay and FoCa clay. If sorption or precipitation prevents the formation of protective surface layers, the glass dissolution can in principle proceed at a high rate. High silica concentrations are assumed to decrease the dissolution rate (by a solution saturation effect or by the impact on the properties of the glass alteration layer). In glass corrosion tests at high clay concentrations, silica concentrations are, however, often higher

  11. The effect of clay on the dissolution of nuclear waste glass

    International Nuclear Information System (INIS)

    Lemmens, K.

    2001-01-01

    In a nuclear waste repository, the waste glass can interact with metals, backfill materials (if present) and natural host rock. Of the various host rocks considered, clays are often reported to delay the onset of the apparent glass saturation, where the glass dissolution rate becomes very small. This effect is ascribed to the sorption of silica or other glass components on the clay. This can have two consequences: (1) the decrease of the silica concentration in solution increases the driving force for further dissolution of glass silica, and (2) the transfer of relatively insoluble glass components (mainly silica) from the glass surface to the clay makes the alteration layer less protective. In recent literature, the latter explanation has gained credibility. The impact of the environmental materials on the glass surface layers is however not well understood. Although the glass dissolution can initially be enhanced by clay, there are arguments to assume that it will decrease to very low values after a long time. Whether this will indeed be the case, depends on the fate of the released glass components in the clay. If they are sorbed on specific sites, it is likely that saturation of the clay will occur. If however the released glass components are removed by precipitation (growth of pre-existing or new secondary phases), saturation of the clay is less likely, and the process can continue until exhaustion of one of the system components. There are indications that the latter mechanism can occur for varying glass compositions in Boom Clay and FoCa clay. If sorption or precipitation prevents the formation of protective surface layers, the glass dissolution can in principle proceed at a high rate. High silica concentrations are assumed to decrease the dissolution rate (by a solution saturation effect or by the impact on the properties of the glass alteration layer). In glass corrosion tests at high clay concentrations, silica concentrations are, however, often higher

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  13. 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).

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-06-01

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

  16. 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.

  17. 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

  18. 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

  19. Study on Waste Composition at Taman Pura Kencana, Batu Pahat

    Directory of Open Access Journals (Sweden)

    Abdul Kadir Aeslina

    2017-01-01

    Full Text Available Municipal solid waste management is a major challenge due to the increase in population and the development of a country. The problems also arise when the lifespan of the landfills available are shorten than estimated. The aim of this study is to determine waste composition at Taman Pura Kencana. The waste collection was carried out for 50 houses on a daily basis. The collection and sorting out method was conducted based on Malaysian Standard MS 2505:2012 and the data collected is recorded. The result showed the moisture content was approximately ranging between 25% – 30%. The density for each waste has been calculated and the result was recorded accordingly. The highest density is metal followed by organic waste, glass, napkins, household hazardous waste, textiles, paper, plastic rigid, plastic film, rubber and tetrapek. The waste collected also were categorized and consisted of food waste/organic (43.75%, paper (17.97%, plastics rigid (13.58%, plastic film (10.62%, napkins (4.43%, glass (3.10%, household hazardous waste (1.68%, metal (1.67%, tetrapek (1.34% , rubber (0.93%, and textiles (0.92%. Results from the analysis illustrated that waste generation in Taman Pura Kencana may influenced by sudden changes in lifestyles, incomes, household size and also increase of population.

  20. Glass-ceramic composition for hermetic seals

    International Nuclear Information System (INIS)

    Ballard, C.P. Jr.

    1979-01-01

    The invention relates to a glass-ceramic composition having a high fracture strength adaptable for hermetically sealing to chromium bearing iron or nickel base alloys at temperatures of between about 950 0 C to about 1100 0 C to form a hermetically sealed insulator body, comprising from about 55 to about 65 weight percent SiO 2 , from about 0 to about 5 weight percent Al 2 O 3 , from about 6 to about 11 weight % Li 2 O, from about 25 to about 32 weight percent BaO, from about 0.5 to about 1.0 weight percent CoO and from about 1.5 to about 3.5 weight percent P 2 O 5

  1. Glass-ceramic composition for hermetic seals

    Science.gov (United States)

    Ballard, Jr., Clifford P.

    1979-01-01

    The invention relates to a glass-ceramic composition having a high fracture strength adaptable for hermetically sealing to chromium bearing iron or nickel base alloys at temperatures of between about 950.degree. C to about 1100.degree. C to form a hermetically sealed insulator body, comprising from about 55 to about 65 weight percent SiO.sub.2, from about 0 to about 5 weight percent Al.sub.2 O.sub.3, from about 6 to about 11 weight % Li.sub.2 O, from about 25 to about 32 weight percent BaO, from about 0.5 to about 1.0 weight percent CoO and from about 1.5 to about 3.5 weight percent P.sub.2 O.sub.5.

  2. 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

  3. Chemical Composition Analysis and Product Consistency Tests of the ORP Phase 5 Nepheline Study 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); Caldwell, M. E. [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)

    2018-02-01

    In this report, the Savannah River National Laboratory (SRNL) provides chemical analyses and Product Consistency Test (PCT) results for a series of simulated high-level waste glass compositions fabricated by the Pacific Northwest National Laboratory (PNNL). These data will be used in the development of improved models for the prediction of nepheline crystallization in support of the Hanford Tank Waste Treatment and Immobilization Plant (WTP).

  4. 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

  5. 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

  6. 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

  7. 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

  8. 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)

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  10. 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.)

  11. Influence of nuclear glasses composition on their liability to deterioration

    International Nuclear Information System (INIS)

    Tovena, I.

