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Sample records for hlw glass degradation

  1. HLW immobilization in glass

    Leroy, P.; Jacquet-Francillon, N.; Runge, S.

    1992-01-01

    The immobilization of High Level Waste in glass in France is a long history which started as early as in the 1950's. More than 30 years of Research and Development have been invested in that field. Two industrial facilities are operating (AVM and R7) and a third one (T7), under cold testing, is planned to start active operation in the mid-92. While vitrification has been demonstrated to be an industrially mastered process, the question of the quality of the final waste product, i.e. the HLW glass, must be addressed. The scope of the present paper is to focus on the latter point from both standpoints of the R and D and of the industrial reality

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

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

    2012-04-02

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

  3. SOURCE TERMS FOR HLW GLASS CANISTERS

    J.S. Tang

    2000-01-01

    This calculation is prepared by the Monitored Geologic Repository (MGR) Waste Package Design Section. The objective of this calculation is to determine the source terms that include radionuclide inventory, decay heat, and radiation sources due to gamma rays and neutrons for the high-level radioactive waste (HLW) from the, West Valley Demonstration Project (WVDP), Savannah River Site (SRS), Hanford Site (HS), and Idaho National Engineering and Environmental Laboratory (INEEL). This calculation also determines the source terms of the canister containing the SRS HLW glass and immobilized plutonium. The scope of this calculation is limited to source terms for a time period out to one million years. The results of this calculation may be used to carry out performance assessment of the potential repository and to evaluate radiation environments surrounding the waste packages (WPs). This calculation was performed in accordance with the Development Plan ''Source Terms for HLW Glass Canisters'' (Ref. 7.24)

  4. HIGH ALUMINUM HLW GLASSES FOR HANFORD'S WTP

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

    2009-01-01

    The world's largest radioactive waste vitrification facility is now under construction at the United State Department of Energy's (DOE's) Hanford site. The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is designed to treat nearly 53 million gallons of mixed hazardous and radioactive waste now residing in 177 underground storage tanks. This multi-decade processing campaign will be one of the most complex ever undertaken because of the wide chemical and physical variability of the waste compositions generated during the cold war era that are stored at Hanford. The DOE Office of River Protection (ORP) has initiated a program to improve the long-term operating efficiency of the WTP vitrification plants with the objective of reducing the overall cost of tank waste treatment and disposal and shortening the duration of plant operations. Due to the size, complexity and duration of the WTP mission, the lifecycle operating and waste disposal costs are substantial. As a result, gains in High Level Waste (HLW) and Low Activity Waste (LAW) waste loadings, as well as increases in glass production rate, which can reduce mission duration and glass volumes for disposal, can yield substantial overall cost savings. EnergySolutions and its long-term research partner, the Vitreous State Laboratory (VSL) of the Catholic University of America, have been involved in a multi-year ORP program directed at optimizing various aspects of the HLW and LAW vitrification flow sheets. A number of Hanford HLW streams contain high concentrations of aluminum, which is challenging with respect to both waste loading and processing rate. Therefore, a key focus area of the ORP vitrification process optimization program at EnergySolutions and VSL has been development of HLW glass compositions that can accommodate high Al 2 O 3 concentrations while maintaining high processing rates in the Joule Heated Ceramic Melters (JHCMs) used for waste vitrification at the WTP. This paper, reviews the

  5. Database and Interim Glass Property Models for Hanford HLW Glasses

    Hrma, Pavel R; Piepel, Gregory F; Vienna, John D; Cooley, Scott K; Kim, Dong-Sang; Russell, Renee L

    2001-01-01

    The purpose of this report is to provide a methodology for an increase in the efficiency and a decrease in the cost of vitrifying high-level waste (HLW) by optimizing HLW glass formulation. This methodology consists in collecting and generating a database of glass properties that determine HLW glass processability and acceptability and relating these properties to glass composition. The report explains how the property-composition models are developed, fitted to data, used for glass formulation optimization, and continuously updated in response to changes in HLW composition estimates and changes in glass processing technology. Further, the report reviews the glass property-composition literature data and presents their preliminary critical evaluation and screening. Finally the report provides interim property-composition models for melt viscosity, for liquidus temperature (with spinel and zircon primary crystalline phases), and for the product consistency test normalized releases of B, Na, and Li. Models were fitted to a subset of the screened database deemed most relevant for the current HLW composition region

  6. Influence of Glass Property Restrictions on Hanford HLW Glass Volume

    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.

