High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 2

International Nuclear Information System (INIS)

Cunnane, J.C.

1994-03-01

The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion.This document is organized into three volumes. Volumes I and II represent a tiered set of information intended for somewhat different audiences. Volume I is intended to provide an overview of waste glass corrosion, and Volume 11 is intended to provide additional experimental details on experimental factors that influence waste glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II. Volume I is intended for managers, decision makers, and modelers, the combined set of Volumes I, II, and III is intended for scientists and engineers working in the field of high-level waste

High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 2

Energy Technology Data Exchange (ETDEWEB)

Cunnane, J.C. [comp.; Bates, J.K.; Bradley, C.R. [Argonne National Lab., IL (United States)] [and others

1994-03-01

The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion.This document is organized into three volumes. Volumes I and II represent a tiered set of information intended for somewhat different audiences. Volume I is intended to provide an overview of waste glass corrosion, and Volume 11 is intended to provide additional experimental details on experimental factors that influence waste glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II. Volume I is intended for managers, decision makers, and modelers, the combined set of Volumes I, II, and III is intended for scientists and engineers working in the field of high-level waste.

The borosilicate glass for 'PAMELA'

International Nuclear Information System (INIS)

Schiewer, E.

1986-01-01

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

Rhenium solubility in borosilicate nuclear waste glass: implications for the processing and immobilization of technetium-99.

Science.gov (United States)

McCloy, John S; Riley, Brian J; Goel, Ashutosh; Liezers, Martin; Schweiger, Michael J; Rodriguez, Carmen P; Hrma, Pavel; Kim, Dong-Sang; Lukens, Wayne W; Kruger, Albert A

2012-11-20

The immobilization of technetium-99 ((99)Tc) in a suitable host matrix has proven to be a challenging task for researchers in the nuclear waste community around the world. In this context, the present work reports on the solubility and retention of rhenium, a nonradioactive surrogate for (99)Tc, in a sodium borosilicate glass. Glasses containing target Re concentrations from 0 to 10,000 ppm [by mass, added as KReO(4) (Re(7+))] were synthesized in vacuum-sealed quartz ampules to minimize the loss of Re from volatilization during melting at 1000 °C. The rhenium was found as Re(7+) in all of the glasses as observed by X-ray absorption near-edge structure. The solubility of Re in borosilicate glasses was determined to be ~3000 ppm (by mass) using inductively coupled plasma optical emission spectroscopy. At higher rhenium concentrations, additional rhenium was retained in the glasses as crystalline inclusions of alkali perrhenates detected with X-ray diffraction. Since (99)Tc concentrations in a glass waste form are predicted to be wastes, assuming Tc as Tc(7+) and similarities between Re(7+) and Tc(7+) behavior in this glass system.

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.

A review of phase separation in borosilicate glasses, with reference to nuclear fuel waste immobilization

International Nuclear Information System (INIS)

Taylor, P.

1990-08-01

This report reviews information on miscibility limits in borosilicate glass-forming systems. It includes both a literature survey and an account of experimental work performed within the Canadian Nuclear Fuel Waste Management Program. Emphasis is placed on the measurement and depiction of miscibility limits in multicomponent (mainly quaternary) systems, and the effects of individual components on the occurrence of phase separation. The behaviour of the multicomponent system is related to that of simpler (binary and ternary) glass systems. The possible occurrence of phase separation, as well as its avoidance, during processing of nuclear waste glasses is discussed

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

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

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

Science.gov (United States)

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

2017-05-01

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

Borosilicate glass as a matrix for immobilization of SRP high-level waste

International Nuclear Information System (INIS)

Wicks, G.G.

1980-01-01

Approximately 22 million gallons of high-level radioactive defense waste are currently being stored in large underground tanks located on the Savannah River Plant (SRP) site in Aiken, South Carolina. One option now being considered for long-term management of this waste involves removing the waste from the tanks, chemically processing the waste, and immobilizing the potentially harmful radionuclides in the waste into a borosilicate glass matrix. The technology for producing waste glass forms is well developed and has been demonstrated on various scales using simulated as well as radioactive SRP waste. Recently, full-scale prototypical equipment has been made operational at SRP. This includes both a joule-heated ceramic melter and an in-can melter. These melters are a part of an integrated vitrification system which is under evaluation and includes a spray calciner, direct liquid feed apparatus, and various elements of an off-gas system. Two of the most important properties of the waste glass are mechanical integrity and leachability. Programs are in progress at SRL aimed at minimizing thermally induced cracking by carefully controlling cooling cycles and using ceramic liners or coatings. The leachability of SRP waste glass has been studied under many different conditions and consistently found to be low. For example, the leachability of actual SRP waste glass was found to be 10 -6 to 10 -5 g/(cm 2 )(day) initially and decreasing to 10 -9 to 10 -8 g/(cm 2 )(day) after 100 days. Waste glass is also being studied under anticipated storage conditions. In brine at 90 0 C, the leachability is about 5 x 10 -8 g/(cm 2 )(day) after 60 days. The effects of other geological media including granite, basalt, shale, and tuff are also being studied as part of the multibarrier isolation system

Method of manufacturing borosilicate glass solidification products

International Nuclear Information System (INIS)

Tanaka, Tsuneya.

1986-01-01

Purpose: To obtain glass solidification products efficiently in a dry process from medium and high level radioactive liquid wastes discharged from PWR type reactors. Method: Boric acid-containing radioactive liquid wastes generated from primary coolants of PWR type reactors are evaporated to condensate as the pre-treatment. The concentrated liquid wastes are supplied to a drum type rotary kiln. While on the other hand, usual glass frits are introduced into the kiln. The liquid wastes are dried in the rotary kiln, as well as B 2 O 3 and the glass frits in the liquid wastes are combined into glass particles. In this case, since the kiln is rotated, no glass particles are deposited on the wall of the kiln. Then, the glass particles are introduced for melting into a high frequency melting furnace made of metal. The melting temperature is set to 1100 - 1150 deg C. The molten borosilicate glass is recovered from the bottom of the melting furance, contained in a canister and cooled for several hours, and then a cover is welded to the canister. (Ikeda, J.)

Effects of alpha radiation on hardness and toughness of the borosilicate glass applied to radioactive wastes immobilization

International Nuclear Information System (INIS)

Prado, Miguel Oscar; Bernasconi, Norma B. Messi de; Bevilacqua, Arturo Miguel; Arribere, Maria Angelica; Heredia, Arturo D.; Sanfilippo, Miguel

1999-01-01

Borosilicate german glass SG7 samples, obtained by frit sintering, were irradiated with different fluences of thermal neutrons in the nucleus of a nuclear reactor. The nuclear reaction 10 B(n,α) 7 Li, where the 10 B isotope is one of the natural glass components, was used to generate alpha particles throughout the glass volume. The maximum alpha disintegration per unit volume achieved was equivalent to that accumulated in a borosilicate glass with nuclear wastes after 3.8 million years. Through Vickers indentations values for microhardness, stress for 50% fracture probability (Weibull statistics) and estimation of the toughness were obtained as a function of alpha radiation dose. Two counterbalanced effects were found: that due to the disorder created by the alpha particles in the glass and that due to the annealing during irradiation (temperature below 240 deg C). Considering the alpha radiation effect, glasses tend decrease Vickers hardness, and to increase thr 50% fracture probability stress with the dose increase. (author)

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

RHENIUM SOLUBILITY IN BOROSILICATE NUCLEAR WASTE GLASS IMPLICATIONS FOR THE PROCESSING AND IMMOBILIZATION OF TECHNETIUM-99 (AND SUPPORTING INFORMATION WITH GRAPHICAL ABSTRACT)

Energy Technology Data Exchange (ETDEWEB)

AA KRUGER; A GOEL; CP RODRIGUEZ; JS MCCLOY; MJ SCHWEIGER; WW LUKENS; JR, BJ RILEY; D KIM; M LIEZERS; P HRMA

2012-08-13

The immobilization of 99Tc in a suitable host matrix has proved a challenging task for researchers in the nuclear waste community around the world. At the Hanford site in Washington State in the U.S., the total amount of 99Tc in low-activity waste (LAW) is {approx} 1,300 kg and the current strategy is to immobilize the 99Tc in borosilicate glass with vitrification. In this context, the present article reports on the solubility and retention of rhenium, a nonradioactive surrogate for 99Tc, in a LAW sodium borosilicate glass. Due to the radioactive nature of technetium, rhenium was chosen as a simulant because of previously established similarities in ionic radii and other chemical aspects. The glasses containing target Re concentrations varying from 0 to10,000 ppm by mass were synthesized in vacuum-sealed quartz ampoules to minimize the loss of Re by volatilization during melting at 1000 DC. The rhenium was found to be present predominantly as Re7 + in all the glasses as observed by X-ray absorption near-edge structure (XANES). The solubility of Re in borosilicate glasses was determined to be {approx}3,000 ppm (by mass) using inductively coupled plasma-optical emission spectroscopy (ICP-OES). At higher rhenium concentrations, some additional material was retained in the glasses in the form of alkali perrhenate crystalline inclusions detected by X-ray diffraction (XRD) and laser ablation-ICP mass spectrometry (LA-ICP-MS). Assuming justifiably substantial similarities between Re7 + and Tc 7+ behavior in this glass system, these results implied that the processing and immobilization of 99Tc from radioactive wastes should not be limited by the solubility of 99Tc in borosilicate LAW glasses.

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-05-15

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

Tc and Re Behavior in Borosilicate Waste Glass Vapor Hydration Tests

International Nuclear Information System (INIS)

McKeown, David A.; Buechele, Andrew C.; Pegg, Ian L.; Lukens, Wayne W.; Shuh, David K.

2007-01-01

Technetium (Tc), found in some nuclear wastes, is of particular concern with regard to long-term storage, because of its long-lived radioactivity and high mobility in the environment. Tc and rhenium (Re), commonly used as a non-radioactive surrogate for Tc, were studied to assess their behavior in borosilicate glass under hydrothermal conditions in the Vapor Hydration Test (VHT). X-ray absorption spectroscopy (XAS) and scanning electron microscopy (SEM) measurements were made on the original Tc- and Re-containing glasses and their corresponding VHT samples, and show different behavior for Tc and Re under VHT conditions. XAS indicates that, despite starting with different Tc(IV) and Tc(VII) distributions in each glass, the VHT samples have 100% Tc(IV)O 6 environments. SEM shows complete alteration of the original glass, Tc enrichment near the sample surface, and Tc depletion in the center. Perrhenate (Re(VII)O 4 - ) is dominant in both Re-containing samples before and after the VHT, where Re is depleted near the VHT sample surface and more concentrated toward the center. (authors)

Chemical durability of lead borosilicate glass matrix under simulated geological conditions

International Nuclear Information System (INIS)

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

2002-03-01

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

Phase separation of cesium from lead borosilicate glass by heat treatment under a reducing atmosphere

Energy Technology Data Exchange (ETDEWEB)

Xu, Zhanglian; Okada, Takashi, E-mail: t-okada@u-fukui.ac.jp; Nishimura, Fumihiro; Yonezawa, Susumu

2016-11-05

Highlights: • Cesium was phase separated from lead borosilicate glass under a reductive atmosphere. • The phase separation occurred on the glass surface that was in contact with the gas. • The leachability of cesium was enhanced by the phase separation. • The degree of such enhancement varied depending on the heat treatment conditions. - Abstract: A phase-separation technique for removing sodium from glass using a heat-treatment method under a reducing atmosphere was previously developed for sodium recovery from waste glass. In this study, this technique was applied to cesium-containing lead borosilicate glass to concentrate the cesium in phase-separated sodium-rich materials for efficient cesium extraction. The theoretical phase-separation temperature of the sodium-rich phase was simulated by thermodynamic equilibrium calculations and was predicted to occur below 700 °C for lead borosilicate glass. Experimentally, a simulated lead borosilicate glass was melted at 1000 °C and subsequently annealed below 700 °C under a CO-containing reducing atmosphere. The phase separation of cesium was found to occur with sodium enrichment on the glass surface that was in contact with the gas phase, promoting cesium extraction from the treated glass using water. The cesium extraction efficiency was affected by the surface area of the treated glass that was in contact with water, and under the examined conditions, the cesium extraction efficiency was up to 66%. Phase separation using reductive heat treatment, combined with a water leaching technique, is suggested to be effective for extracting cesium incorporated in borosilicate glass waste.

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-04-01

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

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

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)

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

Aqueous corrosion of borosilicate glasses. Nature and properties of alteration layers

International Nuclear Information System (INIS)

Trotignon, Laurent

1990-01-01

This research thesis addresses physical and chemical processes which occur during aqueous corrosion of silicates, and the study of the properties of their interfaces with solutions, and thus issues related to the fate of high activity nuclear wastes which are embedded in a vitreous matrix as the potential release of radionuclides towards the environment then depends on the glass parcel behaviour submitted to chemical attacks which could alter it, notably by aqueous corrosion. The objective is then to model the dissolution of nuclear glass over long periods of time, and to predict the behaviour of radionuclides. The author compared the corrosion and alteration layers of gradually more complex borosilicate glasses, from a ternary sodium borosilicate glass to a simulated nuclear glass (the French reference glass R7T7). Complexity is increased by adding oxides. After some theoretical recalls on the structure and corrosion of borosilicate glasses, the author presents the studied materials, the corrosion experiments, and analytical techniques used to study alteration layers. The mechanism of formation of altered layers is studied based on corrosion experiments performed at 90 C on the whole set of glasses. Alteration layers formed on corroded glasses are studied and compared by using various techniques: electronic microscopy, high energy ion beams, spectroscopy, infrared, photo-electron spectroscopy. Implications for underground storage of nuclear glasses are discussed

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

International Nuclear Information System (INIS)

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

1998-02-01

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

A kinetic approach of sulphur behaviour in borosilicate glasses and melts: implications for sulphate incorporation in nuclear waste glasses

Energy Technology Data Exchange (ETDEWEB)

Lenoir, Marion [Service de Confinement des Dechets et Vitrification - Laboratoire d' Etudes de Base sur les Verres, CEA Valrho, Centre de Marcoule, 30207 Bagnols sur Ceze (France); Physique des Mineraux et des Magmas, UMR 7047 - CNRS, Institut de Physique du Globe de Paris, 7 place Jussieu, 75252 Paris Cedex 05 (France); Grandjean, Agnes [Service de Confinement des Dechets et Vitrification - Laboratoire d' Etudes de Base sur les Verres, CEA Valrho, Centre de Marcoule, 30207 Bagnols sur Ceze (France); Neuville, Daniel R. [Physique des Mineraux et des Magmas, UMR 7047 - CNRS, Institut de Physique du Globe de Paris, 7 place Jussieu, 75252 Paris Cedex 05 (France)

2008-07-01

The kinetics of sulphate decomposition in a borosilicate melt were studied using in situ Raman spectroscopy. This technique permits the quantification of the amount of sulphate dissolved in a borosilicate glass as a function of heating time by comparison with measurements obtained by microprobe WDS (Wavelength Dispersive Spectrometry). In order to quantify the content of sulphate obtained by Raman spectroscopy, the integrated intensity of the sulphate band at 985 cm{sup -1} was scaled to the sum of the integrated bands between 800 and 1200 cm{sup -1}, bands that are assigned to Q{sup n} silica units on the basis of previous literature. Viscosities of some borosilicate glasses are also presented here in order to study the kinetics of sulphate decomposition as a function of the viscosity of the melt. This underlines the importance of variations in viscosity depending on the composition of the melt and thus shows that viscosity is an important parameter governing the kinetics of decomposition of sulphate in borosilicate glasses. (authors)

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

International Nuclear Information System (INIS)

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

1979-01-01

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

Diffusion of cesium in sodium-borosilicate glasses for nuclear waste immobilisation. Diffusie van cesium in natrium borosilicaat glazen voor het immobiliseren van radioaktief afval

Energy Technology Data Exchange (ETDEWEB)

Janssen, F.J.J.G.; Sengers, E.G.F. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands)); Waal, H. de (TPD-TNO-Glass technology, Eindhoven (Netherlands))

1989-09-26

Diffusion of cesium in borosilicate glass for high-level radioactive waste is discussed. For this purpose model glasses with non-radioactive elements are being made, in accordance with the specifications of the reprocessing plants, from which concentration couples are composed. A concentration couple consists of two cylinders of borosilicate glass which contain different amounts of cesium. After heat treatment the couples are studied by means of the scanning electron microscopy and X-ray microanalysis. The model study will provide a basis for predictions of the containment achieved over a longer period of time. (author). 11 refs.; 2 figs.; 2 tabs.

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Volatilization from borosilicate glass melts of simulated Savannah River Plant waste

International Nuclear Information System (INIS)

Wilds, G.W.

1979-01-01

Laboratory scale studies determined the rates at which the semivolatile components sodium, boron, lithium, cesium, and ruthenium volatilized from borosilicate glass melts that contained simulated Savannah River Plant waste sludge. Sodium and boric oxides volatilize as the thermally stable compound sodium metaborate, and accounted for approx. 90% of the semivolatiles that evolved. The amounts of semivolatiles that evolved increased linearly with the logarithm of the sodium content of the glass-forming mixture. Cesium volatility was slightly suppressed when titanium dioxide was added to the melt, but was unaffected when cesium was added to the melt as a cesium-loaded zeolite rather than as a cesium carbonate solution. Volatility of ruthenium was not suppressed when the glass melt was blanketed with a nonoxidizing atmosphere. Trace quantities of mercury were removed from vapor streams by adsorption onto a silver-exchanged zeolite. A bed containing silver in the ionic state removed more than 99.9% of the mercury and had a high chemisorption capacity. Beds of lead-, copper-, and copper sulfide-exchanged zeolite-X and also an unexchanged zeolite-X were tested. None of these latter beds had high removal efficiency and high chemisorption capacity

Volatilization from borosilicate glass melts of simulated Savannah River Plant waste

International Nuclear Information System (INIS)

Wilds, G.W.

1978-01-01

Laboratory scale studies determined the rates at which the semivolatile components sodium, boron, lithium, cesium, and ruthenium volatilized from borosilicate glass melts that contained simulated Savannah River Plant waste sludge. Sodium and boric oxides volatilize as the thermally stable compound sodium metaborate, and accounted for approx. 90% of the semivolatiles that evolved. The amounts of semivolatiles that evolved increased linearly with the logarithm of the sodium content of the glass-forming mixture. Cesium volatility was slightly suppressed when titanium dioxide was added to the melt, but was unaffected when cesium was added to the melt as a cesium-loaded zeolite rather than as a cesium carbonate solution. Volatility of ruthenium was not suppressed when the glass melt was blanketed with a nonoxidizing atmosphere. Trace quantities of mercury were removed from vapor streams by adsorption onto a silver-exchanged zeolite. A bed containing silver in the ionic state removed more than 99.9% of the mercury and had a high chemisorption capacity. Beds of lead-, copper-, and copper sulfide-exchanged zeolite-X and also an unexchanged zeolite-X were tested. None of these latter beds had high removal efficiency and high chemisorption capacity

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

International Nuclear Information System (INIS)

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

2002-01-01

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

57Fe Moessbauer effect in borosilicate glasses

International Nuclear Information System (INIS)

Music, S.

1989-01-01

The present study was carried out to elucidate the valence state of iron and its co-ordination in borosilicate glasses, which are being investigated as possible solidification matrices for the immobilization of a simulated nuclear waste. 57 Fe Mossbauer spectroscopy was used as the experimental technique. The chemical compositions of glass samples and the experimental conditions for the preparation of these samples are given. Iron in the form of haematite (α-Fe 2 O 3 ) was used as doping material. Details of the experimental procedure have previously been described. Isomer shifts are calculated relative to α-iron. The results indicate a strong dependence of the valency of the iron and its coordination on the chemical composition of the glass and the Fe 2 O 3 content. The method of preparing the glasses also influences the state of the iron in oxide glasses. (Author)

Alteration of borosilicate glass under water saturated condition

International Nuclear Information System (INIS)

Tsukimura, K.; Matsuhisa, Y.; Kanai, Y.; Hirai, H.

1991-01-01

In the long term of the high-level nuclear waste disposal, the dissolving rates of radionuclides from the waste will be controlled by crystalline phases secondarily formed in the borosilicate glass by reaction with ground water. We have determined the crystalline phases formed from borosilicate glass in a hydrothermal experiment. Borosilicate glass powder (50 mg, Si 40 B 20 Al 4 Li 12 Na 14 K 3 Ca 2 Fe 5 O 140 ) and 200 ml of water were reacted at 300degC and 300 bars for 2 weeks with a cold-seal type hydrothermal high pressure apparatus. After the reaction the liquid phase was analyzed with atomic absorption spectroscopy and the solid phases with x-ray powder diffraction and analytical electron microscopy. The liquid phase contains 58 % of Na, 14 % of Li and 11 % of K but less than 1 % of Ca and Fe. The solid phases were found to be zeolite and smectite. The zeolite is isostructural with analcime Na 2 Al 2 (SiO 3 ) 4 · 2H 2 O or wairakite CaAl 2 (SiO 3 ) 4 · 2H 2 O and contains Ca and Si as major constituents but does not contain Al. Although no direct analytical evidence is available, logical induction of the experimental results indicates that the chemical composition of the zeolite is CaB 2 (SiO 3 ) 4 . The remaining x-ray powder diffraction peaks are same as those of smectite Na 06.7 (Al 3.33 Mg 0.67 )Si 8 O 20 (OH) 4 . However, since our starting material was free of Mg, and Na was not detected by AEM, the smectite may contain Fe instead of Mg, and Li instead of Na. (author)

Thermodynamic properties of alkali borosilicate gasses and metaborates

International Nuclear Information System (INIS)

Asano, Mitsuru

1992-01-01

Borosilicate glasses are the proposed solidifying material for storing high level radioactive wastes in deep underground strata. Those have low melting point, and can contain relatively large amount of high level radioactive wastes. When borosilicate glasses are used for this purpose, they must be sufficiently stable and highly reliable in the vitrification process, engineered storage and the disposal in deep underground strata. The main vaporizing components from borosilicate glasses are alkali elements and boron. In this report, as for the vaporizing behavior of alkali borosilicate glasses, the research on thermodynamic standpoint carried out by the authors is explained, and the thermodynamic properties of alkali metaborates of monomer and dimer which are the main evaporation gases are reported. The evaporation and the activity of alkali borosilicate glasses, the thermodynamic properties of alkali borosilicate glasses, gaseous alkali metaborates and alkali metaborate system solid solution and so on are described. (K.I.)

Effect of various lead species on the leaching behavior of borosilicate waste glass

International Nuclear Information System (INIS)

Lehman, R.L.; Kuchinski, F.A.

1984-01-01

A borosilicate nuclear waste glass was static leached in pure water, silicate water, and brine solution. Three different forms of lead were included in specified corrosion cells to assess the extent to which various lead species alter the leaching behavior of the glass. Weight loss data indicated that Pb/sub m/ amd PbO greatly reduce the weight loss of glass when leached in pure water, and similar effects were noted in silicate and brine. Si concentrations, which were substantial in the glass-alone leachate, were reduced to below detection limits in all pure water cells containing a lead form. Lead concentration levels in the leachate were controlled by lead form solubility and appeared to be a significant factor in influencing apparent leaching behavior. Surface analysis revealed surface crystals, which probably formed when soluble lead in the leachate reacted with dissolved or activated silica at the glass surface. The net effect was to reduce the lease of some glass constituents to the leachate, although it was not clear whether the actual corrosion of the glass surface was reduced. Significantly different corrosion inhibiting effects were noted among lead metal and two forms of lead oxide. 9 refs., 7 figs., 3 tabs

Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories

International Nuclear Information System (INIS)

1983-06-01

The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

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

Calcium-borosilicate glass-ceramics wasteforms to immobilize rare-earth oxide wastes from pyro-processing

Energy Technology Data Exchange (ETDEWEB)

Kim, Miae [Department of Materials Science and Engineering and Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), Pohang, Gyeongbuk, 790-784 (Korea, Republic of); Heo, Jong, E-mail: jheo@postech.ac.kr [Department of Materials Science and Engineering and Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), Pohang, Gyeongbuk, 790-784 (Korea, Republic of); Department of Materials Engineering, Adama Science and Technology University (ASTU), PO Box 1888, Adama (Ethiopia)

2015-12-15

Glass-ceramics containing calcium neodymium(cerium) oxide silicate [Ca{sub 2}Nd{sub 8-x}Ce{sub x}(SiO{sub 4}){sub 6}O{sub 2}] crystals were fabricated for the immobilization of radioactive wastes that contain large portions of rare-earth ions. Controlled crystallization of alkali borosilicate glasses by heating at T ≥ 750 °C for 3 h formed hexagonal Ca–silicate crystals. Maximum lanthanide oxide waste loading was >26.8 wt.%. Ce and Nd ions were highly partitioned inside Ca–silicate crystals compared to the glass matrix; the rare-earth wastes are efficiently immobilized inside the crystalline phases. The concentrations of Ce and Nd ions released in a material characterization center-type 1 test were below the detection limit (0.1 ppb) of inductively coupled plasma mass spectroscopy. Normalized release values performed by a product consistency test were 2.64·10{sup −6} g m{sup −2} for Ce ion and 2.19·10{sup −6} g m{sup −2} for Nd ion. Results suggest that glass-ceramics containing calcium neodymium(cerium) silicate crystals are good candidate wasteforms for immobilization of lanthanide wastes generated by pyro-processing. - Highlights: • Glass-ceramic wasteforms containing Ca{sub 2}Nd{sub 8-x}Ce{sub x}(SiO{sub 4}){sub 6}O{sub 2} crystals were synthesized to immobilize lanthanide wastes. • Maximum lanthanide oxide waste loading was >26.8 wt.%. • Ce and Nd ions were highly partitioned inside Ca–Nd–silicate crystals compared to glass matrix. • Amounts of Ce and Nd ions released in the material characterization center-type 1 were below the detection limit (0.1 ppb). • Normalized release values performed by a PCT were 2.64• 10{sup −6} g m{sup −2} for Ce ions and 2.19• 10{sup −6} g m{sup −2} for Nd ions.

Borosilicate glass for gamma irradiation fields

Science.gov (United States)

Baydogan, N.; Tugrul, A. B.

2012-11-01

Four different types of silicate glass specimens were irradiated with gamma radiation using a Co-60 radioisotope. Glass specimens, with four different chemical compositions, were exposed to neutron and mixed neutron/gamma doses in the central thimble and tangential beam tube of the nuclear research reactor. Optical variations were determined in accordance with standardisation concept. Changes in the direct solar absorbance (αe) of borosilicate glass were examined using the increase in gamma absorbed dose, and results were compared with the changes in the direct solar absorbance of the three different type silicate glass specimens. Solar absorption decreased due to decrease of penetration with absorbed dose. αe of borosilicate increased considerably when compared with other glass types. Changes in optical density were evaluated as an approach to create dose estimation. Mixed/thermal neutron irradiation on glass caused to increse αe.

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Irradiation test of borosilicate glass burnable poison

International Nuclear Information System (INIS)

Feng Mingquan; Liao Zumin; Yang Mingjin; Lu Changlong; Huang Deyang; Zeng Wangchun; Zhao Xihou

1991-08-01

The irradiation test and post-irradiation examinations for borosilicate glass burnable poison are introduced. Examinations include visual examination, measurement of dimensions and density, and determination of He gas releasing and 10 B burnup. The corrosion and phenomenon of irradiation densification are also discussed. Two type glass samples have been irradiated with different levels of neutron flux. It proved that the GG-17 borosilicate glass can be used as burnable poison to replace the 10 B stainless steel in the Qinshan Nuclear Power Plant, and it is safe, economical and reasonable

Preliminary assessment of modified borosilicate glasses for chromium and ruthenium immobilization

Energy Technology Data Exchange (ETDEWEB)

Farid, Osama M. [Reactors Department, Nuclear Research Center, Atomic Energy Authority of Egypt, P.O. 13759, Inshas, Cairo (Egypt); Centre of Nuclear Engineering (CNE), Department of Materials, Imperial College London, London, SW7 2BP (United Kingdom); Abdel Rahman, R.O., E-mail: alaarehab@yahoo.com [Hot Laboratory Center, Atomic Energy Authority of Egypt, P.O. 13759, Inshas, Cairo (Egypt)

2017-01-15

The feasibility of using modified alkali borosilicate glasses for ruthenium and chromium immobilization is preliminary assessed by investigating the immobilization system structure under normal conditions. Within this context, reference alkali borosilicate, and simulated Magnox-modified glasses were prepared and studied. The results indicate that ruthenium is immobilized in the vitreous structure as encapsulated RuO{sub 2} crystallites that act as seeds for heterogeneous nucleation of other crystalline phases. The presence of Zn, as modifier, has contributed to chromium immobilization in zincochromite spinel structure, whereas Ca is accommodated in the vitreous structure. Immobilization performance was evaluated by conducting conservative static leach test and studying the leached glass. Leached glass morphology was altered, where near surface reference glass is leached over 400 nm and simulated Magnox-modified sample is altered over 300 nm. Normalized release rates are within normal range for borosilicate material. For simulated Magnox-modified sample, Ca and alkali structural element, i.e. Na and Li, are leached via ion-exchange reaction. The ion-exchanged fraction equals 1.06 × 10{sup −8} mol/m{sup 2} s and chromium has slightly lower normalized release rate value than ruthenium. - Highlights: • The presence of modifiers and waste oxides led to localized de-vitrification. • Ruthenium is encapsulated within the vitreous glass network as RuO{sub 2} crystals. • Chromium is immobilized within the zincochromite spinel structure. • Pitting and cracks induced by leaching did not affect the immobilization performance.