    1995-01-01

    The aim of this thesis is to contribute to the study of the nuclear glasses composition influence on their liability to deterioration. The methodology of the experimental research used has lead to define between the thirty oxides which form the reference glass light water, six oxides of interest. For each of these oxides, a composition variation area has been defined. A matrix of twenty glass compositions has then been defined. The preparation of materials of these compositions has sometimes lead to materials weakly heterogeneous which have been characterized before deterioration. This study has been completed by those of three glasses in a composition variation area narrower of the light water nuclear glass : the R7T7 and two glasses at limits having respectively an initial dissolution velocity at 100 degrees Celsius theoretically maximum and minimum. Some deterioration parameters in pure water have been experimentally measured on the twenty three glasses : 1) an initial dissolution velocity at 100 degrees (Vo 1 00) Celsius and another one at 90 degrees Celsius (Vo 9 0) 2) a dissolution velocity in conditions near the saturation at 90 degrees Celsius 3) an apparent solubility of glass based on the ortho silicic acid activity 4) the evolution of the dissolution kinetics at 90 degrees Celsius in sub-saturated medium towards saturated medium 5) the alteration films nature developed at the glasses surface during these last alteration tests. Some thermodynamic and structural models have been studied in order to predict Vo 9 0 and Vo 1 00. The dissolution kinetic law developed from reference glass dissolution results has been studied with the calculation code LIXIVER. It has not been able to be used for most of the glasses compositions studied. As a consequence, the glasses dissolution control by a surface reaction which are itself controlled by the only dissolved silica is an hypothesis which is not verified for the greater part of the glasses. (O.L.). refs., figs

  12. 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

  13. 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.)

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

    International Nuclear Information System (INIS)

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

    1982-02-01

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

  15. Waste glass corrosion modeling: Comparison with experimental results

    International Nuclear Information System (INIS)

    Bourcier, W.L.

    1993-11-01

    A chemical model of glass corrosion will be used to predict the rates of release of radionuclides from borosilicate glass waste forms in high-level waste repositories. The model will be used both to calculate the rate of degradation of the glass, and also to predict the effects of chemical interactions between the glass and repository materials such as spent fuel, canister and container materials, backfill, cements, grouts, and others. Coupling between the degradation processes affecting all these materials is expected. Models for borosilicate glass dissolution must account for the processes of (1) kinetically-controlled network dissolution, (2) precipitation of secondary phases, (3) ion exchange, (4) rate-limiting diffusive transport of silica through a hydrous surface reaction layer, and (5) specific glass surface interactions with dissolved cations and anions. Current long-term corrosion models for borosilicate glass employ a rate equation consistent with transition state theory embodied in a geochemical reaction-path modeling program that calculates aqueous phase speciation and mineral precipitation/dissolution. These models are currently under development. Future experimental and modeling work to better quantify the rate-controlling processes and validate these models are necessary before the models can be used in repository performance assessment calculations

  16. Modeling the dissolution behavior of defense waste glass in a salt repository environment

    International Nuclear Information System (INIS)

    McGrain, B.P.

    1988-02-01

    A mechanistic model describing a dynamic mass balance between the production and consumption of dissolved silica was found to describe the dissolution behavior of SRL-165 defense waste glass in a high-magnesium brine (PBB3) at a temperature of 90 0 C. The synergistic effect of the waste package container on the glass dissolution rate was found to depend on a precipitation reaction for a ferrous silicate mineral. The model predicted that the ferrous silicate precipitate should be variable in composition where the iron/silica stoichiometry depended on the metal/glass surface area ratio used in the experiment. This prediction was confirmed experimentally by the variable iron/silica ratios observed in filtered leachates. However, the interaction between dissolved silica and iron corrosion products needs to be much better understood before the model can be used with confidence in predicting radionuclide release rates for a salt repository. 25 refs., 4 figs., 1 tab

  17. 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

  18. 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.

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  20. 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

  1. 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

  2. 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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Ramsey, W.G.

    1993-01-01

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

  5. 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

  6. 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

  7. 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

  8. 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.)

  9. Composition of municipal solid waste in Denmark

    DEFF Research Database (Denmark)

    Edjabou, Vincent Maklawe Essonanawe; Petersen, Claus; Scheutz, Charlotte

    2014-01-01

    Data for the composition of municipal solid waste is a critical basis for any assessment of waste technologies and waste management systems. The detailed quantification of waste fractions is absolutely needed for a better technological development of waste treatment. The current waste composition...... comparability to characterize municipal solid waste. This methodology was applied to residual waste collected from 1,442 households in three municipalities in Denmark. The main fractions contributing to the residual household waste were food waste and miscellaneous waste. Statistical analysis suggested...... of standardised and commonly accepted waste characterization methodologies, various approaches have been reported in literature. This limits both comparability and applicability of the results. The purpose of this study was to introduce a consistent methodology that reduces uncertainties and ensures data...

  10. A general model for the dissolution of nuclear waste glasses in salt brine

    International Nuclear Information System (INIS)

    McGrail, B.P.; Strachan, D.M.