  7. HLW immobilization in glass: industrial operation and product quality

    Jacquet-Francillon, N.; Leroy, P.; Runge, S.

    1992-01-01

    This extended summary discusses the immobilization of high level wastes from the viewpoint of the quality of the final product, i.e. the HLW glass. The R and D studies comprise 3 steps: glass formulation, glass characterization and long term behaviour studies

  8. Development Of Glass Matrices For HLW Radioactive Wastes

    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.

  9. DEVELOPMENT OF GLASS MATRICES FOR HLW RADIOACTIVE WASTES

    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.

  10. Cooling and cracking of technical HLW glass products

    Kienzler, B.

    1989-01-01

    The author discusses various cooling procedures applied to canisters filled with inactive simulated HLW glass and the measured temperature distributions compared with numerically computed data. Stress computations of the cooling process were carried out with a finite element method. Only those volume elements having temperatures below the transformation temperature Tg were assumed to contribute thermoelastically to the developing stresses. Model calculations were extended to include real HLW glass canisters with inherent thermal power. The development of stress as a function of variations of heat flow conditions and of the radioactive decay was studied

  11. Expected behavior of HLW glass in storage

    McElroy, J.L.

    1975-01-01

    Glass produced by solidification of high-level radioactive liquid waste is studied. Conditions to which the waste form will be exposed in a typical handling sequence representative of current U. S. planning are tabulated. The reference matrix for waste form characterization is discussed, and some of the properties of high-level waste glass are described: physical properties, leachability, fracturing, vaporization, and containment in canister. 12 fig, 5 tables

  12. Chemical compatibility of HLW borosilicate glasses with actinides

    Walker, C.T.; Scheffler, K.; Riege, U.

    1978-11-01

    During liquid storage of HLLW the formation of actinide enriched sludges is being expected. Also during melting of HLW glasses an increase of top-to-bottom actinide concentrations can take place. Both effects have been studied. Besides, the vitrification of plutonium enriched wastes from Pu fuel element fabrication plants has been investigated with respect to an isolated vitrification process or a combined one with the HLLW. It is shown that the solidification of actinides from HLLW and actinide waste concentrates will set no principal problems. The leaching of actinides has been measured in salt brine at 23 0 C and 115 0 C. (orig.) [de

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

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

    1999-01-01

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

  14. Using process instrumentation to obviate destructive examination of canisters of HLW glass

    Kuhn, W.L.; Slate, S.C.

    1983-01-01

    An important concern of a manufacturer of packages of solidified high-level waste (HLW) is quality assurance of the waste form. The vitrification of HLW as a borosilicate glass is considered, and, based on a reference vitrification process, it is proposed that information from process instrumentation may be used to assure quality without the need for additional information obtained by destructive examining (core drilling) canisters of glass. This follows mainly because models of product performance and process behavior must be previously established in order to confidently select the desired glass formulation, and to have confidence that the process is well enough developed to be installed and operated in a nuclear facility

  15. DEFENSE HIGH LEVEL WASTE GLASS DEGRADATION

    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

  16. The production of advanced glass ceramic HLW forms using cold crucible induction melter

    Rutledge, V.J.; Maio, V.

    2013-01-01

    Cold Crucible Induction Melters (CCIM) will favorably change how High-Level radioactive Waste (from nuclear fuel recovery) is treated in a near future. Unlike the existing Joule-Heated Melters (JHM) currently in operation for the glass-based immobilization of High-Level Waste (HLW), CCIM offers unique material features that will increase melt temperatures, increase throughput, increase mixing, increase loading in the waste form, lower melter foot prints, eliminate melter corrosion and lower costs. These features not only enhance the technology for producing HLW forms, but also provide advantageous attributes to the waste form by allowing more durable alternatives to glass. It is concluded that glass ceramic waste forms that are tailored to immobilize fission products of HLW can be can be made from the HLW processed with the CCIM. The advantageous higher temperatures reached with the CCIM and unachievable with JHM allows the lanthanides, alkali, alkaline earths, and molybdenum to dissolve into a molten glass. Upon controlled cooling they go into targeted crystalline phases to form a glass ceramic waste form with higher waste loadings than achievable with borosilicate glass waste forms. Natural cooling proves to be too fast for the formation of all targeted crystalline phases

  17. LIQUIDUS TEMPERATURE AND ONE PERCENT CRYSTAL CONTENT MODELS FOR INITIAL HANFORD HLW GLASSES

    Vienna, John D.; Edwards, Tommy B.; Crum, Jarrod V.; Kim, Dong-Sang; Peeler, David K.