Basaltic glasses from Iceland and the deep sea: Natural analogues to borosilicate nuclear waste-form glass

International Nuclear Information System (INIS)

Jercinovic, M.J.; Ewing, R.C.

1987-12-01

The report provides a detailed analysis of the alteration process and products for natural basaltic glasses. Information of specific applicability to the JSS project include: * The identification of typical alteration products which should be expected during the long-term corrosion process of low-silica glasses. The leached layers contain a relatively high proportion of crystalline phases, mostly in the form of smectite-type clays. Channels through the layer provide immediate access of solutions to the fresh glass/alteration layer interface. Thus, glasses are not 'protected' from further corrosion by the surface layer. * Corrosion proceeds with two rates - an initial rate in silica-undersaturated environments and a long-term rate in silica-saturated environments. This demonstrates that there is no unexpected change in corrosion rate over long periods of time. The long-term corrosion rate is consistent with that of borosilicate glasses. * Precipitation of silica-containing phases can result in increased alteration of the glass as manifested by greater alteration layer thicknesses. This emphasizes the importance of being able to predict which phases form during the reaction sequence. * For natural basaltic glasses the flow rate of water and surface area of exposed glass are critical parameters in minimizing glass alteration over long periods of time. The long-term stability of basalt glasses is enhanced when silica concentrations in solution are increased. In summary, there is considerable agreement between corrosion phenomena observed for borosilicate glasses in the laboratory and those observed for natural basalt glasses of great age. (With 121 refs.) (authors)

Surface chemistry and durability of borosilicate glass

International Nuclear Information System (INIS)

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

1994-01-01

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

Topological Principles of Borosilicate Glass Chemistry - An Invited Talk

DEFF Research Database (Denmark)

Mauro, J.C.; Smedskjær, Morten Mattrup; Youngman, R. E.

Borosilicate glasses display a rich complexity of chemical behavior depending on the details of their composition and thermal history. We investigate the topological principles of borosilicate glass chemistry covering the extremes from pure borate to pure silicate end members. Based on NMR...

Stress corrosion in a borosilicate glass nuclear wasteform

International Nuclear Information System (INIS)

Ringwood, A.E.; Willis, P.

1984-01-01

The authors discuss a typical borosilicate glass wasteform which, when exposed to water vapour and water for limited periods, exhibits evidence of stress corrosion cracking arising from the interaction of polar OH groups with stressed glass surfaces. Glass wasteforms may experience similar stress corrosion cracking when buried in a geological repository and exposed to groundwaters over an extended period. This would increase the effective surface areas available for leaching by groundwater and could decrease the lifetime of the wasteform. Conventional leach-testing methods are insensitive to the longer-term effects of stress corrosion cracking. It is suggested that specific fracture-mechanics tests designed to evaluate susceptibility to stress corrosion cracking should be used when evaluating the wasteforms for high-level nuclear wastes. (author)

Commercial Ion Exchange Resin Vitrification in Borosilicate Glass

International Nuclear Information System (INIS)

Cicero-Herman, C.A.; Workman, P.; Poole, K.; Erich, D.; Harden, J.

1998-05-01

Bench-scale studies were performed to determine the feasibility of vitrification treatment of six resins representative of those used in the commercial nuclear industry. Each resin was successfully immobilized using the same proprietary borosilicate glass formulation. Waste loadings varied from 38 to 70 g of resin/100 g of glass produced depending on the particular resin, with volume reductions of 28 percent to 68 percent. The bench-scale results were used to perform a melter demonstration with one of the resins at the Clemson Environmental Technologies Laboratory (CETL). The resin used was a weakly acidic meth acrylic cation exchange resin. The vitrification process utilized represented a approximately 64 percent volume reduction. Glass characterization, radionuclide retention, offgas analyses, and system compatibility results will be discussed in this paper

Preparation and characterization of an improved borosilicate glass matrix for the incorporation of high level radioactive waste (HAW). Pt. 1

International Nuclear Information System (INIS)

Guber, W.; Hussain, M.; Kahl, L.; Ondracek, G.; Saidl, J.; Dippel, T.

1979-08-01

On the basis of laboratory and technical experience with the preparation and the characterization of borosilicate glasses as solidification matrix for high level radioactive waste solution (HAW), a borosilicate glass composition with optimum properties has been developed. Keeping in view the technical and final storage requirements, a number of glass compositions with varying proportions of influential components as Al, Mg, Na were prepared and thoroughly investigated for certain parameters as specific gravity, thermal conductivity, impact resistance, thermal expansion, viscosity, characteristic temperature points, specific heat, evaporation losses from the melt, electrical conductivity, leach resistance, tendency toward recrystallization and second phase formation. All the compositions (some with different amounts of Gd 2 O 3 , an expected neutron poision) contained 15 wt. % simulated HAW oxides. Samples for investigation were fabricated very close to the actual process conditions of vitrification. Two glass products GP12 and GP26 (3.7% Gd 2 O 3 ) have been selected out of 25 glasses as the optimised products for further thorough investigations. Leach resistance, viscosity at 1420 K, tendency towards recrystallization and second phase formation were the most important deciding factors. (orig.) [de

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

The Dissolution Behavior of Borosilicate Glasses in Far-From Equilibrium Conditions

Energy Technology Data Exchange (ETDEWEB)

Neeway, James J.; Rieke, Peter C.; Parruzot, Benjamin PG; Ryan, Joseph V.; Asmussen, Robert M.

2018-02-10

In far-from-equilibrium conditions, the dissolution of borosilicate glasses used to immobilize nuclear waste is known to be a function of both temperature and pH. The aim of this paper is to study effects of these variables on three model waste glasses (SON68, ISG, AFCI). To do this, experiments were conducted at temperatures of 23, 40, 70, and 90 °C and pH(RT) values of 9, 10, 11, and 12 with the single-pass flow-through (SPFT) test method. The results from these tests were then used to parameterize a kinetic rate model based on transition state theory. Both the absolute dissolution rates and the rate model parameters are compared with previous results. Discrepancies in the absolute dissolution rates as compared to those obtained using other test methods are discussed. Rate model parameters for the three glasses studied here are nearly equivalent within error and in relative agreement with previous studies. The results were analyzed with a linear multivariate regression (LMR) and a nonlinear multivariate regression performed with the use of the Glass Corrosion Modeling Tool (GCMT), which is capable of providing a robust uncertainty analysis. This robust analysis highlights the high degree of correlation of various parameters in the kinetic rate model. As more data are obtained on borosilicate glasses with varying compositions, the effect of glass composition on the rate parameter values could possibly be obtained. This would allow for the possibility of predicting the forward dissolution rate of glass based solely on composition

The dissolution behavior of borosilicate glasses in far-from equilibrium conditions

Science.gov (United States)

Neeway, James J.; Rieke, Peter C.; Parruzot, Benjamin P.; Ryan, Joseph V.; Asmussen, R. Matthew

2018-04-01

An area of agreement in the waste glass corrosion community is that, at far-from-equilibrium conditions, the dissolution of borosilicate glasses used to immobilize nuclear waste is known to be a function of both temperature and pH. The aim of this work is to study the effects of temperature and pH on the dissolution rate of three model nuclear waste glasses (SON68, ISG, AFCI). The dissolution rate data are then used to parameterize a kinetic rate model based on Transition State Theory that has been developed to model glass corrosion behavior in dilute conditions. To do this, experiments were conducted at temperatures of 23, 40, 70, and 90 °C and pH (22 °C) values of 9, 10, 11, and 12 with the single-pass flow-through (SPFT) test method. Both the absolute dissolution rates and the rate model parameters are compared with previous results. Rate model parameters for the three glasses studied here are nearly equivalent within error and in relative agreement with previous studies though quantifiable differences exist. The glass dissolution rates were analyzed with a linear multivariate regression (LMR) and a nonlinear multivariate regression performed with the use of the Glass Corrosion Modeling Tool (GCMT), with which a robust uncertainty analysis is performed. This robust analysis highlights the high degree of correlation of various parameters in the kinetic rate model. As more data are obtained on borosilicate glasses with varying compositions, a mathematical description of the effect of glass composition on the rate parameter values should be possible. This would allow for the possibility of calculating the forward dissolution rate of glass based solely on composition. In addition, the method of determination of parameter uncertainty and correlation provides a framework for other rate models that describe the dissolution rates of other amorphous and crystalline materials in a wide range of chemical conditions. The higher level of uncertainty analysis would provide

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

Energy Technology Data Exchange (ETDEWEB)

Tait, J C

1993-05-01

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

Direct conversion of plutonium metal, scrap, residue, and transuranic waste to glass

International Nuclear Information System (INIS)

Forsberg, C.W.; Beahm, E.C.; Parker, G.W.; Malling, J.F.; Rudolph, J.

1995-01-01

A method for the direct conversion of metals, ceramics, organics, and amorphous solids to borosilicate glass has been invented. The process is called the Glass Material Oxidation and Dissolution System (GMODS). Traditional glass-making processes can convert only oxide materials to glass. However, many wastes contain complex mixtures of metals, ceramics, organics, and amorphous solids. Conversion of such mixtures to oxides followed by their conversion to glass is often impractical. GMODS may create a practical method to convert such mixtures to glass. Plutonium-containing materials (PCMS) exist in many forms, including metals, ceramics, organics, amorphous solids, and mixtures thereof. These PCMs vary from plutonium metal to filters made of metal, organic binders, and glass fibers. For storage and/or disposal of PCMS, it is desirable to convert PCMs to borosilicate glass. Borosilicate glass is the preferred repository waste form for high-level waste (HLW) because of its properties. PCMs converted to a transuranic borosilicate homogeneous glass would easily pass all waste acceptance and storage criteria. Conversion of PCMs to a glass would also simplify safeguards by conversion of heterogeneous PCMs to homogeneous glass. Thermodynamic calculations and proof-of-principle experiments on the GMODS process with cerium (plutonium surrogate), uranium, stainless steel, aluminum, Zircaloy-2, and carbon were successfully conducted. Initial analysis has identified potential flowsheets and equipment. Major unknowns remain, but the preliminary data suggests that GMODS may be a major new treatment option for PCMs

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

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

International Nuclear Information System (INIS)

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

2002-12-01

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

Alpha self irradiation effects in nuclear borosilicate glass

International Nuclear Information System (INIS)

Peuget, S.; Roudil, D.; Deschanels, X.; Jegou, C.; Broudic, V.; Bart, J.M.

2004-01-01

The properties of actinide glasses are studied in the context of high-level waste management programs. Reprocessing high burnup fuels in particular will increase the minor actinide content in the glass package, resulting in higher cumulative alpha decay doses in the glass, and raising the question of the glass matrix behavior and especially its containment properties. The effect of alpha self-irradiation on the glass behavior is evaluated by doping the glass with a short-lived actinide ( 244 Cm) to reach in several years the alpha dose received by the future glass packages over several thousand years. 'R7T7' borosilicate glasses were doped with 3 different curium contents (0.04, 0.4 and 1.2 wt% 244 CmO 2 ). The density and mechanical properties of the curium-doped glasses were characterized up to 2. 10 18 α/g, revealing only a slight evolution of the macroscopic behavior of R7T7 glass in this range. The leaching behavior of curium-doped glass was also studied by Soxhlet tests. The results do not show any significant evolution of the initial alteration rate with the alpha dose. (authors)

Enhancement of thermal neutron shielding of cement mortar by using borosilicate glass powder.

Science.gov (United States)

Jang, Bo-Kil; Lee, Jun-Cheol; Kim, Ji-Hyun; Chung, Chul-Woo

2017-05-01

Concrete has been used as a traditional biological shielding material. High hydrogen content in concrete also effectively attenuates high-energy fast neutrons. However, concrete does not have strong protection against thermal neutrons because of the lack of boron compound. In this research, boron was added in the form of borosilicate glass powder to increase the neutron shielding property of cement mortar. Borosilicate glass powder was chosen in order to have beneficial pozzolanic activity and to avoid deleterious expansion caused by an alkali-silica reaction. According to the experimental results, borosilicate glass powder with an average particle size of 13µm showed pozzolanic activity. The replacement of borosilicate glass powder with cement caused a slight increase in the 28-day compressive strength. However, the incorporation of borosilicate glass powder resulted in higher thermal neutron shielding capability. Thus, borosilicate glass powder can be used as a good mineral additive for various radiation shielding purposes. Copyright © 2017 Elsevier Ltd. All rights reserved.

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

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

International Nuclear Information System (INIS)

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

2004-07-01

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

Investigation of U3O8 immobilization in the GP-91 borosilicate glass by induction melter with a cold crucible (CCIM)

International Nuclear Information System (INIS)

Matyunin, Y.I.; Demin, A.V.; Smelova, T.V.; Yudintsev, S.V.; Lapina, M.I.

1997-01-01

One of the most promising and intensively developed methods for the solidification of high-level wastes is their vitrification with the use of a cold crucible induction melter (CCIM), which offers a number of advantages over ceramic melter. This work is concerned with comparison studies on the behavior of uranium in vitreous borosilicate materials synthesized by the traditional technique (melting in muffle furnaces) and CCIM method. The incorporation of uranium oxide U 3 O 8 into the GP-91 borosilicate glass with the use of CCIM technology is investigated. The limiting solubility of uranium in the GP-91 borosilicate glass is evaluated. The phase composition of precipitated dispersed particles based on uranium is determined. Some physicochemical properties of synthesized materials are explored. Investigations into the behavior of uranium in borosilicate glass prepared in the CCIM show a feasibility to synthesize the X-ray amorphous homogeneous borosilicate glasses incorporating as much as 25 - 28 wt% uranium, which is 4 - 5 times larger than that in glasses obtained by the traditional method. (author)

Low Velocity Sphere Impact of a Borosilicate Glass

Energy Technology Data Exchange (ETDEWEB)

Morrissey, Timothy G [ORNL; Ferber, Mattison K [ORNL; Wereszczak, Andrew A [ORNL; Fox, Ethan E [ORNL

2012-05-01

This report summarizes US Army TARDEC sponsored work at Oak Ridge National Laboratory (ORNL) involving low velocity (< 30 m/s or < 65 mph) ball impact testing of Borofloat borosilicate glass, and is a follow-up to a similar study completed by the authors on Starphire soda-lime silicate glass last year. The response of the borosilicate glass to impact testing at different angles was also studied. The Borofloat glass was supplied by the US Army Research Laboratory and its tin-side was impacted or indented. The intent was to better understand low velocity impact response in the Borofloat. Seven sphere materials were used whose densities bracket that of rock: borosilicate glass, soda-lime silicate glass, silicon nitride, aluminum oxide, zirconium oxide, carbon steel, and a chrome steel. A gas gun or a ball-drop test setup was used to produce controlled velocity delivery of the spheres against the glass tile targets. Minimum impact velocities to initiate fracture in the Borofloat were measured and interpreted in context to the kinetic energy of impact and the elastic property mismatch between the seven sphere-Borofloat-target combinations. The primary observations from this low velocity (< 30 m/s or < 65 mph) testing were: (1) BS glass responded similarly to soda-lime silicate glass when spherically indented but quite differently under sphere impact conditions; (2) Frictional effects contributed to fracture initiation in BS glass when it spherically indented. This effect was also observed with soda-lime silicate glass; (3) The force necessary to initiate fracture in BS glass under spherical impact decreases with increasing elastic modulus of the sphere material. This trend is opposite to what was observed with soda-lime silicate glass. Friction cannot explain this trend and the authors do not have a legitimate explanation for it yet; (4) The force necessary to initiate contact-induced fracture is higher under dynamic conditions than under quasi-static conditions. That

Preparation and characterization of an improved borosilicate glass for the solidification of high level radioactive fission product solutions (HLW). Pt. 2

International Nuclear Information System (INIS)

Kahl, L.; Ruiz-Lopez, M.C.; Saidl, J.; Dippel, T.

1982-04-01

In the 'Institut fuer Nuklare Entsorgungstechnik' the borosilicate glass VG 98/12 has been developed for the solidification of the high level radioactive waste (HLW). This borosilicate glass can be used in a direct heated ceramic melter and forms together with the HLW the borosilicate glass product GP 98/12. This borosilicate glass product has been examined in detail both in liquid and solid state. The elements contained in the HLW can be incorporated without problems. Only in a few exceptions the concentration must be kept below certain limits to exclude the formation of a second phase ('yellow phase') by separation. No spontaneous crystallization and no crystallization over a long time could be observed as long as the temperature of the borosilicate glass product is kept below its transformation area. Simulating accidental conditions in the final storage, samples had been leached at temperatures up to 200 0 C and pressures up to 130 bar with saturated rock salt brine and saturated quinary salt brine. The leaching process seems to be stopped by the formed 'leached layer' on the surface of the borosilicate glass product after a limited leaching time. Detailed investigations have been started to explain this phenomenon. (orig.) [de

Economic comparison of crystalline ceramic and glass waste forms for HLW disposal

International Nuclear Information System (INIS)

McKee, R.W.; Daling, P.M.; Wiles, L.E.

1983-05-01

A titanate-based, crystalline ceramic produced by hot isostatic pressing has been proposed as a potentially more stable and improved waste form for high-level nuclear waste disposal compared to the currently favored borosilicate glass waste form. This paper describes the results of a study to evaluate the relative costs for disposal of high-level waste from a 70,000 metric ton equivalent (MTE) system. The entire waste management system, including waste processing and encapsulation, transportation, and final repository disposal, was included in this analysis. The repository concept is based on the current basalt waste isolation project (BWIP) reference design. A range of design basis alternatives is considered to determine if this would influence the relative economics of the two waste forms. A thermal analysis procedure was utilized to define optimum canister sizes to assure that each waste form was compared under favorable conditions. Repository costs are found to favor the borosilicate glass waste form while transportation costs greatly favor the crystalline ceramic waste form. The determining component in the cost comparison is the waste processing cost, which strongly favors the borosilicate glass process because of its relative simplicity. A net cost advantage on the order of 12% to 15% on a waste management system basis is indicated for the glass waste form

Studies on the accelerated chemical durability test and its impact on structural changes in borosilicate glass matrix

International Nuclear Information System (INIS)

Thorat, Vidya S.; Kadam, P.S.; Mishra, R.K.; Kumar, Amar; Kaushik, C.P.; Sudarsan, V.; Tyagi, A.K.

2015-01-01

Sodium borosilicate glass used for the immobilisation of high level nuclear waste with tentative compositions (SiO 2 ) 0.477 (B 2 O 3 ) 0.239 (Na 2 O) 0.170 (TiO 2 ) 0.02 )3 (CaO) 0.068 (Al 2 O 3 ) 0.023 was evaluated for its long term stability in the repository conditions where it may be exposed to high temperature and pressure. 29 Si MAS NMR studies have confirmed that, upon leaching, sodium borosilicate glass undergoes congruent crystallization leading to the formation of an aluminosilicate phase, Na 6 Al 6 Si 10 O 32 . Further the residual glass structure in the composite sample is identical with that of the un-leached glass. Boron structural units are unaffected in the glass compositions upon leaching

Lead-iron phosphate glass: a stable storage medium for high-level nuclear waste

International Nuclear Information System (INIS)

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

1984-01-01

Results are presented which show that lead-iron phosphate glasses are a promising new waste form for the safe immobilization of both high-level defense and high-level commercial radioactive waste. Relative to the borosilicate nuclear waste glasses that are currently the ''reference'' waste form for the long-term disposal of nuclear waste, lead-iron phosphate glasses have several distinct advantages: (1) an aqueous corrosion rate that is about 1000 times lower, (2) a processing temperature that is 100 0 to 250 0 C lower and, (3) a much lower melt viscosity in the temperature range from 800 0 to 1000 0 C. Most significantly, the lead-iron phosphate waste form can be processed using a technology similar to that developed for borosilicate nuclear waste glasses

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

Influence of Mo on the structure of borosilicate glass for the immobilization of high level waste

International Nuclear Information System (INIS)

Petrov, I.; Dimitriev, Y.; Kashchieva, E.

2015-01-01

It has been shown that the classic multi-borosilicate glass with concentration of MoO 3 has a low melting in the range of 1300 to 1400 ° C, but for achieving of complete homogenization and vitrification it is required temperature above 1400 ° C. Under this temperature in all tested samples areas of crystallization were observed, due to incomplete vitrification processes. It is understood that in all multicomponent borosilicate glass featuring MoO 3 , occur micro vitrification processes, and at a concentration of MoO 3 above 20% - macro segregation; It has been shown that the introduction of Nd in compound borosilicate glass featuring MoO 3 is observed crystallization of the phase Na 0.5 Nd 0.5 Mo, and it has been also found that the phase Na 0.5 Nd 0.5 Mo 4 can be synthesized either by solid phase reaction or supercooled melt. The results of surveys show that It is possible to prevent the occurrence of liquid phase separation in the studied multicomponent glass. From a structural point of view, the cause of liquid phase separation is the result of structural incompatibility of molybdenum units with structural units borosilicate network since not been established links Mo-O-B-O and Mo-Si. From a thermodynamic point of view in the lamination multi windows may be due to overlapping areas of delamination in component binary and triple systems. From the kinetic point of view of course the liquid phase by settling, and crystallization may be due to imbalances conditions the cooling process, in the course of which flow various processes of imbalances metastable settling, followed by crystallization of molybdate phases

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

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

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

Topological Principles of Borosilicate Glass Chemistry

DEFF Research Database (Denmark)

Smedskjær, Morten Mattrup; Mauro, J. C.; Youngman, R. E.

2011-01-01

and laboratory glassware to high-tech applications such as liquid crystal displays. In this paper, we investigate the topological principles of borosilicate glass chemistry covering the extremes from pure borate to pure silicate end members. Based on NMR measurements, we present a two-state statistical...

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

International Nuclear Information System (INIS)

Ghezal, A.

1987-09-01

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

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)

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)

Irradiations effects on the structure of boro-silicated glasses: long term behaviour of nuclear waste glassy matrices

International Nuclear Information System (INIS)

Bonfils, J. de

2007-09-01

This work deals with the long term behaviour of R7T7-type nuclear waste glasses and more particularly of non-active boro-silicated glasses made up of 3 or 5 oxides. Radioactivity of active glasses is simulated by multi energies ions implantations which reproduce the same defects. The damages due to the alpha particles are simulated by helium ions implantations and those corresponding to the recoil nucleus are obtained with gold ions ones. Minor actinides, stemming from the used fuel, is simulated by trivalent rare-earths (Eu 3+ and Nd 3+ ). In a first part, we have shown by macroscopic experiments (Vickers hardness - swelling) and optical spectroscopies (Raman - ATR-IR) that the structure of the glassy matrices is modified under implantations until a dose of 2,3.10 13 at.cm -2 , which corresponds to a R7T7 storage time estimated at 300 years. Beyond this dose, no additional modifications have been observed. The second part concerns the local environment of the rare-earth ions in glasses. Two different environments were found and identified as follows: one is a silicate rich one and the other is attributed to a borate rich one. (author)

Influence of processing conditions on the glass-crystal transition into borosilicate glasses

Energy Technology Data Exchange (ETDEWEB)

Deschanels, X.; Cachia, J.N.; Lopez, C.; Peuget, S. [CEA Marcoule, BP 17171, 30207 Bagnols sur Ceze (France)

2008-07-01

The precipitation of a crystalline phase in glass is observed when one element exceeds its loading limit (i.e.: solubility limit). In this work we have studied the solubility of different actinides and surrogates (lanthanides and hafnium) in borosilicate glass used for the immobilization of the high-level nuclear waste (HLW glasses). The results obtained show an increase of the solubility limits of these elements with the processing temperature and the redox potential of the melt. The elements at the oxidation state (III) exhibit a higher solubility than the element at oxidation state (IV). In this framework, cerium is an interesting element because its oxidation state tunes from (IV) to (III) as a function of the processing conditions. It is shown that the solubility of cerium can be multiplied by a factor of 20 at 1100 C. degrees. In order to have a better understanding of the mechanisms that underline the evolution of the solubility, XAFS and NMR investigation has been undertaken. Trivalent elements present the characteristics of network-modified cations while tetravalent elements look like network-former cations.

Effect of MoO3 addition on phase separation in borosilicate glasses

International Nuclear Information System (INIS)

Wakabayashi, Hajimu; Yamanaka, Hiroshi; Terai, Ryohei; Fukumoto, Shinji.

1987-01-01

The effect of MoO 3 addition on the phase separation of sodium borosilicate glasses has been studied by electron microscopic observation and electrical conductivity measurement. With increasing MoO 3 addition, the immiscibility temperatures rose, the growth rate of phase separation increased, and the immiscibility region in the Na 2 O-B 2 O 3 -SiO 2 system extended. When MoO 3 more than 5 mol% was added to the glasses, two stages of phase separation appeared ; one is metastable immiscibility, which takes place below the liquidus temperature, and the other is stable immiscibility, occurring above the liquidus temperature. The present experiment confirmed that the conductivity measurement of molten glasses is usefull to detect phase separation in the glasses, giving some suggestions for preventing the separation of molybdates in nuclear waste glasses and also for controlling the micro-pore sizes of porous glasses. (author)

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

Structure, thermal stability and resistance under external irradiation of rare earths and molybdenum-rich alumino-borosilicate glasses

International Nuclear Information System (INIS)

Chouard, N.

2011-01-01

In France, the highly radioactive nuclear liquid wastes arising from spent nuclear fuel reprocessing (fission products + minor actinides (FPA)) are currently immobilized in an alumino-borosilicate glass called 'R7T7'. In the future, the opportunity of using new alumino-borosilicate glass compositions (HTC glasses) is considered in order to increase the waste loading in glasses and thus significantly decrease the number of glass canisters. However, the increase of the concentration of FPA could lead to the crystallization of rare-earth-rich phases (Ca 2 RE 8 (SiO 4 ) 6 O 2 ) or molybdenum-rich phases (CaMoO 4 , Na 2 MoO 4 ) during melt cooling, which can modify the confinement properties of the glass (chemical durability, self-irradiation resistance..), particularly if they can incorporate radionuclides α or β in their structure. This thesis can be divided into two parts: The first part deals with studying the relationship that can occur between the composition, the structure and the crystallization tendency of simplified seven oxides glasses, belonging to the SiO 2 -B 2 O 3 -Al 2 O 3 -Na 2 O-CaO-MoO 3 -Nd 2 O 3 system and derived from the composition of the HTC glass at 22,5 wt. % in FPA. The impact of the presence of platinoid elements (RuO 2 in our case) on the crystallization of the different phases is also studied. The second part deals with the effect of actinides α decays and more particularly of nuclear interactions essentially coming from recoil nuclei (simulated here by heavy ions external irradiations) on the behaviour under irradiation of an alumino-borosilicate glass containing apatite Ca 2 Nd 8 (SiO 4 ) 6 O 2 crystals, that can incorporate actinides in their structure. Two samples containing apatite crystals with different size are studied, in order to understand the impact of microstructure on the irradiation resistance of this kind of material. (author) [fr

Antagonist effects of calcium on borosilicate glass alteration

Energy Technology Data Exchange (ETDEWEB)

Mercado-Depierre, S. [CEA Marcoule, DTCD SPDE LCLT, 30207 Bagnols sur Cèze (France); Angeli, F., E-mail: frederic.angeli@cea.fr [CEA Marcoule, DTCD SPDE LCLT, 30207 Bagnols sur Cèze (France); Frizon, F. [CEA Marcoule, DTCD SECM LP2C, 30207 Bagnols sur Cèze (France); Gin, S. [CEA Marcoule, DTCD SPDE LCLT, 30207 Bagnols sur Cèze (France)

2013-10-15

Graphical abstract: Display Omitted -- Highlights: •Kinetic study of glass alteration is investigated in calcium-enriched solutions. •New insights into silicon–calcium interactions in glass/cement systems are proposed. •Glass alteration is controlled by pH, Ca concentration and reaction progress. •Evidence of antagonist effects according to the importance of these parameters. -- Abstract: Numerous studies have been conducted on glass and cement durability in contact with water, but very little work to date has focused directly on interactions between the two materials. These interactions are mostly controlled by silicon–calcium reactivity. However, the physical and chemical processes involved remain insufficiently understood to predict the evolution of coupled glass–cement systems used in several industrial applications. Results are reported from borosilicate glass alteration in calcium-rich solutions. Our data show that four distinct behaviors can be expected according to the relative importance of three key parameters: the pH, the reaction progress (short- or long-term alteration) and the calcium concentration. Glass alteration is thus controlled by specific mechanisms depending on the solution chemistry: calcium complexation at the glass surface, precipitation of calcium silicate hydrates (C–S–H) or calcium incorporation in the altered layer. These findings highlight the impact of silicon–calcium interactions on glass durability and open the way for a better understanding of glass–cement mixing in civil engineering applications as well as in nuclear waste storage.

Antagonist effects of calcium on borosilicate glass alteration

International Nuclear Information System (INIS)

Mercado-Depierre, S.; Angeli, F.; Frizon, F.; Gin, S.