    1988-07-01

    A mechanistic model describing a dynamic mass balance between the production and consumption of dissolved silica was found to describe the dissolution of SRL-165 defense waste glass in a high-magnesium (PBB3) brine at a temperature of 90/degree/C. The synergetic effect of the waste package container on the glass dissolution rate was found to depend on a precipitation reaction for a ferrous silicate mineral. The model predicted that the ferrous silicate precipitate should be variable in composition where the iron-silica ratio depended on the metal-to-glass surface area ratio used in the experiment. This prediction was confirmed experimentally by the variable iron-silica ratios observed in filtered leachates. However, the interaction between dissolved silica and iron corrosion products needs to be much better understood before the model could be used with confidence in predicting radionuclide release rates for a salt repository. If the deleterious effects of the iron corrosion products can be shown to be transient, and the fracturing of the glass can be minimized, it appears that the performance of SRL-165 defense waste glass will be near the NRC regulatory criterion for fraction release of one part in 100,000 in PBB3 brine at 90/degree/C under silica-saturated conditions. 47 refs., 6 figs., 1 tab

  11. 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.

  12. 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

  13. 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

  14. 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)

  15. 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.

  16. Glass/Ceramic Composites for Sealing Solid Oxide Fuel Cells

    Science.gov (United States)

    Bansal, Narottam P.; Choi, Sung R.

    2007-01-01

    A family of glass/ceramic composite materials has been investigated for use as sealants in planar solid oxide fuel cells. These materials are modified versions of a barium calcium aluminosilicate glass developed previously for the same purpose. The composition of the glass in mole percentages is 35BaO + 15CaO + 5Al2O3 + 10B2O3 + 35SiO2. The glass seal was found to be susceptible to cracking during thermal cycling of the fuel cells. The goal in formulating the glass/ ceramic composite materials was to (1) retain the physical and chemical advantages that led to the prior selection of the barium calcium aluminosilicate glass as the sealant while (2) increasing strength and fracture toughness so as to reduce the tendency toward cracking. Each of the composite formulations consists of the glass plus either of two ceramic reinforcements in a proportion between 0 and 30 mole percent. One of the ceramic reinforcements consists of alumina platelets; the other one consists of particles of yttria-stabilized zirconia wherein the yttria content is 3 mole percent (3YSZ). In preparation for experiments, panels of the glass/ceramic composites were hot-pressed and machined into test bars.

  17. Viscosity calculations of simulated ion-exchange resin waste glasses

    International Nuclear Information System (INIS)

    Kim, Cheon Woo; Park, Jong Kil; Lee, Kyung Ho; Lee, Myung Chan; Song, Myung Jae; BRUNELOT, Pierre

    2000-01-01

    An induction cold crucible melter (CCM) located in the NETEC-KEPCO has been used to vitrify simulated ion-exchange resin. During vitrification, the CCM operations were tightly constrained by glass viscosity as an important process parameter. Understanding the role of viscosity and quantifying viscosity is highly required in the determination of optimized feed formulations and in the selection of the processing temperature. Therefore, existing process models of glass viscosity based on a relationship between the glass composition, its structure polymerization, and the temperature were searched and adapted to our borosilicate glass systems. Calculated data using a viscosity model based on calculation of non-bridging oxygen (NBO) were in good agreement with the measured viscosity data of benchmark glasses

  18. 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

  19. Development of soda-lime glasses from ornamental rock wastes

    International Nuclear Information System (INIS)

    Babisk, Michelle Pereira

    2009-01-01

    During the ornamental rocks production, among other steps, one saw the rock blocks in order to transform them into semi-finished plates. In this step, expressive amounts of residues are generated, which are not properly discharged in nature, without any programmed utilization. The residues of silicide rocks present, in their compositions, oxides which are raw materials employed to fabricate soda-lime type glasses (containing SiO_2, Al_2O_3, CaO, Na_2O and K_2O). On the other hand the residues of carbonatic rocks are constituted of glass net modifier oxides, like CaO and MgO. In this work it was developed four types of soda-lime glasses using ornamental rock residues, where the glasses compositions were adjusted by adding sand, as silica source, as well as sodium and calcium carbonates as sources of Na_2O and CaO, respectively. The obtained glasses were characterized by means of Archimed's method for densities measurements, microstructure by using optical and electronic microscopy, phases by means of X-ray diffraction (XRD), hardness by Vickers indentation, spectroscopy (UV/VIS), and hydrolytic resistance according to ISO 719. The XRD analyses confirmed the compositions total vitrification, where the greened aspect of the samples was due to the presence of the iron oxides. The produced glasses properties were compared with those of commercial glasses aiming their industrial employment. The main difference between the produced glasses and those commercials varied primarily regarding the amount of carbonates incorporated. The results showed that the ornamental rocks residues may be used as raw materials for glasses fabrication, and they found a useful economic destination rather than discharge which promotes undesirable environmental impact. (author)

  20. Exploring high-strength glass-ceramic materials for upcycling of industrial wastes

    Science.gov (United States)

    Back, Gu-Seul; Park, Hyun Seo; Seo, Sung Mo; Jung, Woo-Gwang

    2015-11-01

    To promote the recycling of industrial waste and to develop value-added products using these resources, the possibility of manufacturing glass-ceramic materials of SiO2-CaO-Al2O3 system has been investigated by various heat treatment processes. Glass-ceramic materials with six different chemical compositions were prepared using steel industry slags and power plant waste by melting, casting and heat treatment. The X-ray diffraction results indicated that diopside and anorthite were the primary phases in the samples. The anorthite phase was formed in SiO2-rich material (at least 43 wt%). In CaO-rich material, the gehlenite phase was formed. By the differential scanning calorimetry analyses, it was found that the glass transition point was in the range of 973-1023 K, and the crystallization temperature was in the range of 1123-1223 K. The crystallization temperature increased as the content of Fe2O3 decreased. By the multi-step heat treatment process, the formation of the anorthite phase was enhanced. Using FactSage, the ratio of various phases was calculated as a function of temperature. The viscosities and the latent heats for the samples with various compositions were also calculated by FactSage. The optimal compositions for glass-ceramics materials were discussed in terms of their compressive strength, and micro-hardness.