    2005-01-01

    Preliminary models for liquidus temperature (TL) and temperature at 1 vol% crystal (T01) applicable to WTP HLW glasses in the spinel primary phase field were developed. A series of literature model forms were evaluated using consistent sets of data form model fitting and validation. For TL, the ion potential and linear mixture models performed best, while for T01 the linear mixture model out performed all other model forms. TL models were able to predict with smaller uncertainty. However, the lower T01 values (even with higher prediction uncertainties) were found to allow for a much broader processing envelope for WTP HLW glasses

  18. Long-term product consistency test of simulated 90-19/Nd HLW glass

    Gan, X.Y.; Zhang, Z.T.; Yuan, W.Y.; Wang, L.; Bai, Y.; Ma, H.

    2011-01-01

    Chemical durability of 90-19/Nd glass, a simulated high-level waste (HLW) glass in contact with the groundwater was investigated with a long-term product consistency test (PCT). Generally, it is difficult to observe the long term property of HLW glass due to the slow corrosion rate in a mild condition. In order to overcome this problem, increased contacting surface (S/V = 6000 m -1 ) and elevated temperature (150 o C) were employed to accelerate the glass corrosion evolution. The micro-morphological characteristics of the glass surface and the secondary minerals formed after the glass alteration were analyzed by SEM-EDS and XRD, and concentrations of elements in the leaching solution were determined by ICP-AES. In our experiments, two types of minerals, which have great impact on glass dissolution, were found to form on 90-19/Nd HLW glass surface when it was subjected to a long-term leaching in the groundwater. One is Mg-Fe-rich phyllosilicates with honeycomb structure; the other is aluminosilicates (zeolites). Mg and Fe in the leaching solution participated in the formation of phyllosilicates. The main components of phyllosilicates in alteration products of 90-19/Nd HLW glass are nontronite (Na 0.3 Fe 2 Si 4 O 10 (OH) 2 .4H 2 O) and montmorillonite (Ca 0.2 (Al,Mg) 2 Si 4 O 10 (OH) 2 .4H 2 O), and those of aluminosilicates are mordenite ((Na 2 ,K 2 ,Ca)Al 2 Si 10 O 24 .7H 2 O)) and clinoptilolite ((Na,K,Ca) 5 Al 6 Si 30 O 72 .18H 2 O). Minerals like Ca(Mg)SO 4 and CaCO 3 with low solubility limits are prone to form precipitant on the glass surface. Appearance of the phyllosilicates and aluminosilicates result in the dissolution rate of 90-19/Nd HLW glass resumed, which is increased by several times over the stable rate. As further dissolution of the glass, both B and Na in the glass were found to leach out in borax form.

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

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

    2010-01-01

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

  20. Studies on the long-term characteristics of HLW glass under ultimate storage conditions

    Roggendorf, H.; Conradt, R.; Ostertag, R.

    1987-01-01

    This interim report deals with first results of corrosion investigations of HLW simulation glass (COGEMA glass SON 68) in quinary salt solutions of different concentrations; the aim of these investigations was to find out about the corrosion mechanism at the surface of the glass and the quantitative registration of the corrosion products. It became obvious that the surface layers developed can be easily removed and that a determination of weight losses becomes possible thereby. The corrosion rates for a test period of 30 days were determined. (RB) [de

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

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

    2017-01-01

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

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

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

    2013-07-01

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

  3. Corrosion behaviour of the WAK-HLW glass

    Grambow, B.; Luckscheiter, B.; Nesovic, M.

    1997-01-01

    Sorption studies were performed on corrosion products from the glass GP WAK1 formed over a period of 40 days in deionized water at 80 C and S/V=1000 m -1 . After 40 days the pH of the solution was adjusted to various preselected values in the pH range 2-10. The pH was kept constant during the experiments by daily addition of either HNO 3 or NaOH. The sorption experiments were run at ambient temperature and 80 C for up to 10 days using various starting concentrations of Eu, Th and U. Sorption isotherms of Eu, Th and U(VI) on corrosion products were determined in deionized water, in NaCl-rich and MgCl 2 -rich solution. Presently, data of the sorption studies in deionized water are available.Furthermore the investigations of the pH dependence of saturation concentration of silica and of the release of various glass constituent of the glass GP WAK1 were continued with studies in the MgCl 2 -rich solution 1 at 80 C. Results of these studies (30 days) are given in terms of normalized elemental mass losses. (MM)

  4. HLW vitrification in France industrial experience and glass quality

    Desvaux, J.L.; Delahaye, P.