2013-01-01

Graphical abstract: Display Omitted -- Highlights: •Kinetic study of glass alteration is investigated in calcium-enriched solutions. •New insights into silicon–calcium interactions in glass/cement systems are proposed. •Glass alteration is controlled by pH, Ca concentration and reaction progress. •Evidence of antagonist effects according to the importance of these parameters. -- Abstract: Numerous studies have been conducted on glass and cement durability in contact with water, but very little work to date has focused directly on interactions between the two materials. These interactions are mostly controlled by silicon–calcium reactivity. However, the physical and chemical processes involved remain insufficiently understood to predict the evolution of coupled glass–cement systems used in several industrial applications. Results are reported from borosilicate glass alteration in calcium-rich solutions. Our data show that four distinct behaviors can be expected according to the relative importance of three key parameters: the pH, the reaction progress (short- or long-term alteration) and the calcium concentration. Glass alteration is thus controlled by specific mechanisms depending on the solution chemistry: calcium complexation at the glass surface, precipitation of calcium silicate hydrates (C–S–H) or calcium incorporation in the altered layer. These findings highlight the impact of silicon–calcium interactions on glass durability and open the way for a better understanding of glass–cement mixing in civil engineering applications as well as in nuclear waste storage

Rutherford backscattering investigation of the corrosion of borosilicate glass

International Nuclear Information System (INIS)

Sales, B.C.; Boatner, L.A.; Naramoto, H.; White, C.W.

1981-10-01

The RBS spectra from Frit 21 borosilicate glasses doped with 5 wt % UO 2 , SrO, or Cs 2 O show that: during the initial stages of leaching (0 to 3 h) there is a substantial (300 to 500%) enhancement in the concentration of U, Sr, Ca, and Ti in the outer surface layer and that this enhancement is accompanied by a large depletion of Na, Si, and Cs; and upon further leaching under static conditions (24 h) the leached surface layer composition is indistinguishable from the unleached surface. Other borosilicate glasses such as PNL 76-68 may eventually show the same behavior if the final equilibrium pH value is greater than 9. The technique of Rutherford backscattering depth profile analysis can be a powerful tool for investigating the initial stages of glass corrosion

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

Infrared and x-ray photoelectron spectroscopic studies on sodium borosilicate glass interacted with thermally oxidized aluminides formed on alloy 690

International Nuclear Information System (INIS)

Yusufali, C.; Dutta, R.S.; Dey, G.K.; Kshirsagar, R.J.; Jagannath; Mishra, R.K.

2012-01-01

Thermally oxidized aluminides formed on Ni-Cr-Fe based superalloy 690 substrates were subjected to interaction with sodium borosilicate melt (used as matrices for immobilization of high-level radioactive liquid waste) at 1248 K for 192 hours. After the interaction, Fourier-transform infrared (FT-IR) spectroscopy analysis of glass samples indicated the incorporation of Al in the glass network. X-ray photoelectron spectroscopy (XPS) of glass specimens revealed modified glass structure. (author)

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

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

Fabrication of Low Noise Borosilicate Glass Nanopores for Single Molecule Sensing.

Directory of Open Access Journals (Sweden)

Jayesh A Bafna

Full Text Available We show low-cost fabrication and characterization of borosilicate glass nanopores for single molecule sensing. Nanopores with diameters of ~100 nm were fabricated in borosilicate glass capillaries using laser assisted glass puller. We further achieve controlled reduction and nanometer-size control in pore diameter by sculpting them under constant electron beam exposure. We successfully fabricate pore diameters down to 6 nm. We next show electrical characterization and low-noise behavior of these borosilicate nanopores and compare their taper geometries. We show, for the first time, a comprehensive characterization of glass nanopore conductance across six-orders of magnitude (1M-1μM of salt conditions, highlighting the role of buffer conditions. Finally, we demonstrate single molecule sensing capabilities of these devices with real-time translocation experiments of individual λ-DNA molecules. We observe distinct current blockage signatures of linear as well as folded DNA molecules as they undergo voltage-driven translocation through the glass nanopores. We find increased signal to noise for single molecule detection for higher trans-nanopore driving voltages. We propose these nanopores will expand the realm of applications for nanopore platform.

Assessment of methods for immobilizing reprocessed radioactive waste

Science.gov (United States)

Murthy, M. K.; Baranyi, A. D.

1980-05-01

Nuclear waste forms presently used for the disposal of high level wastes and other potential waste forms under development were studied. The following waste forms were considered: Borosilicate glass, high silica glass, glassceramics, supercalcine ceramics, synroc ceramics, borosilicate glass beads in a metal matrix, supercalcine and synroc ceramics in a metal matrix and coated ceramics. The best developed wasteform, both in terms of overall product quality and process development, is monolithic borosilicate glass. However, hydrothermal instability is a major concern. Borosilicate glass in metal matrix waste form has better properties than monolithic borosilicate glass waste form. The process was proven on a pilot scale. Hence, it is considered very close to monolithic glass in terms of overall development. The product qualities of the other waste forms are better than borosilicate glass. However, process development for these alternative waste forms is still in a conceptual stage.

Development of borosilicate glass compositions for the immobilisation of the UK's separated plutonium stocks

International Nuclear Information System (INIS)

Harrison, M. T.; Scales, C. R.

2008-01-01

The UK inventory of separated civil plutonium is expected to exceed 100 tonnes by 2010. Whilst the majority of this could be used in the manufacture of MOx (Mixed Oxide) fuel in future power generation scenarios, options for the disposal of surplus plutonium are currently being investigated by Nexia Solutions Ltd on behalf of the UK's Nuclear Decommissioning Authority (NDA). One of the options being considered is immobilisation in a durable glass matrix followed by long term storage and subsequent final repository disposal. A preliminary experimental survey assessed a selection of potential glass systems on the basis of Pu-surrogate (cerium) loading, durability, and ease of processing. Following this, a number of borosilicate compositions have been taken forward into a more detailed investigation in order to fully qualify their potential for Pu-immobilisation. The selected compositions are lanthanide borosilicate (LaBS), alkali tin silicate (ATS) and high-lanthanide alkali borosilicate (modified-MW). For this second series of experiments, hafnium was selected as the Pu surrogate, and a study of the potential waste loading as a function of temperature for the three selected compositions is described in this paper. Furthermore, several variations of the LaBS composition were fabricated in order to investigate the effect of total lanthanide content on melting temperature. The benchmark of 10 wt% HfO 2 incorporation is achievable for all three glasses with temperatures of 1200, 1300 and 1400 deg. C required for ATS, modified-MW and LaBS respectively. (authors)

Water leaching of borosilicate glasses: experiments, modeling and Monte Carlo simulations

International Nuclear Information System (INIS)

Ledieu, A.

2004-10-01

This work is concerned with the corrosion of borosilicate glasses with variable oxide contents. The originality of this study is the complementary use of experiments and numerical simulations. This study is expected to contribute to a better understanding of the corrosion of nuclear waste confinement glasses. First, the corrosion of glasses containing only silicon, boron and sodium oxides has been studied. The kinetics of leaching show that the rate of leaching and the final degree of corrosion sharply depend on the boron content through a percolation mechanism. For some glass contents and some conditions of leaching, the layer which appears at the glass surface stops the release of soluble species (boron and sodium). This altered layer (also called the gel layer) has been characterized with nuclear magnetic resonance (NMR) and small angle X-ray scattering (SAXS) techniques. Second, additional elements have been included in the glass composition. It appears that calcium, zirconium or aluminum oxides strongly modify the final degree of corrosion so that the percolation properties of the boron sub-network is no more a sufficient explanation to account for the behavior of these glasses. Meanwhile, we have developed a theoretical model, based on the dissolution and the reprecipitation of the silicon. Kinetic Monte Carlo simulations have been used in order to test several concepts such as the boron percolation, the local reactivity of weakly soluble elements and the restructuring of the gel layer. This model has been fully validated by comparison with the results on the three oxide glasses. Then, it has been used as a comprehensive tool to investigate the paradoxical behavior of the aluminum and zirconium glasses: although these elements slow down the corrosion kinetics, they lead to a deeper final degree of corrosion. The main contribution of this work is that the final degree of corrosion of borosilicate glasses results from the competition of two opposite mechanisms

Aqueous corrosion of borosilicate glasses: experiments, modeling and Monte-Carlo simulations

International Nuclear Information System (INIS)

Ledieu, A.

2004-10-01

This work is concerned with the corrosion of borosilicate glasses with variable oxide contents. The originality of this study is the complementary use of experiments and numerical simulations. This study is expected to contribute to a better understanding of the corrosion of nuclear waste confinement glasses. First, the corrosion of glasses containing only silicon, boron and sodium oxides has been studied. The kinetics of leaching show that the rate of leaching and the final degree of corrosion sharply depend on the boron content through a percolation mechanism. For some glass contents and some conditions of leaching, the layer which appears at the glass surface stops the release of soluble species (boron and sodium). This altered layer (also called the gel layer) has been characterized with nuclear magnetic resonance (NMR) and small angle X-ray scattering (SAXS) techniques. Second, additional elements have been included in the glass composition. It appears that calcium, zirconium or aluminum oxides strongly modify the final degree of corrosion so that the percolation properties of the boron sub-network is no more a sufficient explanation to account for the behavior of these glasses. Meanwhile, we have developed a theoretical model, based on the dissolution and the reprecipitation of the silicon. Kinetic Monte Carlo simulations have been used in order to test several concepts such as the boron percolation, the local reactivity of weakly soluble elements and the restructuring of the gel layer. This model has been fully validated by comparison with the results on the three oxide glasses. Then, it has been used as a comprehensive tool to investigate the paradoxical behavior of the aluminum and zirconium glasses: although these elements slow down the corrosion kinetics, they lead to a deeper final degree of corrosion. The main contribution of this work is that the final degree of corrosion of borosilicate glasses results from the competition of two opposite mechanisms

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)

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.

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

Understanding the structural drivers governing glass-water interactions in borosilicate based model bioactive glasses.

Science.gov (United States)

Stone-Weiss, Nicholas; Pierce, Eric M; Youngman, Randall E; Gulbiten, Ozgur; Smith, Nicholas J; Du, Jincheng; Goel, Ashutosh

2018-01-01

The past decade has witnessed a significant upsurge in the development of borate and borosilicate based resorbable bioactive glasses owing to their faster degradation rate in comparison to their silicate counterparts. However, due to our lack of understanding about the fundamental science governing the aqueous corrosion of these glasses, most of the borate/borosilicate based bioactive glasses reported in the literature have been designed by "trial-and-error" approach. With an ever-increasing demand for their application in treating a broad spectrum of non-skeletal health problems, it is becoming increasingly difficult to design advanced glass formulations using the same conventional approach. Therefore, a paradigm shift from the "trial-and-error" approach to "materials-by-design" approach is required to develop new-generations of bioactive glasses with controlled release of functional ions tailored for specific patients and disease states, whereby material functions and properties can be predicted from first principles. Realizing this goal, however, requires a thorough understanding of the complex sequence of reactions that control the dissolution kinetics of bioactive glasses and the structural drivers that govern them. While there is a considerable amount of literature published on chemical dissolution behavior and apatite-forming ability of potentially bioactive glasses, the majority of this literature has been produced on silicate glass chemistries using different experimental and measurement protocols. It follows that inter-comparison of different datasets reveals inconsistencies between experimental groups. There are also some major experimental challenges or choices that need to be carefully navigated to unearth the mechanisms governing the chemical degradation behavior and kinetics of boron-containing bioactive glasses, and to accurately determine the composition-structure-property relationships. In order to address these challenges, a simplified

Investigation on neptunium in a borosilicate glass

International Nuclear Information System (INIS)

Poirot, I.

1988-03-01

The oxidization state and coordination of neptunium, introduced as dopant in borosilicate glasses were studied through optical, Mossbauer spectroscopies and magnetic measurements. The neptunium oxide, introduced previously as NpO 2 is reduced during the melting process of the glass. This leads to an equilibrium in which the ratio of Np 4+ to Np 3+ valences depends on experimental conditions. Spectroscopic analysises conduct to postulate the presence of different sites for each of the oxydation states of neptunium [fr

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)

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)

Conversion of plutonium-containing materials into borosilicate glass using the glass material oxidation and dissolution system

International Nuclear Information System (INIS)

Forsberg, C.W.; Beahm, E.C.; Parker, G.W.

1996-01-01

The end of the cold war has resulted in excess plutonium-containing materials (PCMs) in multiple chemical forms. Major problems are associated with the long-term management of these materials: safeguards and nonproliferation issues; health, environment, and safety concerns; waste management requirements; and high storage costs. These issues can be addressed by conversion of the PCMs to glass: however, conventional glass processes require oxide-like feed materials. Conversion of PCMs to oxide-like materials followed by vitrification is a complex and expensive process. A new vitrification process has been invented, the Glass Material Oxidation and Dissolution System (GMODS) to allow direct conversion of PCMs to glass. GMODS directly converts metals, ceramics, and amorphous solids to glass; oxidizes organics with the residue converted to glass; and converts chlorides to borosilicate glass and a secondary sodium chloride stream. Laboratory work has demonstrated the conversion of cerium (a plutonium surrogate), uranium (a plutonium surrogate), Zircaloy, stainless steel, multiple oxides, and other materials to glass. Equipment options have been identified for processing rates between 1 and 100,000 t/y. Significant work, including a pilot plant, is required to develop GMODS for applications at an industrial scale

Mechanical behavior of cellular borosilicate glass with pressurized Ar-filled closed pores

International Nuclear Information System (INIS)

Wang Bo; Matsumaru, Koji; Yang Jianfeng; Fu Zhengyi; Ishizaki, Kozo

2012-01-01

High strength borosilicate foams were fabricated by melting glass powder under high-pressure argon gas and subsequent heat treatment of the glass bulk at atmospheric pressure. In the first step, borosilicate glass powder was melted at 1100 °C for 1 h by capsule-free hot isostatic pressing (HIPing) under a high gas pressure of 10–70 MPa. Pressurized Ar-filled spherical pores were introduced into the glass, and argon atoms were dissolved in the glass network structure. The expansion of argon-filled pores and the release of the dissolved Ar gas resulted in the formation of pressurized Ar-filled closed pores by isothermal heat treatment at 800 °C for 10 min. A high porosity of up to 80% with a bimodal distribution of micro-size cells was obtained for the resultant cellular borosilicate glass. By increasing the total gas pressure from 10 to 70 MPa, the compressive strength and the Young’s modulus were increased considerably from 15 to 52 MPa and from 4.1 to 12.6 GPa, respectively, which can be substantially attributed to the high collapse stress from the high enclosed gas pressure. The cellular glass with a high porosity showed a large failure strain under uniaxial compression.

Er3+-Al2O3 nanoparticles doping of borosilicate glass

International Nuclear Information System (INIS)

Massera, Jonathan; Petit, Laeticia; Hupa, Leena; Hupa, Mikko; Koponen, Joona; Glorieux, Benoit

2015-01-01

Novel borosilicate glasses were developed by adding in the glass batch Er 3+ -Al 2 O 3 nanoparticles synthetized by using a soft chemical method. A similar nanoparticle doping with modified chemical vapour deposition (MCVD) process was developed to increase the efficiency of the amplifying silica fibre in comparison to using MCVD and solution doping. It was shown that with the melt quench technique, a Er 3+ -Al 2 O 3 nanoparticle doping neither leads to an increase in the Er 3+ luminescence properties nor allows one to control the rare-earth chemical environment in a borosilicate glass. The site of Er 3+ in the Er 3+ -Al 2 O 3 nanoparticle containing glass seems to be similar as in glasses with the same composition prepared using standard raw materials. We suspect the Er 3+ ions to diffuse from the nanoparticles into the glass matrix. There was no clear evidence of the presence of Al 2 O 3 nanoparticles in the glasses after melting. (author)

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)

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)

Design and test of the borosilicate glass burnable poison rod for Qinshan nuclear power plant core

International Nuclear Information System (INIS)

Huang Jinhua; Sun Hanhong

1988-08-01

Material for the burnable poison of Qinshan Nuclear Power Plant core is GG-17 borosilicate glass. The chemical composition and physico-chemical properties of GG-17 is very close to Pyrex-7740 glass used by Westinghouse. It is expected from the results of the experiments that the borosilicate glass burnable poison rod can be successfully used in Qinshan Nuclear Power Plant due to good physical, mechanical, corrosion-resistant and irradiaton properties for both GG-17 glass and cold-worked stainless steel cladding. Change of material for burnable poison from boron-bearing stainless steel to borosilicate glass will bring about much more economic benefit to Qinshan Naclear Power Plant

Immobilization of Technetium Waste from Pyro-processing Using Tellurite Glasses

Energy Technology Data Exchange (ETDEWEB)

Heo, Jong; Pyo, Jae-Young; Lee, Cheong-Won [POSTECH, Pohang (Korea, Republic of); Yang, Jae-Hwan; Park, Hwan-Seo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Vitrification of Tc wastes has been challenging because of the low solubility in the silicate glass and high volatility in the melting process. In previous studies, the measured solubility of Tc and Re was ⁓ 3000 ppm at 1000 .deg. C in low activity waste (LAW) glass. And retention of Tc has been reported within 12 - 77% during the borosilicate vitrification process. Tellurite glasses have been studied for halide waste immobilization due to low melting temperatures (Tm= 600-800 .deg. C) and flexibility of network with foreign ions. Tellurite glasses offered higher halide retention than borosilicate glasses. The structure of pure tellurite (TeO{sub 2}) consists of TeO{sub 4} trigonal bipyramids (tbp), but TeO{sub 4} units are converted to TeO{sub 3} trigonal pyramids (tp) having non-bridging oxygen (NBO) as the modifiers added. Objectives of this study are to investigate the tellurite glasses for Tc immobilization using Re as a surrogate. Retention and waste loading of Re were analyzed during the vitrification process of tellurite glass. We investigated local structures of Re ions in glasses by Raman and X-ray absorption spectroscopies. The tellurite glass was investigated to immobilize the Ca(TcO{sub 4}){sub 2}, surrogated by Ca(ReO{sub 4}){sub 2}. The average of Re retention in tellurite glass was 86%. The 7-day PCT results were satisfied with U.S requirement up to 9 mass% of Ca(ReO{sub 4}){sub 2} content. Re in the tellurite glass exists +7 oxidation state and was coordinated with 4 oxygen.

Photon emission induced by brittle fracture of borosilicate glasses

Energy Technology Data Exchange (ETDEWEB)

Shiota, Tadashi, E-mail: tshiota@ceram.titech.ac.jp [Department of Metallurgy and Ceramic Science, Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro-ku, Tokyo 152-8550 (Japan); Sato, Yoshitaka [Department of Metallurgy and Ceramic Science, Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro-ku, Tokyo 152-8550 (Japan); Kishi, Tetsuo [Department of Chemistry and Materials Science, Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro-ku, Tokyo 152-8550 (Japan); Yasuda, Kouichi [Department of Metallurgy and Ceramic Science, Tokyo Institute of Technology, 2-12-1 Ookayama, Meguro-ku, Tokyo 152-8550 (Japan)

2016-05-15

Photon emission (PE) at wavelength ranges of 430–490 nm (B-PE), 500–600 nm (G-PE) and 610–680 nm (R-PE) caused by brittle fracture was simultaneously measured in the nanosecond-to-microsecond and millisecond time domains for two types of borosilicate glasses: Pyrex-type Tempax glass and BK7 glass. The results were compared to those for silica and soda lime glasses. The time dependence of the PE of Tempax glass was similar to that of silica glass, while the PE intensity was lower. Because Tempax glass contains both silica-rich and borate-rich amorphous phases, the PE must be mainly produced by the fracture of the silica-rich phase. Moreover, the proportion of B-PE of Tempax glass was higher than that of silica glass. This suggests that the measured B-PE might also include very weak PE caused by the fracture of the borate-rich phase. The PE time dependence of BK7 glass was similar to that of soda lime glass, which was different from the case for Tempax glass. The PE intensity of BK7 glass was slightly higher than that of soda lime glass, but much lower than that of Tempax glass. The result indicates that non-bridging oxygen in the glasses affects crack propagation behavior and reduces the PE. - Highlights: • Photon emission (PE) upon brittle fracture of borosilicate glasses was measured. • Pyrex-type Tempax and BK7 glasses showed different PE characteristics. • The rupture of Si–O bonds produces much stronger PE than that of B–O bonds. • Non-bridging oxygen in glass affects crack propagation behavior and reduces the PE.

High-level nuclear waste borosilicate glass: A compendium of characteristics

International Nuclear Information System (INIS)

Cunnane, J.C.; Bates, J.K.; Ebert, W.L.; Feng, X.; Mazer, J.J.; Wronkiewicz, D.J.; Sproull, J.; Bourcier, W.L.; McGrail, B.P.

1992-01-01

With the imminent startup, in the United States, of facilities for vitrification of high-level nuclear waste, a document has been prepared that compiles the scientific basis for understanding the alteration of the waste glass products under the range of service conditions to which they may be exposed during storage, transportation, and eventual geologic disposal. A summary of selected parts of the content of this document is provided. Waste glass alterations in a geologic repository may include corrosion of the glass network due to groundwater and/or water vapor contact. Experimental testing results are described and interpreted in terms of the underlying chemical reactions and physical processes involved. The status of mechanistic modeling, which can be used for long-term predictions, is described and the remaining uncertainties associated with long-term simulations are summarized

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)

Determination of the free enthalpies of formation of borosilicate glasses

International Nuclear Information System (INIS)

Linard, Y.

2000-01-01

This work contributes to the study of the thermochemical properties of nuclear waste glasses. Results are used to discuss mechanisms and parameters integrated in alteration models of conditioning materials. Glass is a disordered material defined thermodynamically as a non-equilibrium state. Taking into account one order parameter to characterise its configurational state, the metastable equilibrium for the glass was considered and the main thermochemical properties were determined. Calorimetric techniques were used to measure heat capacities and formation enthalpies of borosilicate glasses (from 3 to 8 constitutive oxides). Formation Entropies were measured too, using the entropy theory of relaxation processes proposed by Adam and Gibbs (1965). The configurational entropy contribution were determined from viscosity measurements. This set of data has allowed the calculation of Gibb's free energies of dissolution of glasses in pure water. By comparison with leaching experiments, it has been demonstrated that the decreasing of the dissolution rate at high reaction progress cannot be associated to the approach of an equilibrium between the sound glass and the aqueous solution. The composition changes of the reaction area at the glass surface need to be considered too. To achieve a complete description of the thermodynamic stability, the equilibrium between hydrated de-alkalinized glass and/or the gel layer with the aqueous solution should also be evaluated. (author)

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1993-06-01

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

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

Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

Science.gov (United States)

Boatner, Lynn A.; Sales, Brian C.

1989-01-01

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

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

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

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

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

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

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

Silicate glasses

International Nuclear Information System (INIS)

Lutze, W.

1988-01-01

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

A kinetic model for borosilicate glass dissolution based on the dissolution affinity of a surface alteration layer

International Nuclear Information System (INIS)

Bourcier, W.L.; Peiffer, D.W.; Knauss, K.G.; McKeegan, K.D.; Smith, D.K.

1989-11-01

A kinetic model for the dissolution of borosilicate glass is used to predict the dissolution rate of a nuclear waste glass. In the model, the glass dissolution rate is controlled by the rate of dissolution of an alkali-depleted amorphous surface (gel) layer. Our model predicts that all components concentrated in the surface layer, affect glass dissolution rates. The good agreement between predicted and observed elemental dissolution rates suggests that the dissolution rate of the gel layer limits the overall rate of glass dissolution. The model predicts that the long-term rate of glass dissolution will depend mainly on ion concentrations in solution, and therefore on the secondary phases which precipitate and control ion concentrations. 10 refs., 5 figs., 1 tab

Effect of ThO2 concentration on thermo-physical properties of barium borosilicate glass

International Nuclear Information System (INIS)

Mishra, R.K.; Munshi, A.K.; Kaushik, C.P.; Raj, Kanwar; Shobha, M.; Srikhande, V.K.; Kothiyal, G.P.; Tyagi, A.K.

2006-01-01

An advanced heavy water reactor is being developed at Bhabha Atomic Research Centre (BARC) India with an aim of utilizing thorium for power generation. The high level radioactive liquid waste (HLW) generated from reprocessing of Th-based spent fuel is expected to contain Th as one of the main constituents other than fission products, corrosion products, actinides and added chemicals. Barium borosilicate (BBS) glass is being used for vitrification of the HLW generated at the reprocessing plant at BARC, Trombay, Mumbai

Silicate glasses. Chapter 1

International Nuclear Information System (INIS)

Lutze, W.

1988-01-01

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

IR study of Pb–Sr titanate borosilicate glasses

Indian Academy of Sciences (India)

Administrator

IR study of Pb–Sr titanate borosilicate glasses. C R GAUTAM*, DEVENDRA KUMAR. † and OM PARKASH. †. Department of Physics, University of Lucknow, Lucknow 226 007, India. †. Department of Ceramic Engineering, Institute of Technology, Banaras Hindu University, Varanasi 221 005, India. MS received 3 January ...

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

Sulphate Incorporation in Borosilicate Glasses and Melts: a Kinetic Approach

Energy Technology Data Exchange (ETDEWEB)

Lenoir, M. [CEA, DEN, Laboratoire d' etude et de Developpement de Matrices de Conditionnement, Centre de Marcoule, 30207 Bagnols-sur-Ceze (France); Physique des Mineraux et Magmas, UMR 7047, CNRS- Institut de Physique du Globe de Paris, 7 place Jussieu, 75252 Paris 05 (France); Grandjean, A. [Institut de Chimie Separative de Marcoule, UMR 5257, Laboratoire des Nanomateriaux Autoreparants, Marcoule, 30207 Bagnols-sur-Ceze (France); Dussossoy, J.L. [CEA, DEN, Laboratoire d' etude et de Developpement de Matrices de Conditionnement, Centre de Marcoule, 30207 Bagnols-sur-Ceze (France); Neuville, D.R. [Physique des Mineraux et Magmas, UMR 7047, CNRS- Institut de Physique du Globe de Paris, 7 place Jussieu, 75252 Paris 05 (France)

2008-07-01

The kinetics of sulphate departure in a sodium borosilicate melt were studied using in situ Raman spectroscopy. This technique allows the quantification of the amount of sulphate dissolved in a borosilicate glass as a function of heating time by comparison with measurements obtained by microprobe wavelength dispersive spectrometry. To quantify the sulphate content obtained with Raman spectroscopy, the integrated intensity of the sulphate band at 990 cm{sup -1} was scaled to the sum of the integrated bands between 800 and 1200 cm{sup -1}, bands that are assigned to Qn silica units on the basis of previous literature. Calibration curves were then determined for two different samples. An evaluation of the kinetics of departure of sulphate could thus be made as a function of the viscosity of the borosilicate glass, showing that the kinetics were controlled by the diffusion of sulphate and its volatilization from the melt. This experimental method allows in situ measurements of sulphate content at high temperature which cannot be obtained by any other simple technique. (authors)

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

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

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.

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

International Nuclear Information System (INIS)

Willis, T.A.

1980-01-01

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

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

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.

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.

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

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

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

Behaviour of ruthenium dioxide particles in borosilicate glasses and melts

Energy Technology Data Exchange (ETDEWEB)

Pflieger, Rachel [DEN/DTCD-SCDV/CEA Valrho, Centre de Marcoule, BP17171, 30207 Bagnols-sur-Ceze (France); Institut de Chimie Separative de Marcoule, UMR5257, Centre de Marcoule, BP17171, 30207 Bagnols-sur-Ceze Cedex (France)], E-mail: rachel_pflieger@yahoo.fr; Lefebvre, Leila [DEN/DTCD-SCDV/CEA Valrho, Centre de Marcoule, BP17171, 30207 Bagnols-sur-Ceze (France); Malki, Mohammed [CNRS/CEMHTI-1D Av. de la Recherche Scientifique, 45701 Orleans cedex 2 (France); Polytech Orleans, Universite d' Orleans, 8 rue Leonard de Vinci, 45072 Orleans cedex 2 (France); Allix, Mathieu [CNRS/CEMHTI-1D Av. de la Recherche Scientifique, 45701 Orleans cedex 2 (France); Grandjean, Agnes [DEN/DTCD-SCDV/CEA Valrho, Centre de Marcoule, BP17171, 30207 Bagnols-sur-Ceze (France); Institut de Chimie Separative de Marcoule, UMR5257, Centre de Marcoule, BP17171, 30207 Bagnols-sur-Ceze Cedex (France)

2009-06-01

Ruthenium-glass systems are formed during the vitrification of nuclear waste. They are also widely used in micro-electronics because of their unique electrical properties. However, the interaction of this element with the glass matrix remains poorly understood. This work focuses on a RuO{sub 2} particles-nuclear alumino-borosilicate glass system in which the electrical conductivity is known to vary considerably with the RuO{sub 2} content and to become electronic above about 0.5-0.7 vol.% RuO{sub 2} [R. Pflieger, M. Malki, Y. Guari, J. Larionova, A. Grandjean, J. Am. Ceram. Soc., accepted for publication]. Some RuO{sub 2} segregation was observed in SEM/TEM investigations but no continuous chain of RuO{sub 2} particles could be seen. Electron relays between the particles are then necessary for a low-rate percolation, such as the nanoclusters suggested by Adachi et al. [K. Adachi, S. Iida, K. Hayashi, J. Mater. Res. 9 (7) (1994) 1866; K. Adachi, H. Kuno, J. Am. Ceram. Soc. 83 (10) (2000) 2441], which could consist in dissolved ruthenium. Indeed, several observations made here clearly indicate the presence of dissolved ruthenium in the glass matrix, like the modification of the glass density in presence of RuO{sub 2} particles or the diffusion-limited growth of RuO{sub 2} particles in the melt.