  1. 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

  2. Investigation of sizing - from glass fibre surface to composite interface

    DEFF Research Database (Denmark)

    Petersen, Helga Nørgaard; Kusano, Yukihiro

    significantly. The usage span wide, from furniture and car components to construction materials. Even though, the concept of composites is well known and widely applied, the fundamental principles of the interaction of the constituents, in the composites are still not fully understood. This thesis is a part...... of the sizing from the glass fibre surface to the interface in composites. Through soxhlet extraction with acetone it was possible to remove a part of the sizing from the glass fibres for analysis. By burning off the sizing at 565 ºC a higher mass loss was obtained than from the extraction, indicating...... increased after the removal of sizing by extraction but also when the sizing was removed by burning. This could partly be explained by the sizing being less dense than the glass fibres. For the burned glass fibres compactment of the glass structure also yields an increase in stiffness. The fibre strength...

  3. 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

  4. Application of the final flotation waste for obtaining the glass-ceramic materials

    Directory of Open Access Journals (Sweden)

    Cocić Mira

    2017-01-01

    Full Text Available This work describes the investigation of the final flotation waste (FFW, originating from the RTB Bor Company (Serbia, as the main component for the production of glass-ceramic materials. The glass-ceramics was synthesized by the sintering of FFW, mixtures of FFW with basalt (10%, 20%, and 40%, and mixtures of FFW with tuff (20% and 40%. The sintering was conducted at the different temperatures and with the different time duration in order to find the optimal composition and conditions for crystallization. The increase of temperature, from 1100 to 1480°C, and sintering time, from 4 to 6h resulted in a higher content of hematite crystal in the obtained glass-ceramic (up to 44%. The glass-ceramics sintered from pure FFW (1080°C/36h has good mechanical properties, such as high propagation speed (4500 m/s and hardness (10800 MPa, as well as very good thermal stability. The glass-ceramics obtained from mixtures shows weaker mechanical properties compared to that obtained from pure FFW. The mixtures of FFW with tuff have a significantly lower bulk density compared to other obtained glass-ceramics. Our results indicate that FFW can be applied as a basis for obtaining the construction materials. [Project of the Serbian Ministry of Education, Science and Technological Development, Grant no. 176010: Composition, genesis, application, and contribution to the environmental sustainability

  5. 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

  6. 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

  7. Near net-shape fabrication of alumina glass composites

    NARCIS (Netherlands)

    Zhu, Q.; With, de G.; Dortmans, L.J.M.G.; Feenstra, F.

    2005-01-01

    The purpose of the present study is to fabricate alumina glass composites by melt infiltration with better dimensional control through reducing both the presintering and infiltration temperature. Main efforts were put to develop glasses that are chemically compatible with alumina. After extensive

  8. Mechanical, dielectric and optical assessment of glass composites ...

    Indian Academy of Sciences (India)

    The band gap, Urbach energy and the extinction coefficient of the glass composites have been ... icate glasses during our previous studies.7 Lanthanum oxide. (La2O3) .... vs. incident photon energy hν are used to determine the type of optical ...

  9. 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

  10. 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

  11. In vitro behaviour of three biocompatible glasses in composite implants.

    Science.gov (United States)

    Varila, Leena; Lehtonen, Timo; Tuominen, Jukka; Hupa, Mikko; Hupa, Leena

    2012-10-01

    Poly(L,DL-lactide) composites containing filler particles of bioactive glasses 45S5 and S53P4 were compared with a composite containing a slowly dissolving glass S68. The in vitro reactivity of the composites was studied in simulated body fluid, Tris-buffered solution, and phosphate buffered saline. The high processing temperature induced thermal degradation giving cavities in the composites containing 45S5 and S53P4, while good adhesion of S68 to the polymer was observed. The cavities partly affected the in vitro reactivity of the composites. The degradation of the composites containing the bioactive glasses was faster in phosphate buffered saline than in the two other solutions. Hydroxyapatite precipitation suggesting bone tissue bonding capability was observed on these two composites in all three solutions. The slower dissolution of S68 glass particles and the limited hydroxyapatite precipitation suggested that this glass has potential as a reinforcing composition with the capability to guide bone tissue growth in biodegradable polymer composites.

  12. Composition of municipal solid waste in Denmark

    DEFF Research Database (Denmark)

    Edjabou, Maklawe Essonanawe

    In response to continuous pressure on resources, and the requirement for secure and sustainable consumption, public authorities are pushing the efficient use of resources. Among other initiatives, the prevention, reduction and recycling of solid waste have been promoted. In this context, reliable...... data for the material and resource content of waste flows are crucial to establishing baselines, setting targets and tracking progress on waste prevention, reduction and recycling goals. Waste data are also a critical basis for the planning, development and environmental assessment of technologies...... the comparison of waste data with various objectives. Analysis revealed that Danish residual household waste constitutes mainly food waste (42 – 45% mass per wet basis). Misplaced recyclable materials in residual waste bins, such as paper, board, glass, metal and plastic, amounted to 20% (mass per wet basis...

  13. 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

  14. Ceramic fiber reinforced glass-ceramic matrix composite

    Science.gov (United States)

    Bansal, Narottam P. (Inventor)

    1993-01-01

    A slurry of BSAS glass powders is cast into tapes which are cut to predetermined sizes. Mats of continuous chemical vapor deposition (CVD)-SiC fibers are alternately stacked with these matrix tapes. This tape-mat stack is warm-pressed to produce a 'green' composite which is heated to burn out organic constituents. The remaining interim material is then hot-pressed to form a BSAS glass-ceramic fiber-reinforced composite.