    1994-01-01

    This paper describes the vitrification process, the technology and process improvements at the La Hague plant in R 7 and T 7 facilities. The main achievements relate to the process flexibility, the reliability of the equipment and solid waste management. The quality of the vitrified glass produced and canisters compliance with agreed specifications are demonstrated through characterization studies. Since the active start-up of R 7/T 7 facilities, canisters compliance with specifications relies upon a complete quality assurance/quality control program including process control. 1 tab., 1 fig

  5. Determination of alpha dose rate profile at the HLW nuclear glass/water interface

    Mougnaud, S., E-mail: sarah.mougnaud@cea.fr [CEA Marcoule, DEN/DTCD/SECM, BP 17171, 30207 Bagnols-sur-Cèze cedex (France); Tribet, M.; Rolland, S. [CEA Marcoule, DEN/DTCD/SECM, BP 17171, 30207 Bagnols-sur-Cèze cedex (France); Renault, J.-P. [CEA Saclay, NIMBE UMR 3685 CEA/CNRS, 91191 Gif-sur-Yvette cedex (France); Jégou, C. [CEA Marcoule, DEN/DTCD/SECM, BP 17171, 30207 Bagnols-sur-Cèze cedex (France)

    2015-07-15

    Highlights: • The nuclear glass/water interface is studied. • The way the energy of alpha particles is deposited is modeled using MCNPX code. • A model giving dose rate profiles at the interface using intrinsic data is proposed. • Bulk dose rate is a majoring estimation in alteration layer and in surrounding water. • Dose rate is high in small cracks; in larger ones irradiated volume is negligible. - Abstract: Alpha irradiation and radiolysis can affect the alteration behavior of High Level Waste (HLW) nuclear glasses. In this study, the way the energy of alpha particles, emitted by a typical HLW glass, is deposited in water at the glass/water interface is investigated, with the aim of better characterizing the dose deposition at the glass/water interface during water-induced leaching mechanisms. A simplified chemical composition was considered for the nuclear glass under study, wherein the dose rate is about 140 Gy/h. The MCNPX calculation code was used to calculate alpha dose rate and alpha particle flux profiles at the glass/water interface in different systems: a single glass grain in water, a glass powder in water and a water-filled ideal crack in a glass package. Dose rate decreases within glass and in water as distance to the center of the grain increases. A general model has been proposed to fit a dose rate profile in water and in glass from values for dose rate in glass bulk, alpha range in water and linear energy transfer considerations. The glass powder simulation showed that there was systematic overlapping of radiation fields for neighboring glass grains, but the water dose rate always remained lower than the bulk value. Finally, for typical ideal cracks in a glass matrix, an overlapping of irradiation fields was observed while the crack aperture was lower than twice the alpha range in water. This led to significant values for the alpha dose rate within the crack volume, as long as the aperture remained lower than 60 μm.

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

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

    2010-01-04

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

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

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

    2010-01-01

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

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

    Veronica J Rutledge; Vince Maio

    2013-10-01

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

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

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

    2017-01-15

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

  10. Viability for controlling long-term leaching of radionuclides from HLW glass by amorphous silica additives

    Inagaki, Y.; Uehara, S.

    2004-01-01

    Dissolution and deterioration experiments in coexistence system of amorphous silica and vitrified wastes have been executed in order to evaluating the effects of amorphous silica addition to high level radioactive vitrified waste (HLW glass) on suppression of nuclide leaching. Geo-chemical reaction mechanism among the vitrified waste, the amorphous silica and water was also evaluated. Dissolution of the silica network was suppressed by addition of the amorphous silica. However, the leaching of soluble nuclides like B proceeded depending on the hydration deterioration reaction. (A. Hishinuma)