Crystalline phase, microstructure, and aqueous stability of zirconolite-barium borosilicate glass-ceramics for immobilization of simulated sulfate bearing high-level liquid waste

Science.gov (United States)

Wu, Lang; Xiao, Jizong; Wang, Xin; Teng, Yuancheng; Li, Yuxiang; Liao, Qilong

2018-01-01

The crystalline phase, microstructure, and aqueous stability of zirconolite-barium borosilicate glass-ceramics with different content (0-30 wt %) of simulated sulfate bearing high-level liquid waste (HLLW) were evaluated. The sulfate phase segregation in vitrification process was also investigated. The results show that the glass-ceramics with 0-20 wt% of HLLW possess mainly zirconolite phase along with a small amount baddeleyite phase. The amount of perovskite crystals increases while the amount of zirconolite crystals decreases when the HLLW content increases from 20 to 30 wt%. For the samples with 20-30 wt% HLLW, yellow phase was observed during the vitrification process and it disappeared after melting at 1150 °C for 2 h. The viscosity of the sample with 16 wt% HLLW (HLLW-16) is about 27 dPa·s at 1150 °C. The addition of a certain amount (≤20 wt %) of HLLW has no significant change on the aqueous stability of glass-ceramic waste forms. After 28 days, the 90 °C PCT-type normalized leaching rates of Na, B, Si, and La of the sample HLLW-16 are 7.23 × 10-3, 1.57 × 10-3, 8.06 × 10-4, and 1.23 × 10-4 g·m-2·d-1, respectively.

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

Remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

International Nuclear Information System (INIS)

Callan, J.E.

1986-01-01

The Defense Waste Processing Facility is being constructed at the Savannah River Plant at a cost of $870 million to immobilize the defense high-level radioactive waste. This radioactive waste is being added to borosilicate glass for later disposal in a federal repository. The borosilicate glass is poured into stainless steel canisters for storage. These canisters must be handled remotely because of their high radioactivity, up to 5000 R/h. After the glass has been poured into the canister which will be temporarily sealed, it is transferred to a decontamination cell and decontaminated. The canister is then transferred to the weld cell where a permanent cap is welded into place. The canisters must then be transported from the processing building to a storage vault on the plant until the federal repository is available. A shielded canister transporter (SCT) has been designed and constructed for this purpose. The design of the SCT vehicle allows the safe transport of a highly radioactive canister containing borosilicate glass weighing 2300 kg with a radiation level up to 5000 R/h from one building to another. The design provides shielding for the operator in the cab of the vehicle to be below 0.5 rem/h. The SCT may also be used to load the final shipping cask when the federal repository is ready to receive the canisters

An assessment of methods for immobilizing reprocessed radioactive waste

International Nuclear Information System (INIS)

Murthy, M.K.; Baranyi, A.D.

1980-05-01

Nuclear waste forms presently used for the disposal of high-level wastes and other potential waste forms under development were studied using information available in the literature and by visits to the laboratories. The following waste forms were considered: Borosilicate glass, high-silica glass, glassceramics, supercalcine ceramics, synroc ceramics, borosilicate glass beads in a metal matrix, supercalcine and synroc ceramics in a metal matrix and coated ceramics. The following conclusions have been reached: To date the best developed wasteform, both in terms of overall product quality and process development, is monolithic borosilicate glass. However, hydrothermal instability is a major concern. Borosilicate glass in metal matrix waste form has better properties than monolithic borosilicate glass waste form. The process has been proven on a pilot scale. Hence, it is considered very close to monolithic glass in terms of overall development. The product qualities of the other waste forms are better than borosilicate glass. However, process development for these alternative waste forms is still in a conceptual stage. The technological basis for processing ceramic waste forms exists in a well developed state. Nevertheless, adaptation of the technology to continuous hot-cell operation, although feasible, has not been demonstrated. In view of the product potential of ceramic waste forms it is felt that their development should be given emphasis at this time. (auth)

Aqueous corrosion of borosilicate glasses: experiments, modeling and Monte-Carlo simulations; Alteration par l'eau des verres borosilicates: experiences, modelisation et simulations Monte-Carlo

Energy Technology Data Exchange (ETDEWEB)

Ledieu, A

2004-10-01

This work is concerned with the corrosion of borosilicate glasses with variable oxide contents. The originality of this study is the complementary use of experiments and numerical simulations. This study is expected to contribute to a better understanding of the corrosion of nuclear waste confinement glasses. First, the corrosion of glasses containing only silicon, boron and sodium oxides has been studied. The kinetics of leaching show that the rate of leaching and the final degree of corrosion sharply depend on the boron content through a percolation mechanism. For some glass contents and some conditions of leaching, the layer which appears at the glass surface stops the release of soluble species (boron and sodium). This altered layer (also called the gel layer) has been characterized with nuclear magnetic resonance (NMR) and small angle X-ray scattering (SAXS) techniques. Second, additional elements have been included in the glass composition. It appears that calcium, zirconium or aluminum oxides strongly modify the final degree of corrosion so that the percolation properties of the boron sub-network is no more a sufficient explanation to account for the behavior of these glasses. Meanwhile, we have developed a theoretical model, based on the dissolution and the reprecipitation of the silicon. Kinetic Monte Carlo simulations have been used in order to test several concepts such as the boron percolation, the local reactivity of weakly soluble elements and the restructuring of the gel layer. This model has been fully validated by comparison with the results on the three oxide glasses. Then, it has been used as a comprehensive tool to investigate the paradoxical behavior of the aluminum and zirconium glasses: although these elements slow down the corrosion kinetics, they lead to a deeper final degree of corrosion. The main contribution of this work is that the final degree of corrosion of borosilicate glasses results from the competition of two opposite mechanisms

Water leaching of borosilicate glasses: experiments, modeling and Monte Carlo simulations; Alteration par l'eau des verres borosilicates: experiences, modelisation et simulations Monte Carlo

Energy Technology Data Exchange (ETDEWEB)

Ledieu, A

2004-10-15

This work is concerned with the corrosion of borosilicate glasses with variable oxide contents. The originality of this study is the complementary use of experiments and numerical simulations. This study is expected to contribute to a better understanding of the corrosion of nuclear waste confinement glasses. First, the corrosion of glasses containing only silicon, boron and sodium oxides has been studied. The kinetics of leaching show that the rate of leaching and the final degree of corrosion sharply depend on the boron content through a percolation mechanism. For some glass contents and some conditions of leaching, the layer which appears at the glass surface stops the release of soluble species (boron and sodium). This altered layer (also called the gel layer) has been characterized with nuclear magnetic resonance (NMR) and small angle X-ray scattering (SAXS) techniques. Second, additional elements have been included in the glass composition. It appears that calcium, zirconium or aluminum oxides strongly modify the final degree of corrosion so that the percolation properties of the boron sub-network is no more a sufficient explanation to account for the behavior of these glasses. Meanwhile, we have developed a theoretical model, based on the dissolution and the reprecipitation of the silicon. Kinetic Monte Carlo simulations have been used in order to test several concepts such as the boron percolation, the local reactivity of weakly soluble elements and the restructuring of the gel layer. This model has been fully validated by comparison with the results on the three oxide glasses. Then, it has been used as a comprehensive tool to investigate the paradoxical behavior of the aluminum and zirconium glasses: although these elements slow down the corrosion kinetics, they lead to a deeper final degree of corrosion. The main contribution of this work is that the final degree of corrosion of borosilicate glasses results from the competition of two opposite mechanisms

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.

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.

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

International Nuclear Information System (INIS)

Glasser, F.P.

1985-01-01

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

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

Test Summary Report Vitrification Demonstration of an Optimized Hanford C-106/AY-102 Waste-Glass Formulation

International Nuclear Information System (INIS)

Goles, Ronald W.; Buchmiller, William C.; Hymas, Charles R.; MacIsaac, Brett D.

2002-01-01

In order to further the goal of optimizing Hanford?s HLW borosilicate flowsheet, a glass formulation effort was launched to develop an advanced high-capacity waste form exhibiting acceptable leach and crystal formation characteristics. A simulated C-106/AY-102 waste envelop inclusive of LAW pretreatment products was chosen as the subject of these nonradioactive optimization efforts. To evaluate this optimized borosilicate waste formulation under continuous dynamic vitrification conditions, a research-scale Joule-heated ceramic melter was used to demonstrate the advanced waste form?s flowsheet. The main objectives of this melter test was to evaluate (1) the processing characteristics of the newly formulated C-106/AY-102 surrogate melter-feed stream, (2) the effectiveness of sucrose as a glass-oxidation-state modifier, and (3) the impact of this reductant upon processing rates

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

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

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Porous glass with high silica content for nuclear waste storage : preparation, characterization and leaching

International Nuclear Information System (INIS)

Aegerter, M.A.; Santos, D.I. dos; Ventura, P.C.S.

1984-01-01

Aqueous solutions simulating radioactive nuclear wastes (like Savanah River Laboratory) were incorporated in porous glass matrix with high silica content prepared by decomposition of borosilicate glass like Na 2 O - B 2 O 3 - SiO 2 . After sintering, the samples were submitted, during 28 days, to standard leaching tests MCC1, MCC5 (Soxhlet) and stagnating. The total weight loss, ph, as well as the integral and differential leaching rates and the accumulated concentrations in the leach of Si, Na, B, Ca, Mn, Al, Fe and Ni. The results are compared with the results from reference borosilicate glass, made by fusion, ceramic, synroc, concrets, etc... (E.G.) [pt

Effects of magnesium minerals representative of the Callovian-Oxfordian clay-stone on borosilicate glass alteration

International Nuclear Information System (INIS)

Debure, M.

2012-01-01

Borosilicate glasses dissolution has been studied in presence of magnesium minerals. Those minerals (dolomite, illite, smectite...) belong to the Callovo-Oxfordian (COx) clay-stone layer, studied in France as a potential site for nuclear waste disposal. Such minerals contain magnesium, an element able to sustain glass alteration when it is available in solution. In the confined media of the wastes disposal, the solids reactivity controls the solution composition and can be the driving force of nuclear glass alteration. Experiments show that magnesium carbonates (hydro-magnesite and dolomite) increase in the glass alteration: the precipitation of magnesium silicates consumes silicon which slows down the formation of the glass passivating layer. The lower the magnesium mineral solubility, the lower the glass alteration. The purified clay phases (illite, smectite...) from the COx layer increase the glass alteration. Half the magnesium was replaced by sodium during the purification process. In such conditions, the effect of clay phases on glass alteration is in part due to the acidic pH-buffering effect of the clay fraction. The GRAAL model implemented in the geochemical transport code HYTEC has confirmed and quantified the mechanisms put in evidence in the experiments. Cells diffusion experiments where the two solids were separated by an inert diffusion barrier allow to valid reactive transport modelling. Such experiments are more representative of the glass package which will be separated from the COx by corrosion products. They show that glass alteration rate is reduced when solids are not close. (author) [fr

Determination of iron redox ratio in borosilicate glasses and melts from Raman spectra

Energy Technology Data Exchange (ETDEWEB)

Cochain, B. [SCDV-Laboratoire d' Etudes de Base sur les Verres, CEA Valrho, Centre de Marcoule, 30207 Bagnols-sur-ceze (France); Physique des Mineraux et des Magmas, CNRS-IPGP, 4 place Jussieu, 75252 Paris Cedex05 (France); Neuville, D.R.; Richet, P. [Physique des Mineraux et des Magmas, CNRS-IPGP, 4 place Jussieu, 75252 Paris Cedex05 (France); Henderson, G.S. [Dept of Geology, University of Toronto, 22 Russell Street, Toronto (Canada); Pinet, O. [SCDV-Laboratoire d' Etudes de Base sur les Verres, CEA Valrho, Centre de Marcoule, 30207 Bagnols-sur-ceze (France)

2008-07-01

A method is presented to determine the redox ratio of iron in borosilicate glasses and melts relevant to nuclear waste storage from an analysis of Raman spectra recorded at room or high temperature. The basis of this method is the strong variation of the spectral feature observed between 800 and 1200 cm{sup -1}, in which it is possible to assign a band to vibrational modes involving ferric iron in tetrahedral coordination whose intensity increases with iron content and iron oxidation. After baseline correction and normalization, fits to the Raman spectra made with Gaussian bands enable us to determine the proportion of ferric iron provided the redox ratio is known independently for at least two redox states for a given glass composition. This method is particularly useful for in situ determinations of the kinetics and mechanisms of redox reactions. (authors)

Determination of iron redox ratio in borosilicate glasses and melts from Raman spectra

International Nuclear Information System (INIS)

Cochain, B.; Neuville, D.R.; Richet, P.; Henderson, G.S.; Pinet, O.

2008-01-01

A method is presented to determine the redox ratio of iron in borosilicate glasses and melts relevant to nuclear waste storage from an analysis of Raman spectra recorded at room or high temperature. The basis of this method is the strong variation of the spectral feature observed between 800 and 1200 cm -1 , in which it is possible to assign a band to vibrational modes involving ferric iron in tetrahedral coordination whose intensity increases with iron content and iron oxidation. After baseline correction and normalization, fits to the Raman spectra made with Gaussian bands enable us to determine the proportion of ferric iron provided the redox ratio is known independently for at least two redox states for a given glass composition. This method is particularly useful for in situ determinations of the kinetics and mechanisms of redox reactions. (authors)

Coordination Environments of Highly Charged Cations (Ti, Cr, and Light REE's) in Borosilicate Glass/Melts to 1120C

Energy Technology Data Exchange (ETDEWEB)

Farges, Francois; /Museum Natl. Hist. Natur. /Stanford U., Geo. Environ. Sci.; Brown, Gordon E., Jr.; /Stanford U., Geo. Environ Sci. /SLAC, SSRL

2007-01-02

The local environments around Ti, Cr, and several light rare-earth elements (La, Ce, and Nd) were investigated by in-situ XANES spectroscopy in a number of complex borosilicate glasses and melts (to 1120 C) that are used for nuclear waste storage. Examination of the high-resolution XANES spectra at the Ti K-edge shows that the average coordination of Ti changes from {approx}5 to {approx}4.5. Cr is dominantly trivalent in the melts studied. However, its average coordination is probably lower in the melt (tetrahedral ?) as revealed by the more intense Cr-K pre-edge feature. Ce also changes its average valence from dominantly +4 to +3.5 upon glass melting. These changes are reversible at T{sub g}, the glass transition temperature ({approx}500-550 C for these glasses). In contrast, the local environments of Nd, Pr, and La are unaffected by melting. Therefore, structural reorganization of these borosilicate glass/melts above T{sub g} is variable, not only in terms of valence (as for Ce) but also speciation (Ti and Cr). Both the ability of B to adopt various coordination geometries (triangular and tetrahedral) and the chemical complexity of the glass/melts explain these changes.

Water leaching of borosilicate glasses: experiments, modeling and Monte Carlo simulations; Alteration par l'eau des verres borosilicates: experiences, modelisation et simulations Monte Carlo

Energy Technology Data Exchange (ETDEWEB)

Ledieu, A

2004-10-15

This work is concerned with the corrosion of borosilicate glasses with variable oxide contents. The originality of this study is the complementary use of experiments and numerical simulations. This study is expected to contribute to a better understanding of the corrosion of nuclear waste confinement glasses. First, the corrosion of glasses containing only silicon, boron and sodium oxides has been studied. The kinetics of leaching show that the rate of leaching and the final degree of corrosion sharply depend on the boron content through a percolation mechanism. For some glass contents and some conditions of leaching, the layer which appears at the glass surface stops the release of soluble species (boron and sodium). This altered layer (also called the gel layer) has been characterized with nuclear magnetic resonance (NMR) and small angle X-ray scattering (SAXS) techniques. Second, additional elements have been included in the glass composition. It appears that calcium, zirconium or aluminum oxides strongly modify the final degree of corrosion so that the percolation properties of the boron sub-network is no more a sufficient explanation to account for the behavior of these glasses. Meanwhile, we have developed a theoretical model, based on the dissolution and the reprecipitation of the silicon. Kinetic Monte Carlo simulations have been used in order to test several concepts such as the boron percolation, the local reactivity of weakly soluble elements and the restructuring of the gel layer. This model has been fully validated by comparison with the results on the three oxide glasses. Then, it has been used as a comprehensive tool to investigate the paradoxical behavior of the aluminum and zirconium glasses: although these elements slow down the corrosion kinetics, they lead to a deeper final degree of corrosion. The main contribution of this work is that the final degree of corrosion of borosilicate glasses results from the competition of two opposite mechanisms

Aqueous corrosion of borosilicate glasses: experiments, modeling and Monte-Carlo simulations; Alteration par l'eau des verres borosilicates: experiences, modelisation et simulations Monte-Carlo

Energy Technology Data Exchange (ETDEWEB)

Ledieu, A

2004-10-01

This work is concerned with the corrosion of borosilicate glasses with variable oxide contents. The originality of this study is the complementary use of experiments and numerical simulations. This study is expected to contribute to a better understanding of the corrosion of nuclear waste confinement glasses. First, the corrosion of glasses containing only silicon, boron and sodium oxides has been studied. The kinetics of leaching show that the rate of leaching and the final degree of corrosion sharply depend on the boron content through a percolation mechanism. For some glass contents and some conditions of leaching, the layer which appears at the glass surface stops the release of soluble species (boron and sodium). This altered layer (also called the gel layer) has been characterized with nuclear magnetic resonance (NMR) and small angle X-ray scattering (SAXS) techniques. Second, additional elements have been included in the glass composition. It appears that calcium, zirconium or aluminum oxides strongly modify the final degree of corrosion so that the percolation properties of the boron sub-network is no more a sufficient explanation to account for the behavior of these glasses. Meanwhile, we have developed a theoretical model, based on the dissolution and the reprecipitation of the silicon. Kinetic Monte Carlo simulations have been used in order to test several concepts such as the boron percolation, the local reactivity of weakly soluble elements and the restructuring of the gel layer. This model has been fully validated by comparison with the results on the three oxide glasses. Then, it has been used as a comprehensive tool to investigate the paradoxical behavior of the aluminum and zirconium glasses: although these elements slow down the corrosion kinetics, they lead to a deeper final degree of corrosion. The main contribution of this work is that the final degree of corrosion of borosilicate glasses results from the competition of two opposite mechanisms

Advances in understanding the structure of borosilicate glasses: A Raman spectroscopy study

Energy Technology Data Exchange (ETDEWEB)

Manara, D.; Grandjean, A. [CEA Marcoule, Serv Confinement Deches and Vitrificat, DTCD, DEN - 30 (France); Neuville, D.R. [Institut Physique Globe, Physique des Mineraux et Magmas, CNRS, F-75252 Paris 05 (France)

2009-05-15

This study is focused on the behavior of ternary SiO{sub 2}-Na{sub 2}O-B{sub 2}O{sub 3} borosilicate glasses at temperatures between 298 and 1800 K. Unpolarized Raman spectra were measured up to high temperature. SiO{sub 2}-Na{sub 2}O-B{sub 2}O{sub 3} glass samples were prepared with different values of the ratio R [Na{sub 2}O]/[B{sub 2}O{sub 3}], while the ratio K = [SiO{sub 2}]/[B{sub 2}O{sub 3}] was kept constant and equal 2.12. Spectra were measured at room temperature in samples with 0.43 {<=} R {<=} 1.68, and the effect of the modifier content was clearly observed in these glasses, only in partial agreement with previous literature results. In particular, the formation in the glass of sodium-danburite units Na{sub 2}O-B{sub 2}O{sub 3}-2SiO{sub 2} was postulated. This feature led to a new assessment of R{sup *}, the critical value of R above which every new alkali atom added to the system breaks a Fo-O-Fo (Fo=glass former) bridge causing depolymerization of the glass. A revised formula is proposed to obtain the value of R{sup *} as a function of K. Raman spectra measured at high temperature yielded important information about the temperature-dependent evolution of the borosilicate system. In particular, borate and borosilicate units including tetra-coordinated boron seem to be unstable at high temperature, where the formation of metaborate chains or rings is fostered. Above 1500 degrees C, evaporation of borate compounds is clearly observed, stemming from the small sample size. (authors)

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.

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

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

International Nuclear Information System (INIS)

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

1981-12-01

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

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

Glass Ceramic Formulation Data Package

International Nuclear Information System (INIS)

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

2012-01-01

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

Electron irradiation induced nanocrystal formation in Cu-borosilicate glass

Energy Technology Data Exchange (ETDEWEB)

Sabri, Mohammed Mohammed; Möbus, Günter, E-mail: g.moebus@sheffield.ac.uk [University of Sheffield, Department of Materials Science and Engineering (United Kingdom)

2016-03-15

Nanoscale writing of Cu nanoparticles in glasses is introduced using focused electron irradiation by transmission electron microscopy. Two types of copper borosilicate glasses, one with high and another with low Cu loading, have been tested at energies of 200–300 keV, and formation of Cu nanoparticles in a variety of shapes and sizes using different irradiation conditions is achieved. Electron energy loss spectroscopy analysis, combined with high-resolution transmission electron microscopy imaging, confirmed the irradiation-induced precipitated nanoparticles as metallic, while furnace annealing of the glass triggered dendrite-shaped particles of copper oxide. Unusual patterns of nanoparticle rings and chains under focused electron beam irradiation are also presented. Conclusively, electron beam patterning of Cu-loaded glasses is a promising alternative route to well-established femtosecond laser photoreduction of Cu ions in glass.

An evaluation of electric melter refractories for contact with glass used for the immobilisation of nuclear waste

International Nuclear Information System (INIS)

Hayward, P.J.; George, I.M.

1987-01-01

Corrosion tests have been performed on twelve candidate refractories in contact with borosilicate, titanosilicate, and aluminosilicate melts, in order to rank them for use in an all-electric melter for the production of waste form materials suitable for immobilising nuclear fuel recycle wastes. Viscosities and electrical conductivities of the melts have also been measured to enable optimum processing conditions to be determined. Of the materials tested, the choice of glass contact refractory for the Joule heated melting of the borosilicate and titanosilicate compositions is Monofrax K3 or SEPR 2161, in conjunction with tin oxide electrodes. The aluminosilicate glass waste form would require an alternative method of production (sol-gel processing, or sintering of a precursor frit), because of its high viscosity. (author)

Dielectric behaviour of (Ba,Sr)TiO3 perovskite borosilicate glass ceramics

International Nuclear Information System (INIS)

Yadav, Avadhesh Kumar; Gautam, C.R.

2013-01-01

Various perovskite (Ba,Sr)TiO 3 borosilicate glasses were prepared by rapid melt-quench technique in the glass system ((Ba 1-x Sr x ).TiO 3 )-(2SiO 2 .B 2 O 3 )-(K 2 O)-(La 2 O 3 ). On the basis of differential thermal analysis results, glasses were converted into glass ceramic samples by regulated heat treatment schedules. The dielectric behaviour of crystallized barium strontium titanate borosilicate glass ceramic samples shows diffuse phase transition. The study depicts the dielectric behaviour of glass ceramic sample BST5K1L0.2S814. The double relaxation was observed in glass ceramic samples corresponding 80/20% Ba/Sr due to change in crystal structure from orthorhombic to tetragonal and tetragonal to cubic with variation of temperature. The highest value of dielectric constant was found to be 48289 for the glass ceramic sample BST5K1L0.2S814. The high value of dielectric constant attributed to space charge polarization between the glassy phase and perovskite phase. Due to very high value of dielectric constant, such glass ceramics are used for high energy storage devices. La 2 O 3 acts as nucleating agent for crystallization of glass to glass ceramics and enhances the dielectric constant and retarded dielectric loss. Such glass ceramics can be used in high energy storage devices such as barrier layer capacitors, multilayer capacitors etc. (author)

The structure of leached sodium borosilicate glass

International Nuclear Information System (INIS)

Bunker, B.C.; Tallant, D.R.; Headley, T.J.; Turner, G.L.; Kirkpatrick, R.J.

1988-01-01

Raman spectroscopy, solid state 29 Si, 11 B, 17 O, and 23 Na nuclear magnetic resonance spectroscopy, and transmission electron microscopy have been used to investigate how the structures of sodium borosilicate glasses change during leaching in water at pH 1, 9, and 12. Results show that the random network structure present prior to leaching is transformed into a network of small condensed ring structures and/or colloidal silica particles. The restructuring of leached glass can be rationalised on the basis of simple hydrolysis (depolymerisation) and condensation (repolymerisation) reactions involving Si-O-Si and Si-O-B bonds. The structural changes that occur during leaching influence the properties of the leached layer, including leaching kinetics, crazing and spalling, and slow crack growth. (author)

Modelling the evaporation of boron species. Part 1: Alkali-free borosilicate glass melts

NARCIS (Netherlands)

Limpt, J.A.C. van; Beerkens, R.G.C.; Cook, S.; O'Connor, R.; Simon, J.

2011-01-01

A laboratory test facility has been used to measure the boron evaporation rates from borosilicate glass melts. The impact of furnace atmosphere composition and glass melt composition on the temperature dependent boron evaporation rates has been investigated experimentally. In Part 1 of this paper

Influence of Cu doping in borosilicate bioactive glass and the properties of its derived scaffolds.

Science.gov (United States)

Wang, Hui; Zhao, Shichang; Xiao, Wei; Xue, Jingzhe; Shen, Youqu; Zhou, Jie; Huang, Wenhai; Rahaman, Mohamed N; Zhang, Changqing; Wang, Deping

2016-01-01

Copper doped borosilicate glasses (BG-Cu) were studied by means of FT-IR, Raman, UV-vis and NMR spectroscopies to investigate the changes that appeared in the structure of borosilicate glass matrix by doping copper ions. Micro-fil and immunohistochemistry analysis were applied to study the angiogenesis of its derived scaffolds in vivo. Results indicated that the Cu ions significantly increased the B-O bond of BO4 groups at 980 cm(-1), while they decrease that of BO2O(-) groups at 1440-1470 cm(-1) as shown by Raman spectra. A negative shift was observed from (11)B and (29)Si NMR spectra. The (11)B NMR spectra exhibited a clear transformation from BO3 into BO4 groups, caused by the agglutination effect of the Cu ions and the charge balance of the agglomerate in the glass network, leading to a more stable glass network and lower ions release rate in the degradation process. Furthermore, the BG-Cu scaffolds significantly enhanced blood vessel formation in rat calvarial defects at 8 weeks post-implantation. Generally, it suggested that the introduction of Cu into borosilicate glass endowed glass and its derived scaffolds with good properties, and the cooperation of Cu with bioactive glass may pave a new way for tissue engineering. Copyright © 2015 Elsevier B.V. All rights reserved.

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

Performance Characteristics of Waste Glass Powder Substituting Portland Cement in Mortar Mixtures

Science.gov (United States)

Kara, P.; Csetényi, L. J.; Borosnyói, A.

2016-04-01

In the present work, soda-lime glass cullet (flint, amber, green) and special glass cullet (soda-alkaline earth-silicate glass coming from low pressure mercury-discharge lamp cullet and incandescent light bulb borosilicate glass waste cullet) were ground into fine powders in a laboratory planetary ball mill for 30 minutes. CEM I 42.5N Portland cement was applied in mortar mixtures, substituted with waste glass powder at levels of 20% and 30%. Characterisation and testing of waste glass powders included fineness by laser diffraction particle size analysis, specific surface area by nitrogen adsorption technique, particle density by pycnometry and chemical analysis by X-ray fluorescence spectrophotometry. Compressive strength, early age shrinkage cracking and drying shrinkage tests, heat of hydration of mortars, temperature of hydration, X-ray diffraction analysis and volume stability tests were performed to observe the influence of waste glass powder substitution for Portland cement on physical and engineering properties of mortar mixtures.

Effect of the nature of alkali and alkaline-earth oxides on the structure and crystallization of an alumino-borosilicate glass developed to immobilize highly concentrated nuclear waste solutions

International Nuclear Information System (INIS)

Quintas, A.; Caurant, D.; Majerus, O.; Charpentier, T.; Dussossoy, J.L.

2008-01-01

A complex rare-earth rich alumino-borosilicate glass has been proved to be a good candidate for the immobilization of new high level radioactive wastes. A simplified seven-oxides composition of this glass was selected for this study. In this system, sodium and calcium cations were supposed in other works to simulate respectively all the other alkali (R + = Li + , Rb + , Cs + ) and alkaline-earth (R 2+ = Sr 2+ , Ba 2+ ) cations present in the complex glass composition. Moreover, neodymium or lanthanum are used here to simulate all the rare-earths and actinides occurring in waste solutions. In order to study the impact of the nature of R + and R 2+ cations on both glass structure and melt crystallization tendency during cooling, two glass series were prepared by replacing either Na + or Ca 2+ cations in the simplified glass by respectively (Li + , K + , Rb + , Cs + ) or (Mg 2+ , Sr 2+ , Ba 2+ ) cations. From these substitutions, it was established that alkali ions are preferentially involved in the charge compensation of (AlO 4 ) - entities in the glass network comparatively to alkaline-earth ions. The glass compositions containing calcium give way to the crystallization of an apatite silicate phase bearing calcium and rare-earth ions. The melt crystallization tendency during cooling strongly varies with the nature of the alkaline-earth. (authors)

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

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

Electrical conductivity and viscosity of borosilicate glasses and melts

DEFF Research Database (Denmark)

Ehrt, Doris; Keding, Ralf

2009-01-01

, 0 to 62·5 mol% B2O3, and 25 to 85 mol% SiO2. The glass samples were characterised by different methods. Refractive indices, density and thermal expansion were measured. Phase separation effects were investigated by electron microscopy. The electrical conductivity of glasses and melts were determined......Simple sodium borosilicate and silicate glasses were melted on a very large scale (35 l Pt crucible) to prepare model glasses of optical quality in order to investigate various properties depending on their structure. The composition of the glass samples varied in a wide range: 3 to 33·3 mol% Na2O...... by impedance measurements in a wide temperature range (250 to 1450°C). The activation energies were calculated by Arrhenius plots in various temperature regions: below the glass transition temperature, Tg, above the melting point, Tl, and between Tg and Tl. Viscosity measurements were carried out...