  15. 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

  16. 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

  17. Fiber glass-bioactive glass composite for bone replacing and bone anchoring implants.

    Science.gov (United States)

    Vallittu, Pekka K; Närhi, Timo O; Hupa, Leena

    2015-04-01

    Although metal implants have successfully been used for decades, devices made out of metals do not meet all clinical requirements, for example, metal objects may interfere with some new medical imaging systems, while their stiffness also differs from natural bone and may cause stress-shielding and over-loading of bone. Peer-review articles and other scientific literature were reviewed for providing up-dated information how fiber-reinforced composites and bioactive glass can be utilized in implantology. There has been a lot of development in the field of composite material research, which has focused to a large extent on biodegradable composites. However, it has become evident that biostable composites may also have several clinical benefits. Fiber reinforced composites containing bioactive glasses are relatively new types of biomaterials in the field of implantology. Biostable glass fibers are responsible for the load-bearing capacity of the implant, while the dissolution of the bioactive glass particles supports bone bonding and provides antimicrobial properties for the implant. These kinds of combination materials have been used clinically in cranioplasty implants and they have been investigated also as oral and orthopedic implants. The present knowledge suggests that by combining glass fiber-reinforced composite with particles of bioactive glass can be used in cranial implants and that the combination of materials may have potential use also as other types of bone replacing and repairing implants. Copyright © 2015 Academy of Dental Materials. Published by Elsevier Ltd. All rights reserved.

  18. 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)

  19. 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

  20. Elemental compositions of suspended particles released in glass manufacture

    Energy Technology Data Exchange (ETDEWEB)

    Mamuro, T; Mizohata, A; Kubota, T [Radiation Center of Osaka Prefecture, Sakai (Japan)

    1980-03-01

    Suspended particles released in glass manufacture were subjected to multielement analysis by means of instrumental neutron activation method and energy dispersive X-ray fluorescence spectrometry. Suspended particles emitted from glass manufacture generally consist of both particles emitted from glass fusion and those produced through fuel combustion (mainly oil combustion). Elemental compositions of suspended particles emitted from glass fusion were found to be strongly dependent on the kind and recipe of raw materials and additives. Of the various metallic elements involved in suspended particles emitted from glass fusion, the elements, As, Se, Cd, Sb, Pb and so on are regarded to produce the most serious air pollution. The amount of emission of these elements to the environment is, howerer, quite varied from manufacturer to manufacturer. The replacement of electric furnace by oil combustion in opal glass manufacture remarkably reduced the emission of metallic elements to the environment.

  1. Composition-Structure-Property Relationships in Boroaluminosilicate Glasses

    DEFF Research Database (Denmark)

    Zheng, Qiuju; Potuzak, M.; Mauro, J.C.

    2012-01-01

    boroaluminosilicate glasses from peralkaline to peraluminous compositions by substituting Al2O3 for SiO2. Our results reveal a pronounced change in all the measured physical properties (density, elastic moduli, hardness, glass transition temperature, and liquid fragility) around [Al2O3]–[Na2O]=0. The structural......The complicated structural speciation in boroaluminosilicate glasses leads to a mixed network former effect yielding nonlinear variation in many macroscopic properties as a function of chemical composition. Here we study the composition–structure–property relationships in a series of sodium...

  2. 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.

  3. GLASS FABRICATION AND PRODUCT CONSISTENCY TESTING OF LANTHANIDE BOROSILICATE FRIT X COMPOSITION FOR PLUTONIUM DISPOSITION

    Energy Technology Data Exchange (ETDEWEB)

    Marra, J

    2006-11-15

    The Department of Energy Office of Environmental Management (DOE/EM) plans to conduct the Plutonium Disposition Project at the Savannah River Site (SRS) to disposition excess weapons-usable plutonium. A plutonium glass waste form is the preferred option for immobilization of the plutonium for subsequent disposition in a geologic repository. A reference glass composition (Lanthanide Borosilicate (LaBS) Frit B) was developed during the Plutonium Immobilization Program (PIP) to immobilize plutonium in the late 1990's. A limited amount of performance testing was performed on this baseline composition before efforts to further pursue Pu disposition via a glass waste form ceased. Recent FY05 studies have further investigated the LaBS Frit B formulation as well as development of a newer LaBS formulation denoted as LaBS Frit X. The objectives of this present task were to fabricate plutonium loaded LaBS Frit X glass and perform corrosion testing to provide near-term data that will increase confidence that LaBS glass product is suitable for disposal in the Yucca Mountain Repository. Specifically, testing was conducted in an effort to provide data to Yucca Mountain Project (YMP) personnel for use in performance assessment calculations. Plutonium containing LaBS glass with the Frit X composition with a 9.5 wt% PuO{sub 2} loading was prepared for testing. Glass was prepared to support Product Consistency Testing (PCT) at Savannah River National Laboratory (SRNL). The glass was thoroughly characterized using x-ray diffraction (XRD) and scanning electron microscopy coupled with energy dispersive spectroscopy (SEM/EDS) prior to performance testing. A series of PCTs were conducted at SRNL using quenched Pu Frit X glass with varying exposed surface areas. Effects of isothermal and can-in-canister heat treatments on the Pu Frit X glass were also investigated. Another series of PCTs were performed on these different heat-treated Pu Frit X glasses. Leachates from all these PCTs

  4. 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

  5. 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.