  11. The development of basic glass formulations for solidifying HLW from nuclear fuel reprocessing plant

    Jiang Yaozhong; Tang Baolong; Zhang Baoshan; Zhou Hui

    1995-01-01

    Basic glass formulations 90U/19, 90U/20, 90Nd/7 and 90Nd/10 applied in electric melting process are developed by using the mathematical model of the viscosity and electric resistance of waste glass. The yellow phase does not occur for basic glass formulations 90U/19 and 90U/20 solidifying HLW from nuclear fuel reprocessing plant when the waste loading is 20%. Under the waste loading is 16%, the process and product properties of glass 90U/19 and 90U/20 come up to or surpass the properties of the same kind of foreign waste glasses, and other properties are about the same to them of foreign waste glasses. The process and product properties of basic glass formulations 90Nd/7 and 90Nd/10 used for the solidification of 'U replaced by Nd' liquid waste are almost similar to them of 90U/19 and 90U/20. These properties fairly meet the requirements of 'joint test' (performed at KfK-INE, Germany). Among these formulations, 90Nd/7 is applied in cold engineering scale electric melting test performed at KfK-INE in Germany. The main process properties of cold test is similar to laboratory results

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

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

    2009-12-30

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

  13. DETERMINATION OF HLW GLASS MELT RATE USING X-RAY COMPUTED TOMOGRAPHY

    Choi, A.; Miller, D.; Immel, D.

    2011-10-06

    The purpose of the high-level waste (HLW) glass melt rate study is two-fold: (1) to gain a better understanding of the impact of feed chemistry on melt rate through bench-scale testing, and (2) to develop a predictive tool for melt rate in support of the on-going frit development efforts for the Defense Waste Processing Facility (DWPF). In particular, the focus is on predicting relative melt rates, not the absolute melt rates, of various HLW glass formulations solely based on feed chemistry, i.e., the chemistry of both waste and glass-forming frit for DWPF. Critical to the successful melt rate modeling is the accurate determination of the melting rates of various HLW glass formulations. The baseline procedure being used at the Savannah River National Laboratory (SRNL) is to; (1) heat a 4 inch-diameter stainless steel beaker containing a mixture of dried sludge and frit in a furnace for a preset period of time, (2) section the cooled beaker along its diameter, and (3) measure the average glass height across the sectioned face using a ruler. As illustrated in Figure 1-1, the glass height is measured for each of the 16 horizontal segments up to the red lines where relatively large-sized bubbles begin to appear. The linear melt rate (LMR) is determined as the average of all 16 glass height readings divided by the time during which the sample was kept in the furnace. This 'visual' method has proved useful in identifying melting accelerants such as alkalis and sulfate and further ranking the relative melt rates of candidate frits for a given sludge batch. However, one of the inherent technical difficulties of this method is to determine the glass height in the presence of numerous gas bubbles of varying sizes, which is prevalent especially for the higher-waste-loading glasses. That is, how the red lines are drawn in Figure 1-1 can be subjective and, therefore, may influence the resulting melt rates significantly. For example, if the red lines are drawn too low

  14. Crystallization in high level waste (HLW) glass melters: Savannah River Site operational experience

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

    2015-06-12

    This paper provides a review of the scaled melter testing that was completed for design input to the Defense Waste Processing Facility (DWPF) melter. Testing with prototype melters provided the data to define the DWPF operating limits to avoid bulk (volume) crystallization in the un-agitated DWPF melter and provided the data to distinguish between spinels generated by refractory corrosion versus spinels that precipitated from the HLW glass melt pool. A review of the crystallization observed with the prototype melters and the full-scale DWPF melters (DWPF Melter 1 and DWPF Melter 2) is included. Examples of actual DWPF melter attainment with Melter 2 are given. The intent is to provide an overview of lessons learned, including some example data, that can be used to advance the development and implementation of an empirical model and operating limit for crystal accumulation for a waste treatment and immobilization plant.

  15. Impact Of Particle Agglomeration On Accumulation Rates In The Glass Discharge Riser Of HLW Melter

    Kruger, A. A.; Rodriguez, C. A.; Matyas, J.; Owen, A. T.; Jansik, D. P.; Lang, J. B.