A Glass-Ceramic Waste Forms for the Immobilization of Rare Earth Oxides from the Pyroprocessing Waste salt

International Nuclear Information System (INIS)

Ahn, Byung-Gil; Park, Hwan-Seo; Kim, Hwan-Young; Kim, In-Tae

2008-01-01

The fission product of rare earth (RE) oxide wastes are generates during the pyroprocess . Borosilicate glass or some ceramic materials such as monazite, apatite or sodium zirconium phosphate (NZP) have been a prospective host matrix through lots of experimental results. Silicate glasses have long been the preferred waste form for the immobilization of HLW. In immobilization of the RE oxides, the developed process on an industrial scale involves their incorporation into a glass matrix, by melting under 1200 ∼ 1300 .deg. C. Instead of the melting process, glass powder sintering is lower temperature (∼ 900 .deg. C) required for the process which implies less demanding conditions for the equipment and a less evaporation of volatile radionuclides. This study reports the behaviors, direct vitrification of RE oxides with glass frit, glass powder sintering of REceramic with glass frit, formation of RE-apatite (or REmonazite) ceramic according to reaction temperature, and the leach resistance of the solidified waste forms

Comparison of costs for solidification of high-level radioactive waste solutions: glass monoliths vs metal matrices

International Nuclear Information System (INIS)

Jardine, L.J.; Carlton, R.E.; Steindler, M.J.

1981-05-01

A comparative economic analysis was made of four solidification processes for liquid high-level radioactive waste. Two processes produced borosilicate glass monoliths and two others produced metal matrix composites of lead and borosilicate glass beads and lead and supercalcine pellets. Within the uncertainties of the cost (1979 dollars) estimates, the cost of the four processes was about the same, with the major cost component being the cost of the primary building structure. Equipment costs and operating and maintenance costs formed only a small portion of the building structure costs for all processes

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)

Effect of heat treatment duration on phase separation of sodium borosilicate glass, containing copper

International Nuclear Information System (INIS)

Shejnina, T.G.; Gutner, S.Kh.; Anan'in, N.I.

1989-01-01

The effect of heat treatment duration on phase separation of sodium borosilicate (SBS) glass, containing copper is studied. It is stated that phase separation close to equilibrium one is attained under 12 hours of heat treatment of SBS glass containing copper

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

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

Processability analysis of candidate waste forms

International Nuclear Information System (INIS)

Gould, T.H. Jr.; Dunson, J.B. Jr.; Eisenberg, A.M.; Haight, H.G. Jr.; Mello, V.E.; Schuyler, R.L. III.

1982-01-01

A quantitative merit evaluation, or processability analysis, was performed to assess the relative difficulty of remote processing of Savannah River Plant high-level wastes for seven alternative waste form candidates. The reference borosilicate glass process was rated as the simplest, followed by FUETAP concrete, glass marbles in a lead matrix, high-silica glass, crystalline ceramics (SYNROC-D and tailored ceramics), and coated ceramic particles. Cost estimates for the borosilicate glass, high-silica glass, and ceramic waste form processing facilities are also reported

3 and 4 oxidation state element solubilities in borosilicate glasses. Implement to actinides in nuclear glasses

International Nuclear Information System (INIS)

Cachia, J.N.

2005-12-01

In order to ensure optimal radionuclides containment, the knowledge of the actinide loading limits in nuclear waste glasses and also the comprehension of the solubilization mechanisms of these elements are essential. A first part of this manuscript deals with the study of the differences in solubility of the tri and tetravalent elements (actinides and surrogates) particularly in function of the melting temperature. The results obtained indicate that trivalent elements (La, Gd, Nd, Am, Cm) exhibit a higher solubility than tetravalent elements (Hf, Th, Pu). Consequently, it was planned to reduce plutonium at the oxidation state (III), the later being essentially tetravalent in borosilicate glasses. An innovating reduction process of multi-valent elements (cerium, plutonium) using silicon nitride has been developed in a second part of this work. Reduced plutonium-bearing glasses synthesized by Si 3 N 4 addition made it possible to double the plutonium solubility from 2 to 4 wt% at 1200 deg C. A structural approach to investigate the differences between tri and tetravalent elements was finally undertaken. These investigations were carried out by X-rays Absorption Spectroscopy (EXAFS) and NMR. Trivalent rare earth and actinide elements seem to behave as network modifiers while tetravalent elements rather present true intermediaries' behaviour. (author)

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

International Nuclear Information System (INIS)

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

1986-09-01

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

Effects of system size and cooling rate on the structure and properties of sodium borosilicate glasses from molecular dynamics simulations.

Science.gov (United States)

Deng, Lu; Du, Jincheng

2018-01-14

Borosilicate glasses form an important glass forming system in both glass science and technologies. The structure and property changes of borosilicate glasses as a function of thermal history in terms of cooling rate during glass formation and simulation system sizes used in classical molecular dynamics (MD) simulation were investigated with recently developed composition dependent partial charge potentials. Short and medium range structural features such as boron coordination, Si and B Q n distributions, and ring size distributions were analyzed to elucidate the effects of cooling rate and simulation system size on these structure features and selected glass properties such as glass transition temperature, vibration density of states, and mechanical properties. Neutron structure factors, neutron broadened pair distribution functions, and vibrational density of states were calculated and compared with results from experiments as well as ab initio calculations to validate the structure models. The results clearly indicate that both cooling rate and system size play an important role on the structures of these glasses, mainly by affecting the 3 B and 4 B distributions and consequently properties of the glasses. It was also found that different structure features and properties converge at different sizes or cooling rates; thus convergence tests are needed in simulations of the borosilicate glasses depending on the targeted properties. The results also shed light on the complex thermal history dependence on structure and properties in borosilicate glasses and the protocols in MD simulations of these and other glass materials.

Effects of system size and cooling rate on the structure and properties of sodium borosilicate glasses from molecular dynamics simulations

Science.gov (United States)

Deng, Lu; Du, Jincheng

2018-01-01

Borosilicate glasses form an important glass forming system in both glass science and technologies. The structure and property changes of borosilicate glasses as a function of thermal history in terms of cooling rate during glass formation and simulation system sizes used in classical molecular dynamics (MD) simulation were investigated with recently developed composition dependent partial charge potentials. Short and medium range structural features such as boron coordination, Si and B Qn distributions, and ring size distributions were analyzed to elucidate the effects of cooling rate and simulation system size on these structure features and selected glass properties such as glass transition temperature, vibration density of states, and mechanical properties. Neutron structure factors, neutron broadened pair distribution functions, and vibrational density of states were calculated and compared with results from experiments as well as ab initio calculations to validate the structure models. The results clearly indicate that both cooling rate and system size play an important role on the structures of these glasses, mainly by affecting the 3B and 4B distributions and consequently properties of the glasses. It was also found that different structure features and properties converge at different sizes or cooling rates; thus convergence tests are needed in simulations of the borosilicate glasses depending on the targeted properties. The results also shed light on the complex thermal history dependence on structure and properties in borosilicate glasses and the protocols in MD simulations of these and other glass materials.

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

Materials interactions relating to long-term geologic disposal of nuclear waste glass

International Nuclear Information System (INIS)

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

1987-01-01

In the geologic disposal of nuclear waste glass, the glass will eventually interact with groundwater in the repository system. Interactions can also occur between the glass and other waste package materials that are present. These include the steel canister that holds the glass, the metal overpack over the canister, backfill materials that may be used, and the repository host rock. This review paper systematizes the additional interactions that materials in the waste package will impose on the borosilicate glass waste form-groundwater interactions. The repository geologies reviewed are tuff, salt, basalt, and granite. The interactions emphasized are those appropriate to conditions expected after repository closure, e.g. oxic vs anoxic conditions. Whenever possible, the effect of radiation from the waste form on the interactions is examined. The interactions are evaluated based on their effect on the release and speciation of various elements including radionuclides from the glass. It is noted when further tests of repository interactions are needed before long-term predictions can be made. 63 references, 1 table

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.

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.)

Influence of Cu doping in borosilicate bioactive glass and the properties of its derived scaffolds

Energy Technology Data Exchange (ETDEWEB)

Wang, Hui [School of Materials Science and Engineering, Tongji University, Shanghai 2001804 (China); Zhao, Shichang [Department of Orthopedic Surgery, Shanghai Sixth People' s Hospital, Shanghai Jiao Tong University, Shanghai 200233 (China); Xiao, Wei [Department of Materials Science and Engineering, and Center for Biomedical Science and Engineering, Missouri University of Science and Technology, Rolla, MO 65409-0340 (United States); Xue, Jingzhe [Department of Chemistry, Tongji University, Shanghai 200092 (China); Shen, Youqu [Department of Materials Science and Engineering, and Center for Biomedical Science and Engineering, Missouri University of Science and Technology, Rolla, MO 65409-0340 (United States); Zhou, Jie; Huang, Wenhai [School of Materials Science and Engineering, Tongji University, Shanghai 2001804 (China); Rahaman, Mohamed N. [Department of Materials Science and Engineering, and Center for Biomedical Science and Engineering, Missouri University of Science and Technology, Rolla, MO 65409-0340 (United States); Zhang, Changqing, E-mail: shzhangchangqing@163.com [Department of Orthopedic Surgery, Shanghai Sixth People' s Hospital, Shanghai Jiao Tong University, Shanghai 200233 (China); Wang, Deping, E-mail: wdpshk@tongji.edu.cn [School of Materials Science and Engineering, Tongji University, Shanghai 2001804 (China)

2016-01-01

Copper doped borosilicate glasses (BG–Cu) were studied by means of FT-IR, Raman, UV–vis and NMR spectroscopies to investigate the changes that appeared in the structure of borosilicate glass matrix by doping copper ions. Micro-fil and immunohistochemistry analysis were applied to study the angiogenesis of its derived scaffolds in vivo. Results indicated that the Cu ions significantly increased the B–O bond of BO{sub 4} groups at 980 cm{sup −1}, while they decrease that of BO{sub 2}O{sup −} groups at 1440–1470 cm{sup −1} as shown by Raman spectra. A negative shift was observed from {sup 11}B and {sup 29}Si NMR spectra. The {sup 11}B NMR spectra exhibited a clear transformation from BO{sub 3} into BO{sub 4} groups, caused by the agglutination effect of the Cu ions and the charge balance of the agglomerate in the glass network, leading to a more stable glass network and lower ions release rate in the degradation process. Furthermore, the BG–Cu scaffolds significantly enhanced blood vessel formation in rat calvarial defects at 8 weeks post-implantation. Generally, it suggested that the introduction of Cu into borosilicate glass endowed glass and its derived scaffolds with good properties, and the cooperation of Cu with bioactive glass may pave a new way for tissue engineering. - Highlights: • Agglutination effect of Cu{sup 2+} and charge balance of agglomerate lead to more stable glass. • Lower degradability and lower ions release were found in BG-Cu scaffolds. • Excellent angiogenesis and sustain Cu{sup 2+} release were endowed by doping Cu.

Utilization of Cs137 to generate a radiation barrier for weapons grade plutonium immobilized in borosilicate glass canisters. Revision 1

International Nuclear Information System (INIS)

Jardine, L.J.; Armantrout, G.A.; Collins, L.F.

1995-01-01

One of the ways recommended by a recent National Academy of Sciences study to dispose of excess weapons-grade plutonium is to encapsulate the plutonium in a glass in combination with high-level radioactive wastes (HLW) to generate an intense radiation dose rate field. The objective is to render the plutonium as difficult to access as the plutonium contained in existing US commercial spent light-water reactor (LWR) fuel until it can be disposed of in a permanent geological repository. A radiation dose rate from a sealed canister of 1,000 rem/h (10 Sv/h) at 1 meter for at least 30 years after fabrication was assumed in this paper to be a radiation dose comparable to spent LWR fuel. This can be achieved by encapsulating the plutonium in a borosilicate glass with an adequate amount of a single fission product in the HLWS, namely radioactive Cs 137 . One hundred thousand curies of Cs 137 will generate a dose rate of 1,000 rem/h (10 Sv/h) at 1 meter for at least 30 years when imbedded into canisters of the size proposed for the Savannah River Site's vitrified high-level wastes. The United States has a current inventory of 54 MCi of CS 137 that has been separated from defense HLWs and is in sealed capsules. This single curie inventory is sufficient to spike 50 metric tons of excess weapons-grade plutonium if plutonium can be loaded at 5.5 wt% in glass, or 540 canisters. Additional CS 137 inventories exist in the United States' HLWs from past reprocessing operations, should additional curies be required. Using only one fission product, CS 137 , rather than the multiple chemical elements and compounds in HLWs to generate a high radiation dose rate from a glass canister greatly simplifies the processing engineering retirement for encapsulating plutonium in a borosilicate glass

Determination of occluded helium and oxygen in irradiated borosilicate glass samples

International Nuclear Information System (INIS)

Ramanjaneyulu, P.S.; Kulkarni, A.S.; Shrivastava, K.C.; Yadav, C.S.; Saxena, M.K.; Tomar, B.S.; Ramakumar, K.L.; Shah, J.G.

2015-01-01

Occluded gases in irradiated borosilicate glass were determined at 573, 873 and 1273 K for understanding the radiation damage in glass matrix. Hot vacuum extraction coupled with a quadrupole mass spectrometer technique was employed for the measurements. Relative sensitivity factors of various gases in QMS system were also determined and used for gas composition calculations. At 573 K only helium was found to get released whereas at 873 and 1273 K both helium and oxygen were released with major fraction of oxygen. (author)

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)

Evaluation of long-term leaching of borosilicate glass in pure water

International Nuclear Information System (INIS)

Lanza, F.; Parnisari, E.

1979-01-01

In order to confirm the validity of the model of the dissolving sphere, a series of tests of leaching of borosilicate glasses has been performed. The leaching rate is followed by measurements of weight loss and analysis of the surface composition. An accumulation at the surface of the less soluble cations is observed

Optical and structural investigation on sodium borosilicate glasses doped with Cr2O3

Science.gov (United States)

Ebrahimi, E.; Rezvani, M.

2018-02-01

In this work, Sodium borosilicate glasses with chemical composition of 60% SiO2-20% B2O3-20%Na2O doped with different contents of Cr2O3 were prepared by melting-quenching method. Physical, structural and optical properties of glasses were investigated by studying density and molar volume, Fourier Transform Infrared (FT-IR) Spectra and UV-visible absorption spectroscopy. The results showed an increase in density of glasses with the increase of Cr2O3 that can be due to addition of oxide with high molar mass. The optical absorption spectra of un-doped glass reveals UV absorption due to trace iron impurities with no visible band however Cr2O3 doped glasses shows absorption in visible range that are characteristic. Increasing of Cr3 + ions in the glassy microstructure of samples provides a semiconducting character to Sodium borosilicate glass by reducing the direct and indirect optical band gaps of glass samples from 3.79 to 2.59 (ev) and 3.36 to 2.09 (ev), respectively. These changes could be attributed to the role of Cr3 + ions as the network former which asserts improvement of semiconducting behavior in presence of Cr2O3.

X-ray irradiation induced reduction and nanoclustering of lead in borosilicate glass

NARCIS (Netherlands)

Stanley, H.B.; Banerjee, D.; Breemen, van L.C.A.; Ciston, J.; Liebscher, C.H.; Martis, V.; Merino, D.H.; Longo, A.; Pattison, P.; Peters, G.W.M.; Portale, G.; Sen, Sabyasachi; Bras, W.

2014-01-01

We have studied the formation of nanoparticles in lead sulfide (PbS)-doped borosilicate glass subjected to a two-step nucleation and growth heat treatment using in situ small-angle X-ray scattering (SAXS). The microstructure produced was subsequently characterized using X-ray powder diffraction

There Is Still Room for Improvement: Presentation of a Neutral Borosilicate Glass with Improved Chemical Stability for Parenteral Packaging.

Science.gov (United States)

Boltres, Bettine; Tratzky, Stephan; Kass, Christof; Eichholz, Rainer; Naß, Peter

2016-01-01

For pharmaceutical parenteral packaging the glass compositions have always been either Type I borosilicate or Type III soda-lime glass. As both the compositions and certain chemical and physical properties are mandated by international standards, there has not been room for any changes. However, by applying only minor adjustments, a borosilicate glass was developed that showed improved chemical stability. The chemical composition is still in the range of currently used borosilicate glasses, which makes it a Type I glass according to all current pharmacopeia. A study was performed on glass vials comparing the new glass with the standard FIOLAX(®) and two other publicly available glasses. In an extraction study with water at 121 °C the new glass showed the highest chemical stability with the lowest amount of extractables. In an accelerated ageing study, which was done with water, phosphate, and carbonate buffer at 40 °C for 12 months, the new glass also proved to have the lowest amount of leachables. In this article the new glass and the results from the studies are presented, showing the reader how much of an effect can be attained with only minor adjustments if the scientific fundamentals are clear. The pharmaceutical market has been quite constant and risk-oriented due to the high impact on the safety of the patient. As any change necessitates a complicated change process, this has, in consequence, lead the industry to resist changing the parenteral primary packaging material for decades. The main glasses have either been Type I borosilicate or Type III soda-lime glass. On the other hand, a combination of improved inspection systems and the development of more sensitive biologically based drugs has elevated the standards for parental packaging materials. For example, the measurement of extractables and leachables from the packaging material steadily came into focus. In this article, a new glass is presented that still belongs to the group of Type I borosilicate

Localized chemistry of 99Tc in simulated low activity waste glass

Science.gov (United States)

Weaver, Jamie L.

A priority of the United States Department of Energy (DOE) is to dispose of the nuclear waste accumulated in the underground tanks at the Hanford Nuclear Reservation in Richland, WA. Incorporation and stabilization of technetium (99Tc) from these tanks into vitrified waste forms is a concern to the waste glass community and DOE due to 99Tc's long half-life ( 2.13˙105 y), and its high mobility in the subsurface environment under oxidizing conditions. Working in collaboration with researchers at Pacific Northwest National Laboratory (PNNL) and other national laboratories, plans were formulated to obtain first-of-a-kind chemical structure determination of poorly understood and environmentally relevant technetium compounds that relate to the chemistry of the Tc in nuclear waste glasses. Knowledge of the structure and spectral signature of these compounds aid in refining the understanding of 99Tc incorporation into and release from oxide based waste glass. In this research a first-of-its kind mechanism for the behavior of 99Tc during vitrification is presented, and the structural role of Tc(VII) and (IV) in borosilicate waste glasses is readdressed.

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

New insight into nanoparticle precipitation by electron beams in borosilicate glasses

Science.gov (United States)

Sabri, M. M.; Möbus, G.

2017-06-01

Nanoprecipitation in different oxide glasses by means of electron irradiation in transmission electron microscopy (TEM) has been compared in this study. Upon irradiation, groups or patterns of nanoparticles with various morphologies and sizes were formed in borosilicate glasses, loaded with zinc, copper, and silver. The study successfully includes loading ranges for the target metal from doping level (1%) over medium level (20%) to majority phase (60%). It is found that particle patterning resolution is affected by parallel processes of amorphous phase separation, glass ablation, and delocalised precipitation. In addition, via an in-situ study, it is confirmed that by heating alone without irradiation, no precipitate nanoparticles form.

New insight into nanoparticle precipitation by electron beams in borosilicate glasses

Energy Technology Data Exchange (ETDEWEB)

Sabri, M.M.; Moebus, G. [University of Sheffield, Department of Materials Science and Engineering (United Kingdom)

2017-06-15

Nanoprecipitation in different oxide glasses by means of electron irradiation in transmission electron microscopy (TEM) has been compared in this study. Upon irradiation, groups or patterns of nanoparticles with various morphologies and sizes were formed in borosilicate glasses, loaded with zinc, copper, and silver. The study successfully includes loading ranges for the target metal from doping level (1%) over medium level (20%) to majority phase (60%). It is found that particle patterning resolution is affected by parallel processes of amorphous phase separation, glass ablation, and delocalised precipitation. In addition, via an in-situ study, it is confirmed that by heating alone without irradiation, no precipitate nanoparticles form. (orig.)

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

On confirmation of abandonment of imported waste (glass solidified bodies) outside business places

International Nuclear Information System (INIS)

1996-01-01

Electric power companies entrust the reprocessing of spent fuel generated from nuclear power stations to COGEMA in France, and in April, 1995, 28 high level radioactive wastes (glass solidified bodies) generated by the reprocessing were returned. When these glass solidified wastes are abandoned in the waste management facility of Japan Nuclear Fuel Service Co., it was decided to receive the confirmation of the prime minister on the measures based on the relevant law. Four electric power companies submitted the application and the explanation paper. As to the contents of the glass solidified wastes, the technical inspection was carried out by Bureau Veritas. Considering that this import of glass solidified wastes is the first in Japan, Science and Technology Agency carried out the measurement of all 28 wastes. The results are reported. It was confirmed that the measures for the abandonment taken by four electric power companies conform to the stipulation. The contents of the confirmation are reported in the order of the stipulation. These wastes were solidified with borosilicate glass in 5 mm thick stainless steel vessels, and the welding was done properly. (K.I.)

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

Radionuclides containment in nuclear glasses. An overview

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

Electron irradiation effect on bubble formation and growth in a sodium borosilicate glass

International Nuclear Information System (INIS)

Chen, X.; Birtcher, R. C.; Donnelly, S. E.

2000-01-01

In this study, the authors studied simultaneous and intermittent electron irradiation effects on bubble growth in a simple sodium borosilicate glass during Xe ion implantation at 200 C. Simultaneous electron irradiation increases the average bubble size in the glass. This enhanced diffusion is also shown by the migration of Xe from bubbles into the matrix when the sample is irradiated by an electron beam after the Xe implantation

Sulphate solubility and sulphate diffusion in oxide glasses: implications for the containment of sulphate-bearing nuclear wastes

International Nuclear Information System (INIS)

Lenoir, M.

2009-09-01

The thesis deals with sulphate solubility and sulphate diffusion in oxide glasses, in order to control sulphate incorporation and sulphate volatilization in nuclear waste glasses. It was conducted on simplified compositions, in the SiO 2 -B 2 O 3 -R 2 O (R = Li, Na, K, Cs), SiO 2 -B 2 O 3 -BaO and V 2 O 5 -B 2 O 3 -BaO systems. These compositions allowed us to study the influence of the nature of network-modifying ions (Li + , Na + , K + , Cs + or Ba 2+ ) and also of former elements (Si, B, V), on structure and properties of glasses. Sulphate volatility is studied in sodium borosilicate melts using an innovative technique of sulphate quantitation with Raman spectroscopy. This technique is useful to obtain kinetic curves of sulphate volatilization. The establishment of a model to fit these curves leads to the determination of diffusion coefficients of sulphate. These diffusion coefficients can thus be compared to diffusion coefficients of other species, determined by other techniques and presented in the literature. They are also linked to diffusion coefficients in relation with the viscosity of the melts. Concerning sulphate solubility in glasses, it depends on glass composition and on the nature of sulphate incorporated. Sulphate incorporation in alkali borosilicate glasses leads to the formation of a sulphate layer floating on top of the melt. Sulphate incorporation in barium borosilicate and boro-vanadate glasses leads to the crystallization of sulphate species inside the vitreous matrix. Moreover, sulphate solubility is higher in these glasses than in alkali borosilicates. Finally, exchanges between cations present in glasses and cations present in the sulphate phase are also studied. (author)

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

Determination of boron in borosilicate glasses by neutron capture prompt gamma-ray activation analysis

Energy Technology Data Exchange (ETDEWEB)

Riley, Jr, J E; Lindstrom, R M

1987-01-01

Major levels of boron in borosilicate glasses were determined nondestructively by neutron activation analysis. The effects of neutron self-shielding by boron (1 to 8% by weight) are examined. Results of the analysis of a series of glasses with increasing boron composition are 1.150 +- .005% and 7.766 +- .035% for the low and high members of the series. Once analyzed, the glasses are useful as secondary standards for alpha track counting, and also ion and electron microprobe analyses of glasses. 12 refs.; 3 tables.

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

Energy Technology Data Exchange (ETDEWEB)

Fox, K.; Edwards, T.

2012-05-08

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

Gas gain stability of MSGCs on borosilicate glass

International Nuclear Information System (INIS)

Berg, F.D. van den; Daum, C.; Eijk, B. van; Hartjes, F.G.; Udo, F.; Zhukov, V.

1996-01-01

The gas gain stability in Micro Strip Gas Counters fabricated on borosilicate glass has been investigated for operation at high radiation fluxes up to 10 5 Hz/mm 2 in a X-ray set-up. We discuss effects of short and long term irradiation on the gain for different voltage settings and strip metalisations. The application of a high drift field and a back plane on positive potential provides good high rate operation up to 10 4 Hz/mm 2 . Ageing is observed to be strongly dependent on the stirp metalisation; gold and nickel perform better than aluminium and copper. (orig.)

Alternative-waste-form evaluation for Savannah River Plant high-level waste

International Nuclear Information System (INIS)

Gould, T.H. Jr.; Crandall, J.L.

1982-01-01

Results of the waste form evaluation are summarized as: risks of human exposure are comparable and extremely small for either borosilicate glass or Synroc ceramic. Waste form properties are more than adequate for either form. The waste form decision can therefore be made on the basis of practicality and cost effectiveness. Synroc offers lower costs for transportation and emplacement. The borosilicate glass form offers the lowest total disposal cost, much simpler and less costly production, an established and proven process, lower future development costs, and an earlier startup of the DWPF

Waste package designs for disposal of high-level waste in salt formations

International Nuclear Information System (INIS)

Basham, S.J. Jr.; Carr, J.A.

1984-01-01

In the United States of America the selected method for disposal of radioactive waste is mined repositories located in suitable geohydrological settings. Currently four types of host rocks are under consideration: tuff, basalt, crystalline rock and salt. Development of waste package designs for incorporation in mined salt repositories is discussed. The three pertinent high-level waste forms are: spent fuel, as disassembled and close-packed fuel pins in a mild steel canister; commercial high-level waste (CHLW), as borosilicate glass in stainless-steel canisters; defence high-level waste (DHLW), as borosilicate glass in stainless-steel canisters. The canisters are production and handling items only. They have no planned long-term isolation function. Each waste form requires a different approach in package design. However, the general geometry and the materials of the three designs are identical. The selected waste package design is an overpack of low carbon steel with a welded closure. This container surrounds the waste forms. Studies to better define brine quantity and composition, radiation effects on the salt and brines, long-term corrosion behaviour of the low carbon steel, and the leaching behaviour of the spent fuel and borosilicate glass waste forms are continuing. (author)

Glass melter assembly for the Hanford Waste Vitrification Plant

International Nuclear Information System (INIS)

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

1993-01-01

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

Understanding of the mechanical and structural changes induced by alpha particles and heavy ions in the French simulated nuclear waste glass

International Nuclear Information System (INIS)

Karakurt, G.; Abdelouas, A.; Guin, J.-P.; Nivard, M.; Sauvage, T.; Paris, M.; Bardeau, J.-F.

2016-01-01

Borosilicate glasses are considered for the long-term confinement of high-level nuclear wastes. External irradiations with 1 MeV He + ions and 7 MeV Au 5+ ions were performed to simulate effects produced by alpha particles and by recoil nuclei in the simulated SON68 nuclear waste glass. To better understand the structural modifications, irradiations were also carried out on a 6-oxides borosilicate glass, a simplified version of the SON68 glass (ISG glass). The mechanical and macroscopic properties of the glasses were studied as function of the deposited electronic and nuclear energies. Alpha particles and gold ions induced a volume change up to −0.7% and −2.7%, respectively, depending on the glass composition. Nano-indentations tests were used to determine the mechanical properties of the irradiated glasses. A decrease of about −22% to −38% of the hardness and a decrease of the reduced Young's modulus by −8% were measured after irradiations. The evolution of the glass structure was studied by Raman spectroscopy, and also 11 B and 27 Al Nuclear Magnetic Resonance (MAS-NMR) on a 20 MeV Kr irradiated ISG glass powder. A decrease of the silica network connectivity after irradiation with alpha particles and gold ions is deduced from the structural changes observations. NMR spectra revealed a partial conversion of BO 4 to BO 3 units but also a formation of AlO 5 and AlO 6 species after irradiation with Kr ions. The relationships between the mechanical and structural changes are also discussed. - Highlights: • Mechanical and structural properties of two borosilicate glass compositions irradiated with alpha particles and heavy ions were investigated. • Both kinds of particles induced a decrease of the hardness, reduced Young's modulus and density. • Electronic and nuclear interactions are responsible for the changes observed. • The evolution of the mechanical properties under irradiation is linked to the changes occured in the

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

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

International Nuclear Information System (INIS)

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

1986-01-01

This review paper systematizes the additional interactions that materials in a geologic repository will impose on the borosilicate glass waste form-groundwater interactions. These materials are the steel canister that holds the glass, the steel overpack over the canister, backfill materials that may be used, and last, the repository host rock. The repository geologies reviewed are tuff, salt, basalt, and granite. The interactions emphasized are those appropriate to conditions expected after repository closure, e.g., oxic vs anoxic conditions. Whenever possible, the effect of radiation from the waste form on the interaction(s) is examined. The interactions are evaluated based on their effect on the release and speciation of various elements including radionuclides from the glass. Repository relevant interactions testing that requires further study before long-term predictions can be made are noted. 62 refs

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

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

Irradiations effects on the structure of boro-silicated glasses: long term behaviour of nuclear waste glassy matrices; Effets d'irradiations sur la structure de verres borosilicates - comportement a long terme des matrices vitreuses de stockage des dechets nucleaires

Energy Technology Data Exchange (ETDEWEB)

Bonfils, J. de

2007-09-15

This work deals with the long term behaviour of R7T7-type nuclear waste glasses and more particularly of non-active boro-silicated glasses made up of 3 or 5 oxides. Radioactivity of active glasses is simulated by multi energies ions implantations which reproduce the same defects. The damages due to the alpha particles are simulated by helium ions implantations and those corresponding to the recoil nucleus are obtained with gold ions ones. Minor actinides, stemming from the used fuel, is simulated by trivalent rare-earths (Eu{sup 3+} and Nd{sup 3+}). In a first part, we have shown by macroscopic experiments (Vickers hardness - swelling) and optical spectroscopies (Raman - ATR-IR) that the structure of the glassy matrices is modified under implantations until a dose of 2,3.10{sup 13} at.cm{sup -2}, which corresponds to a R7T7 storage time estimated at 300 years. Beyond this dose, no additional modifications have been observed. The second part concerns the local environment of the rare-earth ions in glasses. Two different environments were found and identified as follows: one is a silicate rich one and the other is attributed to a borate rich one. (author)

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

International Nuclear Information System (INIS)

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

1980-01-01

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

Phase formation during corrosion experiments with two simulated borosilicate nuclear waste glasses

International Nuclear Information System (INIS)

Haaker, R.F.