  6. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-24

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

  8. 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

  9. IMPACTS OF SMALL COLUMN ION EXCHANGE STREAMS ON DWPF GLASS FORMULATION: KT01, KT02, KT03, AND KT04-SERIES GLASS COMPOSITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K.; Edwards, T.

    2010-11-01

    Four series of glass compositions were selected, fabricated, and characterized as part of a study to determine the impacts of the addition of Crystalline Silicotitanate (CST) and Monosodium Titanate (MST) from the Small Column Ion Exchange (SCIX) process on the Defense Waste Processing Facility (DWPF) glass waste form and the applicability of the DWPF process control models. The KT01 and KT02-series of glasses were chosen to allow for the identification of the influence of the concentrations of major components of the glass on the retention of TiO{sub 2}. The KT03 series of glasses was chosen to allow for the identification of these influences when higher Nb{sub 2}O{sub 5} and ZrO{sub 2} concentrations are included along with TiO2. The KT04 series of glasses was chosen to investigate the properties and performance of glasses based on the best available projections of actual compositions to be processed at the DWPF (i.e., future sludge batches including the SCIX streams).

  10. 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

  11. Development of Composite Materials Under Ecological Aspects as Recycling Concept For Borosilicate Glass Containing Iron Slags

    International Nuclear Information System (INIS)

    Khalil, T.K.; Bossert, J.; Aly, H.F.; Bossert, J.

    1999-01-01

    Composite concept in materials science can be conveniently applied in the waste treatment technology to construct specific t ailor made c omposite materials, in which at least one of the phases is built by the waste material. In this work the applicability of this concept for the fixation and recycling of slags wastes is done, whereby different mixtures of blast furnace slags are mixed with two different borosilicate glasses, which serve as matrix material. Thermal behaviour of the produced compacts were studied. Both melting and powder technology are applied for the fabrication of dense products. The microstructure of sintered samples is investigated by electron microscopy. The mechanical properties such as hardness and fracture toughness are determined by a Vickers technique. An improvement of the fracture toughness of more than 50% over the value for the original glass VG 98 is achieved by slag addition

  12. 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

  13. 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

  14. 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)

  15. 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

  16. 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

  17. 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.

  18. Doping influence by some transition elements on the irradiation effects in nuclear waste glasses

    International Nuclear Information System (INIS)

    Florent, Olivier

    2006-06-01

    High-level waste glasses are submitted to auto-irradiation. Modelling it using external irradiations on simple glasses revealed defects production and non negligible structural changes. This thesis aims at determining the impact of a more complex composition on these effects, especially the influence of adding polyvalent transition metals. Silicate, soda-lime and alumino-borosilicate glasses are doped with different iron, chromium and manganese concentrations then β irradiated at different doses up to 10 9 Gy. Non doped glasses show an increase of their density and polymerisation coupled with a molecular oxygen and point defects production. Adding 0.16 mol% Fe decreases the amount of defects by 85 % and all irradiation effects. A Fe 3+ reduction is also observed by EPR, optical absorption and indirectly by Raman spectroscopy. A higher than 0.32 mol% Fe concentration causes complete blockage of the evolution of polymerisation, density and defect production. The same results are obtained on chromium or manganese doped glasses. An original in situ optical absorption device shows the quick decrease of Fe 3+ amount to a 25 % lower level during irradiation. Stopping irradiation causes a lower decrease of 65 %, suggesting a dynamic (h 0 /e-) consuming equilibrium. He + and Kr 3+ ions and γ irradiated glasses tend to confirm these phenomena for all kind of irradiation with electronic excitations. (author)

  19. 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)

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

    Science.gov (United States)

    Christiansen, Mary U.

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

  1. 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.

  2. 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.

  3. Summary report on microstructure and composition of silicate melts containing simulated Hanford waste

    International Nuclear Information System (INIS)

    Daniel, J.L.

    1975-04-01

    Specimens of silicate melt es containing simulated Hanford waste were studied by microscopy and microprobe methods to determine microstructural and compositional characteristics. The two glass specimens were representative of glasses prepared with Hanford basalt and with sea sand as the source of SiO 2 . Samples of both glasses were studied in detail at locations near the top, bottom, center, and sides of the melts. Both glasses were of a highly uniform microstructure and composition. The basalt glass contained metallic iron inclusions around the periphery near the glass/crucible interface, and small increases in Si content adjacent to the pores occurring throughout the glass. The sand glass contained no iron, its Si composition was uniform, and the average pore size was somewhat smaller (about 2 μm) than in the basalt glass. The Ca nominally added to the sand glass could not be detected. Both glasses contained a random scattering of a micron-sized ''bright'' phase whose composition was identical to the matrix or containing elements not detectable by microprobe methods. (U.S.)