1985-10-01

Corrosion products resulting from the reaction of simulated high-level radioactive waste glasses with various solutions have been identified. At 200degC, in saturated NaCl, a degree of reaction of 10 g C31-3 glass or 2.6 g SON 68 glass per liter of solution was obtained. Analcime, vermiculite (a phyllosilicate) and a 2:1 zinc silicate are the major silica containing alteration products for the C31-3 glass. Analcime was the only silicate alteration product which could be identified for SON 68 glass. C31-3 glass appeared to be less reactive with a quinary brine containing Mg ++ than with NaCl. With the quinary brine, montmorillonite (a phyllosilicate) was the predominant silica containing alteration product. Hydrotalcite (a Mg-Al hydroxysulfate) and montmorillonite were the major Al-containing phases. A phyllosilicate, probably montmorillonite, was observed to form during the reaction of SON 68 glass with quinary brine. With either glass, modified NaCl brines which contained small amounts of MgCl 2 seem to have the effect of decreasing the amount of analcime and increasing the amount of phyllosilicate which is formed. In the case of C31-3 glass, there is approximately enough Mg, Al and Zn to precipitate most of the leached Si; measured Si concentrations remain well below that expected for amorphous silica. SON 68 glass has less Zn, Al and Mg than C31-3 glass and much higher Si concentrations of the leachates. (orig./RB)

Immobilization of Uranium Silicides in Sintered Glass

International Nuclear Information System (INIS)

Mateos, P.; Russo, D.O.; Heredia, A.D.; Sanfilippo, M.

2003-01-01

High activity nuclear spent fuels vitrification by fusion is a well known technology which has industrial scale in France, England, Japan, EEUU. Borosilicates glasses are used in this process.Sintered glasses are an alternative to the immobilization task in which there is also a wide experience around the world.The available technics are: cold pressing and sintering , hot-pressing and hot isostatic pressing.This work compares Borosilicates and Iron silicates sintered glasses behaviour when different ammounts of nuclear simulated waste is added

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

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.

Solubility of actinides and surrogates in nuclear glasses

International Nuclear Information System (INIS)

Lopez, Ch.

2003-01-01

The nuclear wastes are currently incorporated in borosilicate glass matrices. The resulting glass must be perfectly homogeneous. The work discussed here is a study of actinide (thorium and plutonium) solubility in borosilicate glass, undertaken to assess the extent of actinide solubility in the glass and to understand the mechanisms controlling actinide solubilization. Glass specimens containing; actinide surrogates were used to prepare and optimize the fabrication of radioactive glass samples. These preliminary studies revealed that actinide Surrogates solubility in the glass was enhanced by controlling the processing temperature, the dissolution kinetic of the surrogate precursors, the glass composition and the oxidizing versus reducing conditions. The actinide solubility was investigated in the borosilicate glass. The evolution of thorium solubility in borosilicate glass was determined for temperatures ranging from 1200 deg C to 1400 deg C.Borosilicate glass specimens containing plutonium were fabricated. The experimental result showed that the plutonium solubility limit ranged from 1 to 2.5 wt% PuO 2 at 1200 deg C. A structural approach based on the determination of the local structure around actinides and their surrogates by EXAFS spectroscopy was used to determine their structural role in the glass and the nature of their bonding with the vitreous network. This approach revealed a correlation between the length of these bonds and the solubility of the actinides and their surrogates. (author)

Quantification of the boron speciation in alkali borosilicate glasses by electron energy loss spectroscopy

DEFF Research Database (Denmark)

Cheng, D.S.; Yang, G.; Zhao, Y.Q.

2015-01-01

developed a method based on electron energy loss spectroscopy (EELS) data acquisition and analyses, which enables determination of the boron speciation in a series of ternary alkali borosilicate glasses with constant molar ratios. A script for the fast acquisition of EELS has been designed, from which...

Investigation of local environment around rare earths (La and Eu) by fluorescence line narrowing during borosilicate glass alteration

Energy Technology Data Exchange (ETDEWEB)

Molières, Estelle [CEA – DEN-DTCD-LCV-SECM Laboratoire d' études du Comportement à Long Terme, 30207 Bagnols-sur-Cèze (France); Panczer, Gérard; Guyot, Yannick [Institut Lumière Matière, UMR5306 Université Lyon 1-CNRS, Université de Lyon, 69622 Villeurbanne cedex (France); Jollivet, Patrick [CEA – DEN-DTCD-LCV-SECM Laboratoire d' études du Comportement à Long Terme, 30207 Bagnols-sur-Cèze (France); Majérus, Odile; Aschehoug, Patrick; Barboux, Philippe [Laboratoire de Chimie de la Matière Condensée de Paris, UMR-CNRS 7574, École Nationale Supérieure de Chimie de Paris (ENSCP Chimie-ParisTech), 11 rue Pierre et Marie Curie, 75231 Paris (France); Gin, Stéphane [CEA – DEN-DTCD-LCV-SECM Laboratoire d' études du Comportement à Long Terme, 30207 Bagnols-sur-Cèze (France); Angeli, Frédéric, E-mail: frederic.angeli@cea.fr [CEA – DEN-DTCD-LCV-SECM Laboratoire d' études du Comportement à Long Terme, 30207 Bagnols-sur-Cèze (France)

2014-01-15

The local environment of europium in soda-lime borosilicate glasses with a range of La{sub 2}O{sub 3} content was probed by continuous luminescence and Fluorescence Line Narrowing (FLN) to investigate the local environment of rare earth elements in pristine and leached glass. After aqueous leaching at 90 °C at pH 7 and 9.5, rare earths were fully retained and homogeneously distributed in the amorphous alteration layer (commonly called gel). Two separate silicate environments were observed in pristine and leached glasses regardless of the lanthanum content and the leaching conditions. A borate environment surrounding europium was not observed in pristine and leached glasses. During glass alteration, OH groups were located around the europium environment, which became more organized (higher symmetry) in the first coordination shell. -- Highlights: • No borate environment surrounding europium was detected in pristine borosilicate glasses. • Up to 12 mol% of REE2O3 in glass, local environment of europium does not significantly change. • Europium environment becomes more ordered and symmetric in gels than in pristine glasses. • Two distinct silicate sites were observed, as well in pristine glass as in gels (leached glasses). • In altered glasses, OH groups were located around europium.

Investigation of local environment around rare earths (La and Eu) by fluorescence line narrowing during borosilicate glass alteration

International Nuclear Information System (INIS)

Molières, Estelle; Panczer, Gérard; Guyot, Yannick; Jollivet, Patrick; Majérus, Odile; Aschehoug, Patrick; Barboux, Philippe; Gin, Stéphane; Angeli, Frédéric

2014-01-01

The local environment of europium in soda-lime borosilicate glasses with a range of La 2 O 3 content was probed by continuous luminescence and Fluorescence Line Narrowing (FLN) to investigate the local environment of rare earth elements in pristine and leached glass. After aqueous leaching at 90 °C at pH 7 and 9.5, rare earths were fully retained and homogeneously distributed in the amorphous alteration layer (commonly called gel). Two separate silicate environments were observed in pristine and leached glasses regardless of the lanthanum content and the leaching conditions. A borate environment surrounding europium was not observed in pristine and leached glasses. During glass alteration, OH groups were located around the europium environment, which became more organized (higher symmetry) in the first coordination shell. -- Highlights: • No borate environment surrounding europium was detected in pristine borosilicate glasses. • Up to 12 mol% of REE2O3 in glass, local environment of europium does not significantly change. • Europium environment becomes more ordered and symmetric in gels than in pristine glasses. • Two distinct silicate sites were observed, as well in pristine glass as in gels (leached glasses). • In altered glasses, OH groups were located around europium

Intrinsic dosimetry. Properties and mechanisms of thermoluminescence in commercial borosilicate glass

Energy Technology Data Exchange (ETDEWEB)

Clark, Richard A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2012-10-01

Intrinsic dosimetry is the method of measuring total absorbed dose received by the walls of a container holding radioactive material. By considering the total absorbed dose received by a container in tandem with the physical characteristics of the radioactive material housed within that container, this method has the potential to provide enhanced pathway information regarding the history of the container and its radioactive contents. The latest in a series of experiments designed to validate and demonstrate this newly developed tool are reported. Thermoluminescence (TL) dosimetry was used to measure dose effects on raw stock borosilicate container glass up to 70 days after gamma ray, x-ray, beta particle or ultraviolet irradiations at doses from 0.15 to 20 Gy. The TL glow curve when irradiated with 60Co was separated into five peaks: two relatively unstable peaks centered near 120 and 165°C, and three relatively stable peaks centered near 225, 285, and 360°C. Depending on the borosilicate glass source, the minimum measurable dose using this technique is 0.15-0.5 Gy, which is roughly equivalent to a 24 hr irradiation at 1 cm from a 50-165 ng source of 60Co. Differences in TL glow curve shape and intensity were observed for the glasses from different geographical origins. These differences can be explained by changes in the intensities of the five peaks. Electron paramagnetic resonance (EPR) and multivariate statistical methods were used to relate the TL intensity and peaks to electron/hole traps and compositional variations.

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Influence of Glass Property Restrictions on Hanford HLW Glass Volume

International Nuclear Information System (INIS)

Kim, Dong-Sang; Vienna, John D.

2001-01-01

A systematic evaluation of Hanford High-Level Waste (HLW) loading in alkali-alumino-borosilicate glasses was performed. The waste feed compositions used were obtained from current tank waste composition estimates, Hanford's baseline retrieval sequence, and pretreatment processes. The waste feeds were sorted into groups of like composition by cluster analysis. Glass composition optimization was performed on each cluster to meet property and composition constraints while maximizing waste loading. Glass properties were estimated using property models developed for Hanford HLW glasses. The impacts of many constraints on the volume of HLW glass to be produced at Hanford were evaluated. The liquidus temperature, melting temperature, chromium concentration, formation of multiple phases on cooling, and product consistency test response requirements for the glass were varied one- or many-at-a-time and the resultant glass volume was calculated. This study shows clearly that the allowance of crystalline phases in the glass melter can significantly decrease the volume of HLW glass to be produced at Hanford.

Understanding of the mechanical and structural changes induced by alpha particles and heavy ions in the French simulated nuclear waste glass

Energy Technology Data Exchange (ETDEWEB)

Karakurt, G., E-mail: karakurt_gokhan@yahoo.fr [SUBATECH, UMR 6457CNRS-IN2P3, Ecole des Mines de Nantes, 4 rue Alfred Kastler, 44307 Nantes (France); Abdelouas, A. [SUBATECH, UMR 6457CNRS-IN2P3, Ecole des Mines de Nantes, 4 rue Alfred Kastler, 44307 Nantes (France); Guin, J.-P.; Nivard, M. [Institut de Physique de Rennes, Université de Rennes 1 – UMR 62051 IPR, 263 avenue du Général Leclerc, 35042 Rennes (France); Sauvage, T. [Laboratoire CEMHTI (Conditions Extrêmes et Matériaux: Haute Température et Irradiation), CNRS UPR, 3079 Orléans (France); Paris, M. [Institut des Matériaux Jean ROUXEL, Université de Nantes, UMR 6502 CNRS, 2 rue de la Houssinière, BP 32229, 44322 Nantes Cedex 03 (France); Bardeau, J.-F. [Institut des Molécules et Matériaux du Mans, UMR CNRS 6283, avenue Olivier Messiaen, 72085 Le Mans (France)

2016-07-15

Borosilicate glasses are considered for the long-term confinement of high-level nuclear wastes. External irradiations with 1 MeV He{sup +} ions and 7 MeV Au{sup 5+} ions were performed to simulate effects produced by alpha particles and by recoil nuclei in the simulated SON68 nuclear waste glass. To better understand the structural modifications, irradiations were also carried out on a 6-oxides borosilicate glass, a simplified version of the SON68 glass (ISG glass). The mechanical and macroscopic properties of the glasses were studied as function of the deposited electronic and nuclear energies. Alpha particles and gold ions induced a volume change up to −0.7% and −2.7%, respectively, depending on the glass composition. Nano-indentations tests were used to determine the mechanical properties of the irradiated glasses. A decrease of about −22% to −38% of the hardness and a decrease of the reduced Young's modulus by −8% were measured after irradiations. The evolution of the glass structure was studied by Raman spectroscopy, and also {sup 11}B and {sup 27}Al Nuclear Magnetic Resonance (MAS-NMR) on a 20 MeV Kr irradiated ISG glass powder. A decrease of the silica network connectivity after irradiation with alpha particles and gold ions is deduced from the structural changes observations. NMR spectra revealed a partial conversion of BO{sub 4} to BO{sub 3} units but also a formation of AlO{sub 5} and AlO{sub 6} species after irradiation with Kr ions. The relationships between the mechanical and structural changes are also discussed. - Highlights: • Mechanical and structural properties of two borosilicate glass compositions irradiated with alpha particles and heavy ions were investigated. • Both kinds of particles induced a decrease of the hardness, reduced Young's modulus and density. • Electronic and nuclear interactions are responsible for the changes observed. • The evolution of the mechanical properties under irradiation is linked

Small-scale, joule-heated melting of Savannah River Plant waste glass. I. Factors affecting large-scale vitrification tests

International Nuclear Information System (INIS)

Plodinec, M.J.; Chismar, P.H.

1979-10-01

A promising method of immobilizing SRP radioactive waste solids is incorporation in borosilicate glass. In the reference vitrification process, called joule-heated melting, a mixture of glass frit and calcined waste is heated by passage of an electric current. Two problems observed in large-scale tests are foaming and formation of an insoluble slag. A small joule-heated melter was designed and built to study problems such as these. This report describes the melter, identifies factors involved in foaming and slag formation, and proposes ways to overcome these problems

DEFENSE HIGH LEVEL WASTE GLASS DEGRADATION

International Nuclear Information System (INIS)

Ebert, W.

2001-01-01

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

Assessment of processes, facilities, and costs for alternative solid forms for immobilization of SRP defense waste

International Nuclear Information System (INIS)

Dunson, J.B. Jr.; Eisenberg, A.M.; Schuyler, R.L. III; Haight, H.G. Jr.; Mello, V.E.; Gould, T.H. Jr.; Butler, J.L.; Pickett, J.B.

1982-03-01

A quantitative merit evaluation which assesses the relative difficulty of remote processing of Savannah River Plant high-level wastes for seven alternative waste forms is presented. The reference borosilicate glass process is rated as the simplest, followed by FUETAP concrete. The other processes evaluated in order of increasing complexity were: glass marbles in a lead matrix, high-silica glass, crystalline ceramic (Synroc-D and tailored ceramic), and coated ceramic particles. Cost appraisals are summarized for the borosilicate glass, high-silica glass, and ceramic waste form processing facilities

Modelling the local atomic structure of molybdenum in nuclear waste glasses with ab initio molecular dynamics simulations.

Science.gov (United States)

Konstantinou, Konstantinos; Sushko, Peter V; Duffy, Dorothy M

2016-09-21

The nature of chemical bonding of molybdenum in high level nuclear waste glasses has been elucidated by ab initio molecular dynamics simulations. Two compositions, (SiO 2 ) 57.5 -(B 2 O 3 ) 10 -(Na 2 O) 15 -(CaO) 15 -(MoO 3 ) 2.5 and (SiO 2 ) 57.3 -(B 2 O 3 ) 20 -(Na 2 O) 6.8 -(Li 2 O) 13.4 -(MoO 3 ) 2.5 , were considered in order to investigate the effect of ionic and covalent components on the glass structure and the formation of the crystallisation precursors (Na 2 MoO 4 and CaMoO 4 ). The coordination environments of Mo cations and the corresponding bond lengths calculated from our model are in excellent agreement with experimental observations. The analysis of the first coordination shell reveals two different types of molybdenum host matrix bonds in the lithium sodium borosilicate glass. Based on the structural data and the bond valence model, we demonstrate that the Mo cation can be found in a redox state and the molybdate tetrahedron can be connected with the borosilicate network in a way that inhibits the formation of crystalline molybdates. These results significantly extend our understanding of bonding in Mo-containing nuclear waste glasses and demonstrate that tailoring the glass composition to specific heavy metal constituents can facilitate incorporation of heavy metals at high concentrations.

Physical and optical properties of lithium borosilicate glasses doped with Dy3+ ions

Science.gov (United States)

Ramteke, D. D.; Gedam, R. S.; Swart, H. C.

2018-04-01

The borosilicate glasses with Dy3+ ions were prepared by the melt quench technique with varying concentration of Dy2O3. The glasses were characterized by the density calculation, absorbance and photoluminescence (PL) spectroscopy measurements. Density and molar volume of the glasses increases with increase in Dy3+ ions in the glass matrix. This behavior is correlated with the higher molecular weight and larger ionic radius of Dy3+ ion compared to the other constituents of glass matrix. Emission of Dy3+ doped glasses showed three bands at 482, 573 and at 665 nm which correspond to 6H15/2 (blue), 6H13/2 (yellow) and 6H11/2 (red) transitions. The emission spectra of glasses with different concentration of Dy3+ ions shows that, glasses with 0.5 mol% of Dy2O3 shows highest emission and decreases with further doping. CIE 1931 chromaticity diagram showed that the emission of these glasses was in the white region. Photographs of these glasses under 349 nm Light emitting diode excitation also confirmed the white light emission from these glasses.

Study of phase separation and crystallization phenomena in soda-lime borosilicate glass enriched in MoO3

International Nuclear Information System (INIS)

Magnin, M.

2009-09-01

Molybdenum oxide immobilization (MoO 3 , as fission product) is one of the major challenges in the nuclear glass formulation issues for high level waste solutions conditioning since many years, these solutions arising from spent nuclear fuel reprocessing. Phase separation and crystallisation processes may arise in molten glass when the MoO 3 content is higher than its solubility limit that may depend on glass composition. Molybdenum combined with other elements such as alkali and alkaline-earth may form crystalline molybdates, known as 'yellow phases' in nuclear glasses which may decrease the glass durability. In order to confine high level wastes (HLW) such as the fission product solutions arising from the reprocessing of high burn-up UOX-type nuclear spent fuels, a new glass composition (HLW glass) is being optimized. This work is devoted to the study of the origin and the mechanism of phase separation and crystallization phenomena induced by molybdenum oxide incorporation in the HLW glass. From microstructural and structural point of view, the molybdenum oxide behavior was studied in glass compositions belonging to the SiO 2 -B 2 O 3 - Na 2 O-CaO simplified system which constituted basis for the HLW glass formulation. The structural role of molybdenum oxide in borosilicate network explaining the phase separation and crystallization tendency was studied through the coupling of structural ( 95 Mo, 29 Si, 11 B, 23 Na MAS NMR, XRD) and microstructural (SEM, HRTEM) analysis techniques. The determination of phase separation (critical temperature) and crystallization (liquidus temperature) appearance temperatures by in situ viscosimetry and Raman spectroscopy experiments allowed us to propose a transformation scenario during melt cooling. These processes and the nature of the crystalline phases formed (CaMoO 4 , Na 2 MoO 4 ) that depend on the evolution of MoO 3 , CaO and B 2 O 3 contents were correlated with changes of sodium and calcium cations proportions in the

Deformation mechanisms during nanoindentation of sodium borosilicate glasses of nuclear interest

Energy Technology Data Exchange (ETDEWEB)

Kilymis, D. A.; Delaye, J.-M., E-mail: jean-marc.delaye@cea.fr [CEA Marcoule, DEN/DTCD, Service d’Etude et Comportement des Matériaux de Conditionnement, BP17171 30207 Bagnols-sur-Cèze Cedex (France)

2014-07-07

In this paper we analyze results of Molecular Dynamics simulations of Vickers nanoindentation, performed for sodium borosilicate glasses of interest in the nuclear industry. Three glasses have been studied in their pristine form, as well as a disordered one that is analogous to the real irradiated glass. We focused in the behavior of the glass during the nanoindentation in order to reveal the mechanisms of deformation and how they are affected by microstructural characteristics. Results have shown a strong dependence on the SiO{sub 2} content of the glass, which promotes densification due to the open structure of SiO{sub 4} tetrahedra and also due to the strength of Si-O bonds. Densification for the glasses is primarily expressed by the relative decrease of the Si-O-Si and Si-O-B angles, indicating rotation of the structural units and decrease of free volume. The increase of alkali content on the other hand results to higher plasticity of the matrix and increased shear flow. The most important effect on the deformation mechanism of the disordered glasses is that of the highly depolymerized network that will also induce shear flow and, in combination with the increased free volume, will result in the decreased hardness of these glasses, as has been previously observed.

Deformation mechanisms during nanoindentation of sodium borosilicate glasses of nuclear interest.

Science.gov (United States)

Kilymis, D A; Delaye, J-M

2014-07-07

In this paper we analyze results of Molecular Dynamics simulations of Vickers nanoindentation, performed for sodium borosilicate glasses of interest in the nuclear industry. Three glasses have been studied in their pristine form, as well as a disordered one that is analogous to the real irradiated glass. We focused in the behavior of the glass during the nanoindentation in order to reveal the mechanisms of deformation and how they are affected by microstructural characteristics. Results have shown a strong dependence on the SiO2 content of the glass, which promotes densification due to the open structure of SiO4 tetrahedra and also due to the strength of Si-O bonds. Densification for the glasses is primarily expressed by the relative decrease of the Si-O-Si and Si-O-B angles, indicating rotation of the structural units and decrease of free volume. The increase of alkali content on the other hand results to higher plasticity of the matrix and increased shear flow. The most important effect on the deformation mechanism of the disordered glasses is that of the highly depolymerized network that will also induce shear flow and, in combination with the increased free volume, will result in the decreased hardness of these glasses, as has been previously observed.

Waste glass corrosion modeling: Comparison with experimental results

International Nuclear Information System (INIS)

Bourcier, W.L.

1994-01-01

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

Segregation of the elements of the platinum group in a simulated high-level waste glass

International Nuclear Information System (INIS)

Mitamura, H.; Banba, T.; Kamizono, H.; Kiriyama, Y.; Kumata, M.; Murakami, T.; Tashiro, S.

1983-01-01

Segregation of the elements of the platinum group occurred during vitrification of the borosilicate glass containing 20 wt% simulated high-level waste oxides. The segregated materials were composed of two crystalline phases: one was the solid solution of ruthenium and rhodium dioxides and the other was that of palladium and rhodium metals also with tellurium. The segregated materials were not distributed homogeneously throughout the glass: (i) on the surface of the glass, there occurred palladium, rhodium and tellurium alloy alone; and (ii) at the inner part of the glass, the agglomerates of the two phases were concentrated in one part and dispersed in the other

Application of PCT to the EBR II ceramic waste form

International Nuclear Information System (INIS)

Ebert, W. L.; Lewis, M. A.; Johnson, S. G.

2002-01-01

We are evaluating the use of the Product Consistency Test (PCT) developed to monitor the consistency of borosilicate glass waste forms for application to the multiphase ceramic waste form (CWF) that will be used to immobilize waste salts generated during the electrometallurgical conditioning of spent sodium-bonded nuclear fuel from the Experimental Breeder Reactor No. 2 (EBR II). The CWF is a multiphase waste form comprised of about 70% sodalite, 25% borosilicate glass binder, and small amounts of halite and oxide inclusions. It must be qualified for disposal as a non-standard high-level waste (HLW) form. One of the requirements in the DOE Waste Acceptance System Requirements Document (WASRD) for HLW waste forms is that the consistency of the waste forms be monitored.[1] Use of the PCT is being considered for the CWF because of the similarities of the dissolution behaviors of both the sodalite and glass binder phases in the CWF to borosilicate HLW glasses. This paper provides (1) a summary of the approach taken in selecting a consistency test for CWF production and (2) results of tests conducted to measure the precision and sensitivity of the PCT conducted with simulated CWF

Polyphase ceramic and glass-ceramic forms for immobilizing ICPP high-level nuclear waste

International Nuclear Information System (INIS)

Harker, A.B.; Flintoff, J.F.

1984-01-01

Polyphase ceramic and glass-ceramic forms have been consolidated from simulated Idaho Chemical Processing Plant wastes by hot isostatic pressing calcined waste and chemical additives by 1000 0 C or less. The ceramic forms can contain over 70 wt% waste with densities ranging from 3.5 to 3.85 g/cm 3 , depending upon the formulation. Major phases are CaF 2 , CaZrTi 207 , CaTiO 3 , monoclinic ZrO 2 , and amorphous intergranular material. The relative fraction of the phases is a function of the chemical additives (TiO 2 , CaO, and SiO 2 ) and consolidation temperature. Zirconolite, the major actinide host, makes the ceramic forms extremely leach resistant for the actinide simulant U 238 . The amorphous phase controls the leach performance for Sr and Cs which is improved by the addition of SiO 2 . Glass-ceramic forms were also consolidated by HIP at waste loadings of 30 to 70 wt% with densities of 2.73 to 3.1 g/cm 3 using Exxon 127 borosilicate glass frit. The glass-ceramic forms contain crystalline CaF 2 , Al 203 , and ZrSi 04 (zircon) in a glass matrix. Natural mineral zircon is a stable host for 4+ valent actinides. 17 references, 3 figures, 5 tables

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

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

β-Irradiation Effects on the Formation and Stability of CaMoO4 in a Soda Lime Borosilicate Glass Ceramic for Nuclear Waste Storage.

Science.gov (United States)

Patel, Karishma B; Boizot, Bruno; Facq, Sébastien P; Lampronti, Giulio I; Peuget, Sylvain; Schuller, Sophie; Farnan, Ian

2017-02-06

Molybdenum solubility is a limiting factor to actinide loading in nuclear waste glasses, as it initiates the formation of water-soluble crystalline phases such as alkali molybdates. To increase waste loading efficiency, alternative glass ceramic structures are sought that prove resistant to internal radiation resulting from radioisotope decay. In this study, selective formation of water-durable CaMoO 4 in a soda lime borosilicate is achieved by introducing up to 10 mol % MoO 3 in a 1:1 ratio to CaO using a sintering process. The resulting homogeneously dispersed spherical CaMoO 4 nanocrystallites were analyzed using electron microscopy, X-ray diffraction (XRD), Raman and electron paramagnetic resonance (EPR) spectroscopies prior to and post irradiation, which replicated internal β-irradiation damage on an accelerated scale. Following 0.77 to 1.34 GGy of 2.5 MeV electron radiation CaMoO 4 does not exhibit amorphization or significant transformation. Nor does irradiation induce glass-in-glass phase separation in the surrounding amorphous matrix, or the precipitation of other molybdates, thus proving that excess molybdenum can be successfully incorporated into a structure that it is resistant to β-irradiation proportional to 1000 years of storage without water-soluble byproducts. The CaMoO 4 crystallites do however exhibit a nonlinear Scherrer crystallite size pattern with dose, as determined by a Rietveld refinement of XRD patterns and an alteration in crystal quality as deduced by anisotropic peak changes in both XRD and Raman spectroscopy. Radiation-induced modifications in the CaMoO 4 tetragonal unit cell occurred primarily along the c-axis indicating relaxation of stacked calcium polyhedra. Concurrently, a strong reduction of Mo 6+ to Mo 5+ during irradiation is observed by EPR, which is believed to enhance Ca mobility. These combined results are used to hypothesize a crystallite size alteration model based on a combination of relaxation and diffusion

Synroc tailored waste forms for actinide immobilization

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-01

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

Direct conversion of plutonium-containing materials to borosilicate glass for storage or disposal

International Nuclear Information System (INIS)

Forsberg, C.W.; Beahm, E.C.

1995-01-01

A new process, the Glass Material Oxidation and Dissolution System (GMODS), has been invented for the direct conversion of plutonium metal, scrap, and residue into borosilicate glass. The glass should be acceptable for either the long-term storage or disposition of plutonium. Conversion of plutonium from complex chemical mixtures and variable geometries into homogeneous glass (1) simplifies safeguards and security; (2) creates a stable chemical form that meets health, safety, and environmental concerns; (3) provides an easy storage form; (4) may lower storage costs; and (5) allows for future disposition options. In the GMODS process, mixtures of metals, ceramics, organics, and amorphous solids containing plutonium are fed directly into a glass melter where they are directly converted to glass. Conventional glass melters can accept materials only in oxide form; thus, it is its ability to accept materials in multiple chemical forms that makes GMODS a unique glass making process. Initial proof-of-principle experiments have converted cerium (plutonium surrogate), uranium, stainless steel, aluminum, and other materials to glass. Significant technical uncertainties remain because of the early nature of process development

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

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

Upconversion studies of Er3+/Yb3+ doped SrO.TiO2 borosilicate glass ceramic system

International Nuclear Information System (INIS)

Maheshwari, Aditya; Om Prakash; Kumar, Devendra; Rai, S.B.

2011-01-01

Upconversion behaviour has been studied in various matrices and fine powders of SrTiO 3 by previous workers. In present work, Er 3+ /Yb 3+ were doped in appropriate ratio in SrO.TiO 2 borosilicate glass ceramic system to study the upconversion phenomenon. Dielectric properties of this class of glass ceramic system have been extensively investigated by Thakur et al. It has been observed that both upconversion efficiency and dielectric constant increases with transformation of glass into glass ceramic. Therefore, present investigation is based upon the study of optical as well as the electrical properties of same glass ceramic system. In order to prepare different crystalline matrices, two different Er 3+ /Yb 3+ :SrO.TiO 2 borosilicate glasses with same amount of Er 2 O 3 and Yb 2 O 3 were prepared by melt quench method. Glasses were transparent with light-wine colour. Glass ceramics were prepared from the glasses by heat treatment based on DTA (Differential thermal analysis) results. Glass ceramics were fully opaque with brownish-cream colour. Powder X-ray diffraction (XRD) patterns confirmed that two different crystalline matrices, Sr 3 Ti 2 O 7 , Ti 10 O 19 and SrTiO 3 , TiO 2 were present in two glass ceramic samples respectively. Luminescence properties of glass and glass ceramic samples with 976nm laser irradiation showed that the intensities of the green and red emission increased multiple times in glass ceramic than that of the glass. Possible mechanisms responsible for upconversion eg. Energy Transfer (ET) and Excited State Absorption (ESA), were studied through laser pumping power log dependence

Radiation transport in high-level waste form

International Nuclear Information System (INIS)

Arakali, V.S.; Barnes, S.M.

1992-01-01

The waste form selected for vitrifying high-level nuclear waste stored in underground tanks at West Valley, NY is borosilicate glass. The maximum radiation level at the surface of a canister filled with the high-level waste form is prescribed by repository design criteria for handling and disposition of the vitrified waste. This paper presents an evaluation of the radiation transport characteristics for the vitreous waste form expected to be produced at West Valley and the resulting neutron and gamma dose rates. The maximum gamma and neutron dose rates are estimated to be less than 7500 R/h and 10 mRem/h respectively at the surface of a West Valley canister filled with borosilicate waste glass

Photoelastic examination of borosilicate glass discs used in the insulating legs for the NSF tandem

International Nuclear Information System (INIS)

Acton, W.J.; Cundy, D.

1981-04-01

The results are presented of a photoelastic stress analysis carried out to establish the effect of re-annealing borosilicate glass discs used in the insulating legs of the 30 MV tandem van de Graaff accelerator of the NSF. The results show that re-annealing of the glass discs has no measurable effect on reducing the high stress at inclusions and re-emphasise the need to exercise great care in selecting suitable discs for use in the insulating legs. (U.K.)

Aqueous solutions/nuclear glasses interactions

International Nuclear Information System (INIS)

Delage, F.; Advocat, T.; Vernaz, E.; Crovisier, J.L.

1991-01-01

Interactions results of the borosilicate glass used in radioactive wastes confinement and aqueous solutions at various temperature and PH show that for the glass components: - the release rate evolution follows an Arrhenius law, - in acid PH, there is a selective dissolution, - in basic PH, there is a stoechiometric dissolution [fr

Waste vitrification: a historical perspective

International Nuclear Information System (INIS)

McElroy, J.L.; Bjorklund, W.J.; Bonner, W.F.

1982-02-01

The possibility of converting high-level wastes (HLW) to glass was first pursued in Canada and England at a time when other countries were evaluating many other alternatives. By 1966, the British had completed radioactive demonstrations of the FINGAL pot process, converting HLW to borosilicate glass. By this time other countries, including France and the United States, had begun using the glass waste form. Beginning in 1966, several processes, including phosphate and borosilicate glass, were demonstrated by the US in the Waste Solidification Engineering Prototypes (WSEP) program at the Pacific Northwest Laboratory (PNL). Most of the current vitrification processes are adaptations of the FINGAL pot process or the continuous metallic melter used in the WSEP program. One notable exception is the joule-heated ceramic melter, which was adapted from commercial glass technology for HLW by PNL in the mid-1970's. Both batch and continuous processes have been developed to an advanced stage of readiness. These processes are described and compared in this paper

Development, evaluation, and selection of candidate high-level waste forms

International Nuclear Information System (INIS)

Bernadzikowski, T.A.; Allender, J.S.; Gordon, D.E.; Gould, T.H. Jr.

1982-01-01

The seven candidate waste forms, evaluated as potential media for the immobilization and gelogic disposal of high-level nuclear wastes were borosilicate glass, SYNROC, tailored ceramic, high-silica glass, FUETAP concrete, coated sol-gel particles, and glass marbles in a lead matrix. The evaluation, completed on August 1, 1981, combined preliminary waste form evaluations conducted at Department of Energy (DOE) defense waste-sites and at independent laboratories, peer review assessments, a product performance evaluation, and a processability analysis. Based on the combined results of these four inputs, two of the seven forms, borosilicate glass and a titanate-based ceramic, SYNROC, were selected as the reference and alternative forms, respectively, for continued development and evaluation in the National HLW Program. The borosilicate glass and ceramic forms were further compared during FY-1982 on the basis of risk assessments, cost comparisons, properties comparisons, and conformance with proposed regulatory and repository criteria. Both the glass and ceramic forms are viable candidates for use at DOE defense HLW sites; they are also candidates for immobilization of commercial reprocessing wastes. This paper describes the waste form screening process, discusses each of the four major inputs considered in the selection of the two forms in 1981, and presents a brief summary of the comparisons of the two forms during 1982 and the selection process to determine the final form for SRP defense HLW

Corrosion testing of a plutonium-loaded lanthanide borosilicate glass made with Frit B.

Energy Technology Data Exchange (ETDEWEB)

Ebert, W. L.; Chemical Engineering

2006-09-30

Laboratory tests were conducted with a lanthanide borosilicate (LaBS) glass made with Frit B and added PuO2 (the glass is referred to herein as Pu LaBS-B glass) to measure the dependence of the glass dissolution rate on pH and temperature. These results are compared with the dependencies used in the Defense HLW Glass Degradation Model that was developed to account for HLW glasses in total system performance assessment (TSPA) calculations for the Yucca Mountain repository to determine if that model can also be used to represent the release of radionuclides from disposed Pu LaBS glass by using either the same parameter values that are used for HLW glasses or parameter values specific for Pu LaBS glass. Tests were conducted by immersing monolithic specimens of Pu LaBS-B glass in six solutions that imposed pH values between about pH 3.5 and pH 11, and then measuring the amounts of glass components released into solution. Tests were conducted at 40, 70, and 90 C for 1, 2, 3, 4, and 5 days at low glass-surface-area-to-solution volume ratios. As intended, these test conditions maintained sufficiently dilute solutions that the impacts of solution feedback effects on the dissolution rates were negligible in most tests. The glass dissolution rates were determined from the concentrations of Si and B measured in the test solutions. The dissolution rates determined from the releases of Si and B were consistent with the 'V' shaped pH dependence that is commonly seen for borosilicate glasses and is included in the Defense HLW Glass Degradation Model. The rate equation in that model (using the coefficients determined for HLW glasses) provides values that are higher than the Pu LaBS-B glass dissolution rates that were measured over the range of pH and temperature values that were studied (i.e., an upper bound). Separate coefficients for the rate expression in acidic and alkaline solutions were also determined from the test results to model Pu LaBS-B glass dissolution

Entrapment of 137Cs vapour generated during vitrification and casting of cesium borosilicate glass by inorganic materials

International Nuclear Information System (INIS)

Ram, Ramu; Gandhi, Shyamala; Dash, A.; Varma, R.N.

2003-01-01

Efficiency of different inorganic materials like zirconium antimonate (ZrA), ammonium molybdophosphate (AMP), synthetic zeolites, activated charcoal, glass wool etc, towards the entrapment of 137 Cs vapour escaping during vitrification and casting of cesium borosilicate glass required for the preparation of 137 Cs sources for medical and industrial applications have been determined. The recovery of entrapped cesium using dilute acids for subsequent recycling has also been explored. (author)

Development of iodine waste forms using low-temperature sintering glass

International Nuclear Information System (INIS)

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

2010-01-01

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

Composition-Structure-Property Relations of Compressed Borosilicate Glasses

Science.gov (United States)

Svenson, Mouritz N.; Bechgaard, Tobias K.; Fuglsang, Søren D.; Pedersen, Rune H.; Tjell, Anders Ø.; Østergaard, Martin B.; Youngman, Randall E.; Mauro, John C.; Rzoska, Sylwester J.; Bockowski, Michal; Smedskjaer, Morten M.

2014-08-01

Hot isostatic compression is an interesting method for modifying the structure and properties of bulk inorganic glasses. However, the structural and topological origins of the pressure-induced changes in macroscopic properties are not yet well understood. In this study, we report on the pressure and composition dependences of density and micromechanical properties (hardness, crack resistance, and brittleness) of five soda-lime borosilicate glasses with constant modifier content, covering the extremes from Na-Ca borate to Na-Ca silicate end members. Compression experiments are performed at pressures ≤1.0 GPa at the glass transition temperature in order to allow processing of large samples with relevance for industrial applications. In line with previous reports, we find an increasing fraction of tetrahedral boron, density, and hardness but a decreasing crack resistance and brittleness upon isostatic compression. Interestingly, a strong linear correlation between plastic (irreversible) compressibility and initial trigonal boron content is demonstrated, as the trigonal boron units are the ones most disposed for structural and topological rearrangements upon network compaction. A linear correlation is also found between plastic compressibility and the relative change in hardness with pressure, which could indicate that the overall network densification is responsible for the increase in hardness. Finally, we find that the micromechanical properties exhibit significantly different composition dependences before and after pressurization. The findings have important implications for tailoring microscopic and macroscopic structures of glassy materials and thus their properties through the hot isostatic compression method.

Characterization of projected DWPF glasses heat treated to simulate canister centerline cooling

International Nuclear Information System (INIS)

Marra, S.L.; Jantzen, C.M.

1992-05-01

Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). Eventually these canistered waste forms will be sent to a geologic repository for final disposal. In order to assure acceptability by the repository, the Department of Energy has defined requirements which DWPF canistered waste forms must meet. These requirements are the Waste Acceptance Preliminary Specifications (WAPS). The WAPS require DWPF to identify the crystalline phases expected to be present in the final glass product. Knowledge of the thermal history of the borosilicate glass during filling and cooldown of the canister is necessary to determine the amount and type of crystalline phases present in the final glass product. Glass samples of seven projected DWPF compositions were cooled following the same temperature profile as that of glass at the centerline of the full-scale DWPF canister. The glasses were characterized by x-ray diffraction and scanning electron microscopy to identify the crystalline phases present The volume percents of each crystalline phase present were determined by quantitative x-ray diffraction. The Product Consistency Test (PCI) was used to determine the durability of the heat-treated glasses

3 and 4 oxidation state element solubilities in borosilicate glasses. Implement to actinides in nuclear glasses; Solubilite des elements aux degres d'oxydation (3) et (4) dans les verres de borosilicate. Application aux actinides dans les verres nucleaires

Energy Technology Data Exchange (ETDEWEB)

Cachia, J.N

2005-12-15

In order to ensure optimal radionuclides containment, the knowledge of the actinide loading limits in nuclear waste glasses and also the comprehension of the solubilization mechanisms of these elements are essential. A first part of this manuscript deals with the study of the differences in solubility of the tri and tetravalent elements (actinides and surrogates) particularly in function of the melting temperature. The results obtained indicate that trivalent elements (La, Gd, Nd, Am, Cm) exhibit a higher solubility than tetravalent elements (Hf, Th, Pu). Consequently, it was planned to reduce plutonium at the oxidation state (III), the later being essentially tetravalent in borosilicate glasses. An innovating reduction process of multi-valent elements (cerium, plutonium) using silicon nitride has been developed in a second part of this work. Reduced plutonium-bearing glasses synthesized by Si{sub 3}N{sub 4} addition made it possible to double the plutonium solubility from 2 to 4 wt% at 1200 deg C. A structural approach to investigate the differences between tri and tetravalent elements was finally undertaken. These investigations were carried out by X-rays Absorption Spectroscopy (EXAFS) and NMR. Trivalent rare earth and actinide elements seem to behave as network modifiers while tetravalent elements rather present true intermediaries' behaviour. (author)

Determination of the free enthalpies of formation of borosilicate glasses; Determination des enthalpies libres de formation des verres borosilicates. Application a l'etude de l'alteration des verres de confinement de dechets radioactifs

Energy Technology Data Exchange (ETDEWEB)

Linard, Y

2000-07-01

This work contributes to the study of the thermochemical properties of nuclear waste glasses. Results are used to discuss mechanisms and parameters integrated in alteration models of conditioning materials. Glass is a disordered material defined thermodynamically as a non-equilibrium state. Taking into account one order parameter to characterise its configurational state, the metastable equilibrium for the glass was considered and the main thermochemical properties were determined. Calorimetric techniques were used to measure heat capacities and formation enthalpies of borosilicate glasses (from 3 to 8 constitutive oxides). Formation Entropies were measured too, using the entropy theory of relaxation processes proposed by Adam and Gibbs (1965). The configurational entropy contribution were determined from viscosity measurements. This set of data has allowed the calculation of Gibb's free energies of dissolution of glasses in pure water. By comparison with leaching experiments, it has been demonstrated that the decreasing of the dissolution rate at high reaction progress cannot be associated to the approach of an equilibrium between the sound glass and the aqueous solution. The composition changes of the reaction area at the glass surface need to be considered too. To achieve a complete description of the thermodynamic stability, the equilibrium between hydrated de-alkalinized glass and/or the gel layer with the aqueous solution should also be evaluated. (author)

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

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)

Development and evaluation of candidate high-level waste forms

International Nuclear Information System (INIS)

Bernadzikowski, T.A.

1981-01-01

Some seventeen candidate waste forms have been investigated under US Department of Energy programs as potential media for the immobilization and geologic disposal of the high-level radioactive wastes (HLW) resulting from chemical processing of nuclear reactor fuels and targets. Two of these HLW forms were selected at the end of fiscal year (FY) 1981 for intensive development if FY 1982 to 1983. Borosilicate glass was continued as the reference form. A crystalline ceramic waste form, SYNROC, was selected for further product formulation and process development as the alternative to borosilicate glass. This paper describes the bases on which this decision was made

Color-converted remote phosphor prototype of a multiwavelength excitable borosilicate glass for white light-emitting diodes

International Nuclear Information System (INIS)

Tian Hua; Qiu Kun; Song Jun; Wang Da-Jian; Liu Ji-Wen

2012-01-01

We report a unique red light-emitting Eu-doped borosilicate glass to convert color for warm white light-emitting diodes. This glass can be excited from 394 nm-peaked near ultraviolet light, 466 nm-peaked blue light, to 534 nm-peaked green light to emit the desired red light with an excellent transmission in the wavelength range of 400–700 nm which makes this glass suitable for color conversion without a great cost of luminous power loss. In particular, when assembling this glass for commercial white light-emitting diodes, the tested results show that the color rendering index is improved to 84 with a loss of luminous power by 12 percent at average, making this variety of glass promising for inorganic “remote-phosphor” color conversion

Leach rate studies on glass containing actual radioactive waste

International Nuclear Information System (INIS)

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

1980-01-01

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

Mercury reduction and removal during high-level radioactive waste processing and vitrification

International Nuclear Information System (INIS)

Eibling, R.E.; Fowler, J.R.

1981-01-01

A reference process for immobilizing the high-level radioactive waste in borosilicate glass has been developed at the Savannah River Plant. This waste contains a substantial amount of mercury from separations processing. Because mercury will not remain in borosilicate glass at the processing temperature, mercury must be removed before vitrification or must be handled in the off-gas system. A process has been developed to remove mercury by reduction with formic acid prior to vitrification. Additional benefits of formic acid treatment include improved sludge handling and glass melter redox control

Comparison of radiation and quenching rate effects on the structure of a sodium borosilicate glass

International Nuclear Information System (INIS)

Peuget, Sylvain; Maugeri, Emilio-Andrea; Mendoza, Clement; Fares, Toby; Bouty, Olivier; Jegou, Christophe; Charpentier, Thibault; Moskura, Melanie

2013-01-01

The effects of quenching rate and irradiation on the structure of a sodium borosilicate glass were compared using 29 Si, 11 B, and 23 Na nuclear magnetic resonance and Raman spectroscopy. Quenching rate ranging from 0.1 to 3 * 10 4 K min -1 was studied. Various irradiation conditions were performed, i.e. gold-ion irradiation in a multi-energy mode (from 1 to 6.75 MeV), and Kr and Xe ion irradiations with energy of 74 and 92 MeV, respectively. In pile irradiation with thermal neutron flux was performed as well, to study the effect of alpha radiation from the nuclear reaction 10 B(n,α) 7 Li. Both irradiation and high quenching rate induce similar local order modification of the glass structure, mainly a decrease of the mean boron coordination and an increase of Q 3 units. Nevertheless, the variations observed under irradiation are more pronounced than the ones induced by the quenching rate. Moreover, some important modifications of the glass medium range order, i.e. the emergence of the D2 band associated to three members silica rings and a modification of the Si-O-Si angle distribution were only noticed after irradiation. These results suggest that the irradiated structure is certainly not exactly the one obtained by a rapidly quenched equilibrated melt, but rather a more disordered structure that was weakly relaxed during the very rapid quenching phase following the energy deposition step. Raman spectroscopy showed a similar irradiated structure whereas the glass evolutions were controlled by the electronic energy loss in the ion track formation regime for Kr-ion irradiation or by the nuclear energy loss for Au and OSIRIS irradiation. The similar irradiated structure despite different irradiation routes, suggests that the final structural state of this sodium borosilicate glass is mainly controlled by the glass reconstruction after the energy deposition step. (authors)

Preparation and nonlinear optical properties of indium nanocrystals in sodium borosilicate glass by the sol–gel route

International Nuclear Information System (INIS)

Zhong, Jiasong; Xiang, Weidong; Zhao, Haijun; Chen, Zhaoping; Liang, Xiaojuan; Zhao, Wenguang; Chen, Guoxin

2012-01-01

Graphical abstract: The sodium borosilicate glass doped with indium nanocrystals have been successfully prepared by sol–gel methods. And the indium nanocrystals in tetragonal crystal system have formed uniformly in the glass, and the average diameter of indium nanocrystals is about 30 nm. The third-order optical nonlinear refractive index γ, absorption coefficient β, and susceptibility χ (3) of the glass are determined to be −4.77 × 10 −16 m 2 /W, 2.67 × 10 −9 m/W, and 2.81 × 10 −10 esu, respectively. Highlights: ► Indium nanocrystals embedded in glass matrix have been prepared by sol–gel route. ► The crystal structure and composition are investigated by XRD and XPS. ► Size and distribution of indium nanocrystals is determined by TEM. ► The third-order optical nonlinearity is investigated by using Z-scan technique. -- Abstract: The sodium borosilicate glass doped with indium nanocrystals have been successfully prepared by sol–gel route. The thermal stability behavior of the stiff gel is investigated by thermogravimetric (TG) and differential thermal (DTA) analysis. The crystal structure of the glass is characterized by X-ray powder diffraction (XRD). Particle composition is determined by X-ray photoelectron spectroscopy (XPS). Size and distribution of the nanocrystals are characterized by transmission electron microscopy (TEM) as well as high-resolution transmission electron microscopy (HRTEM). Results show that the indium nanocrystals in tetragonal crystal structure have formed in glass, and the average diameter is about 30 nm. Further, the glass is measured by Z-scan technique to investigate the nonlinear optical (NLO) properties. The third-order NLO coefficient χ (3) of the glass is determined to be 2.81 × 10 −10 esu. The glass with large third-order NLO coefficient is promising materials for applications in optical devices.

Proving the role of boron in the structure of fly-ash/borosilicate glass based geopolymers

Czech Academy of Sciences Publication Activity Database

Taveri, Gianmarco; Toušek, J.; Bernardo, E.; Toniolo, N.; Boccaccini, A. R.; Dlouhý, Ivo

2017-01-01

Roč. 200, AUG (2017), s. 105-108 ISSN 0167-577X EU Projects: European Commission(XE) 642557 - CoACH Institutional support: RVO:68081723 Keywords : Fly-ash * Borosillicate * Geopolymerization * Spectroscopy * NMR * FTIR Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass OBOR OECD: Ceramics Impact factor: 2.572, year: 2016 http://www.sciencedirect.com/science/article/pii/S0167577X1730650X

Description of processes for the immobilization of selected transuranic wastes

International Nuclear Information System (INIS)

Timmerman, C.L.

1980-12-01

Processed sludge and incinerator-ash wastes contaminated with transuranic (TRU) elements may require immobilization to prevent the release of these elements to the environment. As part of the TRU Waste Immobilization Program sponsored by the Department of Energy (DOE), the Pacific Northwest Laboratory is developing applicable waste-form and processing technology that may meet this need. This report defines and describes processes that are capable of immobilizing a selected TRU waste-stream consisting of a blend of three parts process sludge and one part incinerator ash. These selected waste streams are based on the compositions and generation rates of the waste processing and incineration facility at the Rocky Flats Plant. The specific waste forms that could be produced by the described processes include: in-can melted borosilicate-glass monolith; joule-heated melter borosilicate-glass monolith or marble; joule-heated melter aluminosilicate-glass monolith or marble; joule-heated melter basaltic-glass monolith or marble; joule-heated melter glass-ceramic monolith; cast-cement monolith; pressed-cement pellet; and cold-pressed sintered-ceramic pellet

Influence of iron ions on the structural properties of Zn-borosilicate glasses

International Nuclear Information System (INIS)

Music, S.; Gotic, M.; Popovic, S.

1989-01-01

The objectives of this study were to determine the influence of iron ions on the appearance of crystalline phases in Zn-borosilicate glasses, and to obtain information about valence state and coordination of iron ions. The systems Na 2 O-ZnO-B 2 O 3 -SiO 2 and ZnO-B 2 O 3 -SiO 2 were doped with α-Fe 2 O 3 . X-ray diffraction, IR spectroscopy and 57 Fe Moessbauer spectroscopy were used as experimental techniques. (author) 32 refs.; 5 figs.; 5 tabs

A Raman spectroscopy study on the effects of intermolecular hydrogen bonding on water molecules absorbed by borosilicate glass surface

Science.gov (United States)

Li, Fabing; Li, Zhanlong; Wang, Ying; Wang, Shenghan; Wang, Xiaojun; Sun, Chenglin; Men, Zhiwei

2018-05-01

The structural forms of water/deuterated water molecules located on the surface of borosilicate capillaries have been first investigated in this study on the basis of the Raman spectral data obtained at different temperatures and under atmospheric pressure for molecules in bulk and also for molecules absorbed by borosilicate glass surface. The strongest two fundamental bands locating at 3063 cm-1 (2438 cm-1) in the recorded Raman spectra are assigned here to the Osbnd H (Osbnd D) bond stretching vibrations and they are compared with the corresponding bands observed at 3124 cm-1 (2325 cm-1) in the Raman spectrum of ice Ih. Our spectroscopic observations have indicated that the structure of water and deuterated water molecules on borosilicate surface is similar to that of ice Ih (hexagonal phase of ice). These observations have also indicated that water molecules locate on the borosilicate surface so as to construct a bilayer structure and that strong and weak intermolecular hydrogen bonds are formed between water/deuterated molecules and silanol groups on borosilicate surface. In accordance with these findings, water and deuterated water molecules at the interface of capillary have a higher melting temperature.

Wetspun poly-L-(lactic acid)-borosilicate bioactive glass scaffolds for guided bone regeneration

Energy Technology Data Exchange (ETDEWEB)

Fernandes, João S., E-mail: joao.fernandes@dep.uminho.pt [3B' s Research Group - Biomaterials, Biodegradables and Biomimetics, University of Minho, Headquarters of the European Institute of Excellence on Tissue Engineering and Regenerative Medicine, AvePark-Parque de Ciência e Tecnologia, Zona Industrial da Gandra, 4805-017 Barco GMR (Portugal); ICVS/3B' s - PT Government Associate Laboratory, Braga/Guimarães (Portugal); Reis, Rui L. [3B' s Research Group - Biomaterials, Biodegradables and Biomimetics, University of Minho, Headquarters of the European Institute of Excellence on Tissue Engineering and Regenerative Medicine, AvePark-Parque de Ciência e Tecnologia, Zona Industrial da Gandra, 4805-017 Barco GMR (Portugal); ICVS/3B' s - PT Government Associate Laboratory, Braga/Guimarães (Portugal); Pires, Ricardo A., E-mail: rpires@dep.uminho.pt [3B' s Research Group - Biomaterials, Biodegradables and Biomimetics, University of Minho, Headquarters of the European Institute of Excellence on Tissue Engineering and Regenerative Medicine, AvePark-Parque de Ciência e Tecnologia, Zona Industrial da Gandra, 4805-017 Barco GMR (Portugal); ICVS/3B' s - PT Government Associate Laboratory, Braga/Guimarães (Portugal)

2017-02-01

We developed a porous poly-L-lactic acid (PLLA) scaffold compounded with borosilicate bioactive glasses (BBGs) endowing it with bioactive properties. Porous PLLA-BBG fibre mesh scaffolds were successfully prepared by the combination of wet spinning and fibre bonding techniques. Micro-computed tomography (μCT) confirmed that the PLLA-BBG scaffolds containing ≈ 25% of BBGs (w/w) exhibited randomly interconnected porous (58 to 62% of interconnectivity and 53 to 67% of porosity) with mean pore diameters higher that 100 μm. Bioactivity and degradation studies were performed by immersing the scaffolds in simulated body fluid (SBF) and ultrapure water, respectively. The PLLA-BBG scaffolds presented a faster degradation rate with a constant release of inorganic species, which are capable to produce calcium phosphate structures at the surface of the material after 7 days of immersion in SBF (Ca/P ratio of ~ 1.7). Cellular in vitro studies with human osteosarcoma cell line (Saos-2) and human adipose-derived stem cells (hASCs) showed that PLLA-BBGs are not cytotoxic to cells, while demonstrating their capacity to promote cell adhesion and proliferation. Overall, we showed that the proposed scaffolds present a tailored kinetics on the release of inorganic species and controlled biological response under conditions that mimic the bone physiological environment. - Highlights: • We prepared borosilicate glasses and their PLLA composites in the form of fibres. • These glasses imparted bioactivity and controlled degradability to the fibres. • The prepared fibres did not elicit cytotoxicity. • hASCs attached and proliferated in the surface and inner sections of the scaffolds. • The composites present appropriate properties to be used in bone tissue engineering.

Wetspun poly-L-(lactic acid)-borosilicate bioactive glass scaffolds for guided bone regeneration

International Nuclear Information System (INIS)

Fernandes, João S.; Reis, Rui L.; Pires, Ricardo A.

2017-01-01

We developed a porous poly-L-lactic acid (PLLA) scaffold compounded with borosilicate bioactive glasses (BBGs) endowing it with bioactive properties. Porous PLLA-BBG fibre mesh scaffolds were successfully prepared by the combination of wet spinning and fibre bonding techniques. Micro-computed tomography (μCT) confirmed that the PLLA-BBG scaffolds containing ≈ 25% of BBGs (w/w) exhibited randomly interconnected porous (58 to 62% of interconnectivity and 53 to 67% of porosity) with mean pore diameters higher that 100 μm. Bioactivity and degradation studies were performed by immersing the scaffolds in simulated body fluid (SBF) and ultrapure water, respectively. The PLLA-BBG scaffolds presented a faster degradation rate with a constant release of inorganic species, which are capable to produce calcium phosphate structures at the surface of the material after 7 days of immersion in SBF (Ca/P ratio of ~ 1.7). Cellular in vitro studies with human osteosarcoma cell line (Saos-2) and human adipose-derived stem cells (hASCs) showed that PLLA-BBGs are not cytotoxic to cells, while demonstrating their capacity to promote cell adhesion and proliferation. Overall, we showed that the proposed scaffolds present a tailored kinetics on the release of inorganic species and controlled biological response under conditions that mimic the bone physiological environment. - Highlights: • We prepared borosilicate glasses and their PLLA composites in the form of fibres. • These glasses imparted bioactivity and controlled degradability to the fibres. • The prepared fibres did not elicit cytotoxicity. • hASCs attached and proliferated in the surface and inner sections of the scaffolds. • The composites present appropriate properties to be used in bone tissue engineering.

The solidification of high-level liquid wastes in glass and ceramics

International Nuclear Information System (INIS)

Krause, H.