  4. 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)

  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. 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

  7. GLASS FABRICATION AND PRODUCT CONSISTENCY TESTING OF LANTHANIDE BOROSHILICATE FRIT X COMPOSITION FOR PLUTONIUM DISPOSITION

    Energy Technology Data Exchange (ETDEWEB)

    Marra, J

    2006-11-21

    The Department of Energy Office of Environmental Management (DOE/EM) plans to conduct the Plutonium Disposition Project at the Savannah River Site (SRS) to disposition excess weapons-usable plutonium. A plutonium glass waste form is the preferred option for immobilization of the plutonium for subsequent disposition in a geologic repository. A reference glass composition (Lanthanide Borosilicate (LaBS) Frit B) was developed during the Plutonium Immobilization Program (PIP) to immobilize plutonium in the late 1990's. A limited amount of performance testing was performed on this baseline composition before efforts to further pursue Pu disposition via a glass waste form ceased. Recent FY05 studies have further investigated the LaBS Frit B formulation as well as development of a newer LaBS formulation denoted as LaBS Frit X. The objectives of this present task were to fabricate plutonium loaded LaBS Frit X glass and perform corrosion testing to provide near-term data that will increase confidence that LaBS glass product is suitable for disposal in the Yucca Mountain Repository. Specifically, testing was conducted in an effort to provide data to Yucca Mountain Project (YMP) personnel for use in performance assessment calculations. Plutonium containing LaBS glass with the Frit X composition with a 9.5 wt% PuO{sub 2} loading was prepared for testing. Glass was prepared to support Product Consistency Testing (PCT) at Savannah River National Laboratory (SRNL). The glass was thoroughly characterized using x-ray diffraction (XRD) and scanning electron microscopy coupled with energy dispersive spectroscopy (SEM/EDS) prior to performance testing. A series of PCTs were conducted at SRNL using quenched Pu Frit X glass with varying exposed surface areas. Effects of isothermal and can-in-canister heat treatments on the Pu Frit X glass were also investigated. Another series of PCTs were performed on these different heat-treated Pu Frit X glasses. Leachates from all these PCTs

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

    International Nuclear Information System (INIS)

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

    2015-01-01

    Highlights: • Spinel crystallization incorporates ZnO from base glass, displacing Mg and Ni. • Raman spectroscopy demonstrates significant impact on glass structure by addition of ZnO to base glass. • Addition of ZnO reduces glass dissolution rate at early time periods (up to 28 days). - Abstract: Compositional modification of United Kingdom high level nuclear waste (HLW) glasses was investigated with the aim of understanding the impact of adopting a ZnO/CaO modified base glass on the vitrified product phase assemblage, glass structure, processing characteristics and dissolution kinetics. Crystalline spinel phases were identified in the vitrified products derived from the Na 2 O/Li 2 O and the ZnO/CaO modified base glass compositions; the volume fraction of the spinel crystallites increased with increasing waste loading from 15 to 20 wt%. The spinel composition was influenced by the base glass components; in the vitrified product obtained with the ZnO/CaO modified base glass, the spinel phase contained a greater proportion of Zn, with a nominal composition of (Zn 0.60 Ni 0.20 Mg 0.20 )(Cr 1.37 Fe 0.63 )O 4 . The addition of ZnO and CaO to the base glass was also found to significantly alter the glass structure, with changes identified in both borate and silicate glass networks using Raman spectroscopy. In particular, these glasses were characterised by a significantly higher Q 3 species, which we attribute to Si–O–Zn linkages; addition of ZnO and CaO to the glass composition therefore enhanced glass network polymerisation. The increase in network polymerisation, and the presence of spinel crystallites, were found to increase the glass viscosity of the ZnO/CaO modified base glass; however, the viscosities were within the accepted range for nuclear waste glass processing. The ZnO/CaO modified glass compositions were observed to be significantly more durable than the Na 2 O/Li 2 O base glass up to 28 days, due to a combination of the enhanced network

  9. 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...

  10. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-01

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

  12. 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.

  13. 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

  14. Boron nitride nanosheets reinforced glass matrix composites

    Czech Academy of Sciences Publication Activity Database

    Saggar, Richa; Porwal, H.; Tatarko, P.; Dlouhý, Ivo; Reece, M. J.

    2015-01-01

    Roč. 114, SEP (2015), S26-S32 ISSN 1743-6753 R&D Projects: GA MŠk(CZ) 7AMB14SK155 EU Projects: European Commission(XE) 264526 Institutional support: RVO:68081723 Keywords : Boron nitride nanosheets * Borosilicate glass * Mechanical properties Subject RIV: JL - Materials Fatigue, Friction Mechanics Impact factor: 1.162, year: 2015

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-12-30

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

  16. Production and characterization of red mud based on glasses for the immobilization of nuclear wastes

    International Nuclear Information System (INIS)

    Vieira, Heveline

    2015-01-01

    Glasses based on red mud, a residual material from bauxite processing, were developed and characterized in this work. In order to promote its use, a minimum 60 wt% of red mud was used in the production of the glasses. According to XRD results, materials containing considerable amorphous phases were produced when using red mud as raw material. These amorphous phases were observed even though crystalline phases associated to Fe coming from the red mud itself were present. The material denominated 60L40S, which has a nominal composition of 60 wt% red mud showed the best properties comparing with the others compositions studied. However, these materials presented a high melting temperature. Changes in the composition of this material were made with the objective of lowering this temperature. Results indicated that the changes made to the material were successful in the reduction of the melting temperature. However, a reduction in the chemical properties of the resulting material was observed. Elements usually found in the chemical composition of nuclear wastes were added to the glasses produced. It was done with the objective of determining the effect of these elements on the chemical and physical properties of the red mud based glasses obtained. It was found that it was possible to add up to 15 wt% of these elements to the materials produced. The addition of these simulant materials promoted a reduction in the melting temperature of the resulting material. Above 15 wt%, the added elements precipitate in the structure of the resulting material. Even though the reduction in the chemical durability of the 60L40S material when simulant elements were added, it was observed that this material contained the simulant elements confined in its structure when in contact with water. This is a promising result, since it indicates that the 60L40S has the potential to immobilize elements from nuclear wastes . (author)

  17. Multifunctionality in epoxy/glass fibers composites with graphene interphase

    OpenAIRE

    Mahmood, Haroon

    2017-01-01

    In this project, the synergetic effect of a graphene interphase in epoxy/glass fibers composites was investigated by coating glass fibers (GF) with graphene oxide (GO) and reduced graphene oxide (rGO) nanosheets by an electrophoretic deposition (EPD) technique. Graphite oxide was prepared using modified Hummers method in which raw graphite powder was oxidized using potassium permanganate (KMnO4) in acidic solution. Using ultrasonic technique, the graphite oxide was dispersed homogenously in w...

  18. Mechanical Characterization of Basalt and Glass Fiber Epoxy Composite Tube

    OpenAIRE

    Lapena, Mauro Henrique; Marinucci, Gerson

    2017-01-01

    The application of basalt fibers are possible in many areas thanks to its multiple and good properties. It exhibits excellent resistance to alkalis, similar to glass fiber, at a much lower cost than carbon and aramid fibers. In the present paper, a comparative study on mechanical properties of basalt and E-glass fiber composites was performed. Results of apparent hoop tensile strength test of ring specimens cut from tubes and the interlaminar shear stress (ILSS) test are presented. Tensile te...

  19. 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

  20. 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

  1. Composition-structure-property relation of oxide glasses

    DEFF Research Database (Denmark)

    Hermansen, Christian

    also increases such properties. Yet, these rules are not strictly followed even for the simplest binary oxide glasses, such as alkali silicates, borates and phosphates. In this thesis it is argued that the missing link between composition and properties is the glass structure. Structural models...... are proposed based on topological selection rules and experimentally verified. The relation between structure and properties is evaluated using topological constraint theory, which in its essence is a theory that quantifies the two intuitions of the glass scientist. The end result is a quantitative model...

  2. In situ one-year burial experiments with simulated nuclear waste glasses

    International Nuclear Information System (INIS)

    Hench, L.L.; Spilman, D.; Buonaquisti, T.; Werme, L.

    1985-01-01

    Two simulated nuclear waste glasses were corroded in an in-situ experiment in the Stripa mine up to one year at 90 degree C and ambient temperature. Changes in compositional in-depth profiles were measured using Fourier transform infrared reflection spectroscopy, SIMS and Rutherford back-scattering. For glass/glass interfaces, both glasses showed depletion of Na, Cs and B, but for the more corrosion resistant glass, the lower depletion is ascribed to the formation of a thin (0.2 nm) coherent and dense outer layer enriched in Mg, Ca, Sr, Ba, Zn-Al and Si, which impedes both ion exchange and network attack of the bulk underneath. For the bentonite interfaces, cation exchange of Ca, Mg, Al and Fe from the bentonite for primarily Na and B is found to produce a glass surface that has three silicate-rich layers. The larger concentrations of M/super2+/ and M/super3+/ cation and the high silica content of the reaction layers result in a considerably retarded rate of ion exchange after the formation of these layers during the first three months of burial. The granite interfaces showed the lowest rate of attack. This appears to be due to a large increase of Fe and Al within the glass surfaces exposed to granite. The results obtained using Rutheford back-scattering confirm the results obtained using the other techniques for surface analysis. Analysis of burial samples cast in steel mini-canisters show no significant effects associated with the steel canister-glass interface. (author)

  3. Viscosity of glasses containing simulated Savannah River Plant waste

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1978-08-01

    The viscosity of glass melts containing four simulated sludge types and two frit candidates (Frits 18 and 21) was measured over the temperature range 750 to 1200 0 C. The viscosity of melts made with either frit was reduced by the addition of high iron sludge, unchanged by average sludge, and increased by composite and high aluminum sludge. High aluminium sludge greatly increased the viscosity. Frit 21 (containing 4 wt % Li 2 O substituted for 4 wt % Na 2 O in Frit 18) was clearly better than Frit 18 in terms of its low viscosity. However, further reductions in viscosity are desirable, especially for glasses containing high aluminum sludge. Changing any frit component by 1 wt % did not significantly affect the viscosity of the glasses. Therefore, variability of 1 wt % in any frit component can be tolerated

  4. Alteration of basaltic glass in Iceland as a natural analogue for nuclear waste glasses

    International Nuclear Information System (INIS)

    Crovisier, J.L.; Advocat, T.; Fritz, B.; Petit, J.C.

    1989-01-01

    The authors describe the longer term geochemical consequences of basaltic glass dissolution in fresh water at 0 0 C calculated with the computer code DISSOL. The clay minerals were represented by an ideal solid solution model (CISSFIT) able to describe variations in chemical composition of a clay phase in response to variations of the solution chemistry. The predicted mineral phases were iron hydroxides followed by kaolinite, TOT clays, chabazite and clinoptilolite. These results are in reasonably good agreement with experimental results and observations of altered subglacial hyaloclastites from Iceland. The formation of secondary products are mainly controlled by thermodynamic constraints. Kinetic effects, such as diffusion in the near glass surface are not important

  5. Chemical composition analysis and product consistency tests supporting refinement of the Nepheline Model for the high aluminum Hanford glass composition region

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

    2016-03-01

    In this report, Savannah River National Laboratory provides chemical analyses and Product Consistency Test (PCT) results for a series of simulated high level waste (HLW) glasses fabricated by Pacific Northwest National Laboratory (PNNL) as part of an ongoing nepheline crystallization study. The results of these analyses will be used to improve the ability to predict crystallization of nepheline as a function of composition and heat treatment for glasses formulated at high alumina concentrations.

  6. Leach rate studies on glass containing actual radioactive