1989-01-01

In spent nuclear fuel reprocessing a highly radioactive waste solution is produced. It must be converted into a solid product, which binds the radionuclides, be hydrolytic as well as radiation and temperature resistant. Borosilicate glasses fulfil these requirements and, jointly with the barriers of a repository, they prevent inadmissible amounts of radionuclides from escaping into the biocycle. Two techniques were developed for industrial-scale vitrification: a rotary kiln calciner combined with an induction heated metallic melter and the electrode heated ceramic melters. Both techniques were already demonstrated on an industrial scale and under radioactive conditions. (AVM, Marcoule and PAMELA, Mol). (orig./MM) [de

Comparative assessment of TRU waste forms and processes. Volume I. Waste form and process evaluations

International Nuclear Information System (INIS)

Ross, W.A.; Lokken, R.O.; May, R.P.; Roberts, F.P.; Timmerman, C.L.; Treat, R.L.; Westsik, J.H. Jr.

1982-09-01

This study provides an assesses seven waste forms and eight processes for immobilizing transuranic (TRU) wastes. The waste forms considered are cast cement, cold-pressed cement, FUETAP (formed under elevated temperature and pressure) cement, borosilicate glass, aluminosilicate glass, basalt glass-ceramic, and cold-pressed and sintered silicate ceramic. The waste-immobilization processes considered are in-can glass melting, joule-heated glass melting, glass marble forming, cement casting, cement cold-pressing, FUETAP cement processing, ceramic cold-pressing and sintering, basalt glass-ceramic processing. Properties considered included gas generation, chemical durability, mechanical strength, thermal stability, and radiation stability. The ceramic products demonstrated the best properties, except for plutonium release during leaching. The glass and ceramic products had similar properties. The cement products generally had poorer properties than the other forms, except for plutonium release during leaching. Calculations of the Pu release indicated that the waste forms met the proposed NRC release rate limit of 1 part in 10 5 per year in most test conditions. The cast-cement process had the lowest processing cost, followed closely by the cold-pressed and FUETAP cement processes. Joule-heated glass melting had the lower cost of the glass processes. In-can melting in a high-quality canister had the highest cost, and cold-pressed and sintered ceramic the second highest. Labor and canister costs for in-can melting were identified. The major contributor to costs of disposing of TRU wastes in a defense waste repository is waste processing costs. Repository costs could become the dominant cost for disposing of TRU wastes in a commercial repository. It is recommended that cast and FUETAP cement and borosilicate glass waste-form systems be considered. 13 figures, 16 tables

Investigating the effect of V{sub 2}O{sub 5} addition on sodium barium borosilicate glasses

Energy Technology Data Exchange (ETDEWEB)

Halder, Rumu, E-mail: rumuhalder24feb@gmail.com; Sengupta, Pranesh; Dey, G. K. [Materials Science Division, Bhabha Atomic Research Centre, Mumbai-700 085 (India); Sudarsan, V. [Chemistry Division, Bhabha Atomic Research Centre, Mumbai-700 085 (India); Kaushik, C. P. [Waste Management Division, Bhabha Atomic Research Centre, Mumbai-700 085 (India)

2016-05-23

V{sub 2}O{sub 5} doped sodium barium borosilicate glasses were characterized by photoluminescence spectroscopy and electron probe microanalyzer (EPMA). The glass remains homogeneous for lower concentration of V{sub 2}O{sub 5} but a phase separation is observed when V{sub 2}O{sub 5} doping is increased beyond 5 mol%. Detailed microanalysis reveals that the phase separated glass consists of a phase containing V, Ba and Si and a separate Si rich phase within the glass matrix. The luminescence study demonstrated that at low concentration the vanadium mainly interacts with the structural units of B/Si while at higher concentrations, V-O-V/ V-O{sup −} Na{sup +}/Ba{sup 2+} linkages are formed.

Simulating the physicochemical properties of borosilicate and lanthanum borosilicate glasses using a polarizable force field

International Nuclear Information System (INIS)

Pacaud, Fabien

2016-01-01

as result of the nuclear waste vitrification, the knowledge and understanding of the dynamic and structural properties of glasses, including the behavior of radionuclides, is important (in liquid and solid phases). It can influence the glass waste properties, the lifetime of the vitrification process and the amount of radionuclides introduced in the glass matrix. Molecular dynamic simulations have been done to study the influence of the glass matrix composition into the structural and dynamic properties of the glass. a simplified glass, with 3 major oxides of the R7T7 glass such as SiO_2, B_2O_3 and Na_2O, have been used to simulate the R7T7 industrial nuclear glass (a 30 oxides glass). The inclusion of La_2O_3 allows us to simulate the impact of fission products and minor actinides into the properties of the glass matrix. Both systems, the SiO_2-B_2O_3-Na_2O and SiO_2-B_2O_3-Na_2O-La_2O_3, allow us to study the sodium and lanthanum effect on the properties of the glass. During this work, a polarizable force field has been developed to do these simulations. The results obtained at room temperature let us reproduce the experimental results of the structure, the distribution of BIII/BIV and the density. a study has been done on the viscosity and electrical conductivity of the liquid. The distribution BIV/BIII and the influence of the structural changes on the density along with the temperature have also been observed with thermal quenching. The current limits of this approach are also described. (author) [fr

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)

Glass viscosity calculation based on a global statistical modelling approach

Energy Technology Data Exchange (ETDEWEB)

Fluegel, Alex

2007-02-01

A global statistical glass viscosity model was developed for predicting the complete viscosity curve, based on more than 2200 composition-property data of silicate glasses from the scientific literature, including soda-lime-silica container and float glasses, TV panel glasses, borosilicate fiber wool and E type glasses, low expansion borosilicate glasses, glasses for nuclear waste vitrification, lead crystal glasses, binary alkali silicates, and various further compositions from over half a century. It is shown that within a measurement series from a specific laboratory the reported viscosity values are often over-estimated at higher temperatures due to alkali and boron oxide evaporation during the measurement and glass preparation, including data by Lakatos et al. (1972) and the recently published High temperature glass melt property database for process modeling by Seward et al. (2005). Similarly, in the glass transition range many experimental data of borosilicate glasses are reported too high due to phase separation effects. The developed global model corrects those errors. The model standard error was 9-17°C, with R^2 = 0.985-0.989. The prediction 95% confidence interval for glass in mass production largely depends on the glass composition of interest, the composition uncertainty, and the viscosity level. New insights in the mixed-alkali effect are provided.

High level waste at Hanford: Potential for waste loading maximization

International Nuclear Information System (INIS)

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

1995-09-01

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

Gamma ray shielding study of barium-bismuth-borosilicate glasses as transparent shielding materials using MCNP-4C code, XCOM program, and available experimental data

Energy Technology Data Exchange (ETDEWEB)

Bagheri, Reza; Yousefinia, Hassan [Nuclear Fuel Cycle Research School (NFCRS), Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of); Moghaddam, Alireza Khorrami [Radiology Department, Paramedical Faculty, Mazandaran University of Medical Sciences, Sari (Iran, Islamic Republic of)

2017-02-15

In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and 10th value layer values of barium-bismuth-borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium-bismuth-borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

Electrical characterization of strontium titanate borosilicate glass ceramics system with bismuth oxide addition using impedance spectroscopy

International Nuclear Information System (INIS)

Thakur, O.P.; Kumar, Devendra; Parkash, Om; Pandey, Lakshman

2003-01-01

The ac electrical data, measured in the frequency range 0.1 kHz-1 MHz, were used to study the electrical response of strontium titanate borosilicate glass ceramic system with bismuth oxide addition. Complex plane plots from these electrical data for various glass ceramic samples reveal contributions from simultaneously operating polarization mechanisms to overall dielectric behavior. The complex modulus (M * ) representation of electrical data for various glass ceramic samples were found to be more informative. Equivalent circuit models, which represent the electrical behavior of glass ceramic samples, were determined using complex non-linear least square (CNLS) fitting. An attempt has been made to understand the dielectric behavior of various glass ceramics in terms of contributions arising from different polarization processes occurring at glassy matrix, crystalline phases, glass to crystal interface region and blocking electrodes. Glass ceramics containing SrTiO 3 and TiO 2 (rutile) phases show thermally stable dielectric behavior

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

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

International Nuclear Information System (INIS)

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

2013-01-01

Chemical durability is not a single material property that can be uniquely measured. Instead it is the response to a host of coupled material and environmental processes whose rates are estimated by a combination of theory, experiment, and modeling. High-level nuclear waste (HLW) glass is perhaps the most studied of any material yet there remain significant technical gaps regarding their chemical durability. The phenomena affecting the long-term performance of HLW glasses in their disposal environment include surface reactions, transport properties to and from the reacting glass surface, and ion exchange between the solid glass and the surrounding solution and alteration products. The rates of these processes are strongly influenced and are coupled through the solution chemistry, which is in turn influenced by the reacting glass and also by reaction with the near-field materials and precipitation of alteration products. Therefore, those processes must be understood sufficiently well to estimate or bound the performance of HLW glass in its disposal environment over geologic time-scales. This article summarizes the current state of understanding of surface reactions, transport properties, and ion exchange along with the near-field materials and alteration products influences on solution chemistry and glass reaction rates. Also summarized are the remaining technical gaps along with recommended approaches to fill those technical gaps

Glass-bonded iodosodalite waste form for immobilization of 129I

Science.gov (United States)

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

2018-06-01

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

Silica based gel as a potential waste form for high level waste from fuel reprocessing

International Nuclear Information System (INIS)

Ford, C.E.; Dempster, T.J.; Melling, P.J.

1983-10-01

To assess the feasibility of safe disposal of high-level radioactive waste as synthetic clay, or material that would react with ground water to form clay, experiments have been carried out to determine the hydrothermal crystallisation and leaching behaviour of silica based gels fired at 900 deg C. Crystallisation rates at a pressure of 500 bars and at temperatures below 400 deg C are negligible and this more or less precludes pre-disposal production of synthetic clay on the scale required. Leaching experiments suggest that the leach rates of Cs from gels by distilled water are higher than those of boro-silicate glasses and SYNROC at the lower temperatures that would be preferred for geological storage. However, amounts of bulk dissolution of gels may be lower than those of boro-silicate glasses. The initial leaching behaviour of gels might be considerably improved by hot compaction at 900 to 1000 deg C. Consideration of likely waste form dissolution behaviour in a repository environment suggests that gels of appropriate composition might perform as well as, or better than, boro-silicate glasses. A novel hypothetical plant is described that could produce the gel waste form on the scale required on a more or less continuous basis. (author)

Testing and evaluation of solidified high-level waste forms

International Nuclear Information System (INIS)

Engelmann, C.

1984-01-01

The report describes research by several laboratories on the behaviour, in aqueous and salt environments, of borosilicate glass ceramics proposed for the solidification of nuclear wastes by the European Community. Results were obtained on inactive simulates, doped materials, and on borosilicate glass containing real radioactive waste. The influence of many important parameters were studied: leaching mode, nature of the leachant, pH, pressure, temperature, duration of the treatment, etc. The results of tests lasting for as little as a few hours or for as long as several hundred days, at temperatures up to 200 0 C or under pressures up to 200 bars, are presented. Numerous analytical techniques (ESCA, EMP, IRR, SEM, etc.) were used to determine the structure and the chemical composition of the altered layer developed by hydration at the glass surface. Information is also given on physical properties of the borosilicate glass: crystallization phase separation, alpha-irradiation stability, mechanical and thermal stability, etc. Finally, preliminary results on the structure and composition of hollandite ceramics are given

Energy transfer and cross-relaxation in Tb3+-doped borosilicate glasses

International Nuclear Information System (INIS)

Kim, Jung Hwan; Sol, Jung Sik

1990-01-01

Energy transfer in Tb 3+ -doped borosilicate glasses has been studied by the analysis of fluorescence intensities and lifetimes of 5 D 3 and 5 D 4 states as a function of Tb 3+ concentration. It is shown that as the Tb 3+ concentration is increased the cross-relaxation produces high population of the 5 D 4 state at the expense of 5 D 3 . It is also found that this interaction is predominantly dipole-dipole transition with critical distance of 13 A. The critical distance for energy transfer 5 D 4 5 D 3 which is responsible for the quenching of 5 D 4 emission at high concentration of Tb 3+ ions is 4.5 A. (Author)

Linear thermal expansion coefficient (at temperatures from 130 to 800 K) of borosilicate glasses applicable for coupling with silicon in microelectronics

OpenAIRE

Sinev, Leonid S.; Petrov, Ivan D.

2017-01-01

Processing results of measurements of linear thermal expansion coefficients and linear thermal expansion of two brands of borosilicate glasses --- LK5 and Borofloat 33 --- are presented. The linear thermal expansion of glass samples have been determined in the temperature range 130 to 800 K (minus 143 to 526 $\\deg$C) using thermomechanical analyzer TMA7100. Relative imprecision of indirectly measured linear thermal expansion coefficients and linear thermal expansion of both glass brands is le...

Vitrification of high-level liquid wastes

International Nuclear Information System (INIS)

Varani, J.L.; Petraitis, E.J.; Vazquez, Antonio.

1987-01-01

High-level radioactive liquid wastes produced in the fuel elements reprocessing require, for their disposal, a preliminary treatment by which, through a series of engineering barriers, the dispersion into the biosphere is delayed by 10 000 years. Four groups of compounds are distinguished among a great variety of final products and methods of elaboration. From these, the borosilicate glasses were chosen. Vitrification experiences were made at a laboratory scale with simulated radioactive wastes, employing different compositions of borosilicate glass. The installations are described. A series of tests were carried out on four basic formulae using always the same methodology, consisting of a dry mixture of the vitreous matrix's products and a dry simulated mixture. Several quality tests of the glasses were made 1: Behaviour in leaching following the DIN 12 111 standard; 2: Mechanical resistance; parameters related with the facility of the different glasses for increasing their surface were studied; 3: Degree of devitrification: it is shown that devitrification turns the glasses containing radioactive wastes easily leachable. From all the glasses tested, the composition SiO 2 , Al 2 O 3 , B 2 O 3 , Na 2 O, CaO shows the best retention characteristics. (M.E.L.) [es

Chemical durability of alkali-borosilicate glasses studied by analytical SEM, IBA, isotopic-tracing and SIMS

Science.gov (United States)

Trocellier, P.; Djanarthany, S.; Chêne, J.; Haddi, A.; Brass, A. M.; Poissonnet, S.; Farges, F.

2005-10-01

Simple and complex alkali-borosilicate glasses were submitted to aqueous corrosion at room temperature, 60 and 90 °C in solutions with pH ranging between 0 and 12. Analytical scanning electron microscopy (SEM), ion beam analysis (IBA) techniques, isotopic tracing and secondary ion mass-depth profiling (SIMS) have been used to investigate the variations of the surface composition of glass. In acidic medium, the glass surface is generally covered by a thick hydrated silica layer, mobile elements like Li, Na and B and transition elements (Fe, Zr, Mo, etc.) are strongly depleted. Near pH 7, relative enrichments of aluminium, iron and rare earths are shown together with strong Li, Na and B depletions. In basic medium, the glass surface exhibits relative enrichments of the major part of transition metals (from Cr to U) whereas mobile elements seem to be kept close to their nominal concentration level at the glass surface and Si is severely impoverished. Hydrogen incorporated at the glass surface after leaching is much more immobile in neutral and basic media than in acid medium.

Chemical durability of alkali-borosilicate glasses studied by analytical SEM, IBA, isotopic-tracing and SIMS

International Nuclear Information System (INIS)

Trocellier, P.; Djanarthany, S.; Chene, J.; Haddi, A.; Brass, A.M.; Poissonnet, S.; Farges, F.

2005-01-01

Simple and complex alkali-borosilicate glasses were submitted to aqueous corrosion at room temperature, 60 and 90 deg. C in solutions with pH ranging between 0 and 12. Analytical scanning electron microscopy (SEM), ion beam analysis (IBA) techniques, isotopic tracing and secondary ion mass-depth profiling (SIMS) have been used to investigate the variations of the surface composition of glass. In acidic medium, the glass surface is generally covered by a thick hydrated silica layer, mobile elements like Li, Na and B and transition elements (Fe, Zr, Mo, etc.) are strongly depleted. Near pH 7, relative enrichments of aluminium, iron and rare earths are shown together with strong Li, Na and B depletions. In basic medium, the glass surface exhibits relative enrichments of the major part of transition metals (from Cr to U) whereas mobile elements seem to be kept close to their nominal concentration level at the glass surface and Si is severely impoverished. Hydrogen incorporated at the glass surface after leaching is much more immobile in neutral and basic media than in acid medium

Conversion of highly active waste to solids

International Nuclear Information System (INIS)

Scheffler, K.

Borosilicate glasses were selected as matrix material for solidification of highly radioactive wastes. Current laboratory work on the VERA process is described. Goals were met by a five-component glass VG-38 and a glass-ceramic VC-15. The VERA process is described: flowsheet, denitration, calcinator, fusion facility

High-level waste processing and disposal

International Nuclear Information System (INIS)

Crandall, J.L.; Krause, H.; Sombret, C.; Uematsu, K.

1984-11-01

Without reprocessing, spent LWR fuel itself is generally considered an acceptable waste form. With reprocessing, borosilicate glass canisters, have now gained general acceptance for waste immobilization. The current first choice for disposal is emplacement in an engineered structure in a mined cavern at a depth of 500-1000 meters. A variety of rock types are being investigated including basalt, clay, granite, salt, shale, and volcanic tuff. This paper gives specific coverage to the national high level waste disposal plans for France, the Federal Republic of Germany, Japan and the United States. The French nuclear program assumes prompt reprocessing of its spent fuels, and France has already constructed the AVM. Two larger borosilicate glass plants are planned for a new French reprocessing plant at La Hague. France plans to hold the glass canisters in near-surface storage for a forty to sixty year cooling period and then to place them into a mined repository. The FRG and Japan also plan reprocessing for their LWR fuels. Both are currently having some fuel reprocessed by France, but both are also planning reprocessing plants which will include waste vitrification facilities. West Germany is now constructing the PAMELA Plant at Mol, Belgium to vitrify high level reprocessing wastes at the shutdown Eurochemic Plant. Japan is now operating a vitrification mockup test facility and plans a pilot plant facility at the Tokai reprocessing plant by 1990. Both countries have active geologic repository programs. The United State program assumes little LWR fuel reprocessing and is thus primarily aimed at direct disposal of spent fuel into mined repositories. However, the US have two borosilicate glass plants under construction to vitrify existing reprocessing wastes

Gamma Ray Shielding Study of Barium–Bismuth–Borosilicate Glasses as Transparent Shielding Materials using MCNP-4C Code, XCOM Program, and Available Experimental Data

Directory of Open Access Journals (Sweden)

Reza Bagheri

2017-02-01

Full Text Available In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and 10th value layer values of barium–bismuth–borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium–bismuth–borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

Performance testing of waste forms in a tuff environment

International Nuclear Information System (INIS)

Oversby, V.M.

1983-11-01

This paper describes experimental work conducted to establish the chemical composition of water which will have reacted with Topopah Spring Member tuff prior to contact with waste packages. The experimental program to determine the behavior of spent fuel and borosilicate glass in the presence of this water is then described. Preliminary results of experiments using spent fuel segments with defects in the Zircaloy cladding are presented. Some results from parametric testing of a borosilicate glass with tuff and 304L stainless steel are also discussed. Experiments conducted using Topopah Spring tuff and J-13 well water have been conducted to provide an estimate of the post-emplacement environment for waste packages in a repository at Yucca Mountain. The results show that emplacement of waste packages should cause only small changes in the water chemistry and rock mineralogy. The changes in environment should not have any detrimental effects on the performance of metal barriers or waste forms. The NNWSI waste form testing program has provided preliminary results related to the release rate of radionuclides from the waste package. Those results indicate that release rates from both spent fuel and borosilicate glass should be below 1 part in 10 5 per year. Future testing will be directed toward making release rate testing more closely relevant to site specific conditions. 17 references, 7 figures

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

Reinforcement of poly-l-lactic acid electrospun membranes with strontium borosilicate bioactive glasses for bone tissue engineering.

Science.gov (United States)

Fernandes, João S; Gentile, Piergiorgio; Martins, Margarida; Neves, Nuno M; Miller, Cheryl; Crawford, Aileen; Pires, Ricardo A; Hatton, Paul; Reis, Rui L

2016-10-15

Herein, for the first time, we combined poly-l-lactic acid (PLLA) with a strontium borosilicate bioactive glass (BBG-Sr) using electrospinning to fabricate a composite bioactive PLLA membrane loaded with 10% (w/w) of BBG-Sr glass particles (PLLA-BBG-Sr). The composites were characterised by scanning electron microscopy (SEM) and microcomputer tomography (μ-CT), and the results showed that we successfully fabricated smooth and uniform fibres (1-3μm in width) with a homogeneous distribution of BBG-Sr microparticles (bone marrow-derived mesenchymal stem cells (BM-MSCs) demonstrated that PLLA-BBG-Sr membranes promoted the osteogenic differentiation of the cells as demonstrated by increased alkaline phosphatase activity and up-regulated osteogenic gene expression (Alpl, Sp7 and Bglap) in relation to PLLA alone. These results strongly suggest that the composite PLLA membranes reinforced with the BBG-Sr glass particles have potential as an effective biomaterial capable of promoting bone regeneration. PLLA membranes were reinforced with 10% (w/w) of strontium-bioactive borosilicate glass microparticles, and their capacity to induce the osteogenic differentiation of bone marrow mesenchymal stem cells (BM-MSCs) was evaluated. These membranes presented an increased: degradability, water uptake, Young modulus and tensile strength. We also demonstrated that these membranes are non-cytotoxic and promote the attachment of BM-MSCs. The addition of the glass microparticles into the PLLA membranes promoted the increase of ALP activity (under osteogenic conditions), as well as the BM-MSCs osteogenic differentiation as shown by the upregulation of Alpl, Sp7 and Bglap gene expression. Overall, we demonstrated that the reinforcement of PLLA with glass microparticles results in a biomaterial with the appropriate properties for the regeneration of bone tissue. Copyright © 2016 Acta Materialia Inc. All rights reserved.

Immobilization of high-level wastes into sintered glass: 2

International Nuclear Information System (INIS)

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

1987-01-01

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

Processing and optical characterization of lead calcium titanate borosilicate glass doped with germanium

Science.gov (United States)

Gautam, C. R.; Das, Sangeeta; Gautam, S. S.; Madheshiya, Abhishek; Singh, Anod Kumar

2018-04-01

In this study, various compositions of lead calcium titanate borosilicate glass doped with a fixed amount of germanium were synthesized using the rapid melt quench technique. The amorphous nature of the synthesized glass was confirmed by X-ray diffraction and scanning electron microscopy analyses. The structural and optical properties were deduced using Raman, Fourier transform infrared (FTIR), and ultraviolet-visible (UV-Vis) spectroscopy. FTIR spectroscopy confirmed the presence of borate groups in triangular and tetrahedral coordination. Infrared and Raman analyses detected the vibrational bonds of Gesbnd Osbnd Ge, Bsbnd Osbnd Ge, Sisbnd Osbnd Ge, Sisbnd Osbnd Si, and Pbsbnd Osbnd Ge. The energy band gaps were evaluated for the prepared glass samples based on Tauc plots of the UV-Vis spectra. The calculated values of the optical band gap decreased from 2.91 to 2.85 eV as the PbO content increased from x = 0.0 to x = 0.7. Furthermore, the Urbach energy was studied based on the UV-Vis results to confirm the disordered structure of the glass. The calculated densities of the glass samples (1.5835 g/cm3 to 3.9184 g/cm3) increased as the concentration of PbO increased, whereas they decreased with the molar volume.

Phase evolution and dielectric properties of MgTi2O5 ceramic sintered with lithium borosilicate glass

International Nuclear Information System (INIS)

Shin, Hyunho; Shin, Hee-Kyun; Jung, Hyun Suk; Cho, Seo-Yong; Hong, Kug Sun

2005-01-01

Phase evolution, densification, and dielectric properties of MgTi 2 O 5 dielectric ceramic, sintered with lithium borosilicate (LBS) glass, were studied. Reaction between LBS glass and MgTi 2 O 5 was significant in forming secondary phases such as TiO 2 and (Mg,Ti) 2 (BO 3 )O. The glass addition was not necessarily deleterious to the dielectric properties due to the formation of TiO 2 : permittivity increased and temperature coefficient of resonance frequency could be tuned to zero with the addition of LBS glass, although the inevitable glass-induced decrease of quality factor was not retarded by the formation of TiO 2 . The sintered specimen with 10 wt% LBS fired at 950 deg. C for 2 h showed permittivity of 19.3, quality factor of 6800 GHz, and τ f of -16 ppm/ deg. C

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

to the low solubility of RuO 2 in borosilicate glass. These particles tended to form agglomerates with varying sizes and shapes that were located close to the bottom of crucibles. The results of this study provide further insight into the ability of borosilicate waste glass to incorporate increased (>16 wt %) concentrations of aluminum. The glass composition and properties data will be incorporated into a database of glass composition-property relationships (ComPro) to support further optimization of waste glass compositions at DOE sites

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

O{sub 2} in all glasses due to the low solubility of RuO{sub 2} in borosilicate glass. These particles tended to form agglomerates with varying sizes and shapes that were located close to the bottom of crucibles. The results of this study provide further insight into the ability of borosilicate waste glass to incorporate increased (>16 wt %) concentrations of aluminum. The glass composition and properties data will be incorporated into a database of glass composition-property relationships (ComPro) to support further optimization of waste glass compositions at DOE sites.

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Influence of iron ions on the structural properties of some inorganic glasses

International Nuclear Information System (INIS)

Music, S.; Gotic, M.; Popovic, S.; Grzeta, B.

1987-01-01

The effects of iron on the structural properties of Zn-borosilicate glass and Pb-metaphosphate glass were studied using x-ray diffraction, 57 Fe Moessbauer spectroscopy and IR spectroscopy. At high concentration of iron the crystallization of zinc ferrite in the glass matrix takes place. X-ray diffraction and 57 Fe Moessbauer spectroscopy showed that the amount of zinc ferrite in Zn-borosilicate glass decreases. In Pb-metaphosphate glass doped with high concentration of α-Fe 2 O 3 , the crystallization of Fe 3 (PO 4 ) 2 is pronounced. The assignments of IR band positions and the corresponding interpretation are given. The importance of this study for the technology of vitrification of high-level radioactive wastes is emphasized. (author) 31 refs.; 6 figs,.; 6 tabs

A statistically designed matrix to evaluate solubility, impurity tolerance, and thermal stability of plutonium-bearing glasses

International Nuclear Information System (INIS)

Peeler, D.K.; Meaker, T.F.; Edwards, T.B.; McIntyre, D.S.

1997-01-01

In support of the Department of Energy's (DOE) Office of Fissile Material Disposition (OFDM) Program, Westinghouse Savannah River Company (WSRC) is evaluating a unique lanthanide borosilicate glass to immobilize excess plutonium and other heavy metals. The lanthanide borosilicate (LaBS) glass system met all FY96 programmatic planning objectives. Those objectives were focused on (1) demonstrating 10 wt% Pu solubility, and (2) meeting preliminary product performance criteria. Although 10 wt% Pu solubility was demonstrated with product performance exceeding high level waste glasses based on PCT results, the LaBS system was not optimized

Effects of carbonate and sulphate ions in synthetic groundwater on high-level waste glass leaching

International Nuclear Information System (INIS)

Kamizono, H.

1990-01-01

This laboratory experiment aims to examine the effects of rare earth carbonate and sulphate ions, that are naturally present in underground water, have on glass used to store high-level radioactive waste for disposal underground. Borosilicate glass (or HLW glass) is stored under observation on the land surface for several decades before being buried deep below ground in geological disposal sites. Two types of precipitation occur during leaching from the glass, immediate formation of a hydrated surface layer and slow precipitation from concentration in the leachates. This slow process of some elements precipitating onto the glass surface or into the leachates is examined in this experiment using scanning electron microscopy (SEM) and energy-dispersive X-ray analysis (EDX). Carbonates from rare-earth elements are found in the synthetic groundwater used. It is shown that carbonate and sulphate ions will affect leaching and will occur in geological disposal sites. Other particles were also observed to precipitate using SEM-EDX. (author)

Effects of solution chemistry and atmosphere on leaching of alkali borosilicate glass

International Nuclear Information System (INIS)

Hermansson, H.P.; Christensen, H.; Clark, D.E.; Werme, L.

1983-01-01

The leaching behavior of two alkali-borosilicate glasses containing 9 wt % simulated fission products and 1.6 wt % uranium oxide has been studied. Samples were exposed to one of eight types of leachants including doubly distilled water, simulated ground silicate water, a brine solution, and solutions containing various concentrations of iron, aluminum or sodium maintained at either 25 0 C, 40 0 C or 90 0 C for up to 182 days. The most aggressive leachants were the solutions containing sodium (excluding brine) and simulated ground silicate water. These solutions increased the extent of leaching by a factor of 2 to 3 over that for distilled water for one of the glasses. A partially protective surface film rich in magnesium, potassium, and chlorine was formed on the glasses exposed to the brine solution. In order to evaluate the effects of atmosphere on leaching, samples were also immersed in doubly distilled water over which the relative concentrations of oxygen, nitrogen and carbon dioxide were varied. Increasing the carbon dioxide concentration from 0 to 50% resulted in a factor of 3 increase in the leaching rate

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

Small-scale integrated demonstration of high-level radioactive waste processing and vitrification using actual SRP waste

International Nuclear Information System (INIS)

Ferguson, R.B.; Woolsey, G.B.; Galloway, R.M.; Baumgarten, P.M.; Eibling, R.E.

1980-01-01

Experiments have been made to demonstrate the feasibility of immobilizing SRP high-level waste in borosilicate glass. Results to date are encouraging. Equipment performance and processing characteristics for solidifying small batches of actual SRP waste have agreed well with previous experience with small- and large-scale tests synthetic waste, and with theoretical predictions