    2012-01-01

    The major factor limiting waste loading in continuous high-level radioactive waste (HLW) melters is an accumulation of particles in the glass discharge riser during a frequent and periodic idling of more than 20 days. An excessive accumulation can produce robust layers a few centimeters thick, which may clog the riser, preventing molten glass from being poured into canisters. Since the accumulation rate is driven by the size of particles we investigated with x-ray microtomography, scanning electron microscopy, and image analysis the impact of spinel forming components, noble metals, and alumina on the size, concentration, and spatial distribution of particles, and on the accumulation rate. Increased concentrations of Fe and Ni in the baseline glass resulted in the formation of large agglomerates that grew over the time to an average size of ∼185+-155 μm, and produced >3 mm thick layer after 120 h at 850 deg C. The noble metals decreased the particle size, and therefore significantly slowed down the accumulation rate. Addition of alumina resulted in the formation of a network of spinel dendrites which prevented accumulation of particles into compact layers

  16. Final Report - Crystal Settling, Redox, and High Temperature Properties of ORP HLW and LAW Glasses, VSL-09R1510-1, Rev. 0, dated 6/18/09

    Kruger, Albert A.; Wang, C.; Gan, H.; Pegg, I. L.; Chaudhuri, M.; Kot, W.; Feng, Z.; Viragh, C.; McKeown, D. A.; Joseph, I.; Muller, I. S.; Cecil, R.; Zhao, W.

    2013-11-13

    The radioactive tank waste treatment programs at the U. S. Department of Energy (DOE) have featured joule heated ceramic melter technology for the vitrification of high level waste (HLW). The Hanford Tank Waste Treatment and Immobilization Plant (WTP) employs this same basic technology not only for the vitrification of HLW streams but also for the vitrification of Low Activity Waste (LAW) streams. Because of the much greater throughput rates required of the WTP as compared to the vitrification facilities at the West Valley Demonstration Project (WVDP) or the Defense Waste Processing Facility (DWPF), the WTP employs advanced joule heated melters with forced mixing of the glass pool (bubblers) to improve heat and mass transport and increase melting rates. However, for both HLW and LAW treatment, the ability to increase waste loadings offers the potential to significantly reduce the amount of glass that must be produced and disposed and, therefore, the overall project costs. This report presents the results from a study to investigate several glass property issues related to WTP HLW and LAW vitrification: crystal formation and settling in selected HLW glasses; redox behavior of vanadium and chromium in selected LAW glasses; and key high temperature thermal properties of representative HLW and LAW glasses. The work was conducted according to Test Plans that were prepared for the HLW and LAW scope, respectively. One part of this work thus addresses some of the possible detrimental effects due to considerably higher crystal content in waste glass melts and, in particular, the impact of high crystal contents on the flow property of the glass melt and the settling rate of representative crystalline phases in an environment similar to that of an idling glass melter. Characterization of vanadium redox shifts in representative WTP LAW glasses is the second focal point of this work. The third part of this work focused on key high temperature thermal properties of

  17. Long term corrosion behavior of the WAK-HLW glass in salt solutions

    Luckscheiter, B.; Nesovic, M.

    1998-01-01

    The corrosion behavior of the HLW glass GP WAK1 containing simulated HLW oxides from the WAK reprocessing plant in Karlsruhe is investigated in long-term corrosion experiments at high S/V ratios in two reference brines at 110 and 190 C. In case of the MgCl 2 -rich solution the leachate becomes increasingly acid with reaction time up to a final pH of about 3.5 at 190 C. In the NaCl-rich solution the pH rises to about 8.5 after one year of reaction. The release of soluble elements in MgCl 2 solution, under Si-saturated conditions, is proportional to the surface area of the sample and the release increases at 190 C according to a t 1/2 rate law. This time dependence may be an indication of diffusion controlled matrix dissolution. However, at 110 C the release of the mobile elements cannot be described by a t 1/2 rate law as the time exponents are much lower than 0.5. This difference in corrosion behavior may be explained by the higher pH of about 5 at 110 C. In case of NaCl solution under alkaline conditions, the release of soluble elements is not proportional to the surface area of the sample and it increases with time exponents much lower than 0.5. After one year of reaction at 190 C a sharp increase of the release values of some elements was observed. This increase might be explained by the high pH of the solution attained after one year. The corrosion mechanism in NaCl solution, as well as in MgCl 2 solution at 110 C, has not yet been explained. By corrosion experiments in water at constant pH values between 2 and 10, it could be shown that the time exponents of the release of Li and B decrease with increasing pH of the solution. This result can explain qualitatively the differences found in the corrosion behavior of the glass under the various conditions

  18. Crystallization In High Level Waste (HLW) Glass Melters: Operational Experience From The Savannah River Site

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

    2014-02-27

    processing strategy for the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The basis of this alternative approach is an empirical model predicting the crystal accumulation in the WTP glass discharge riser and melter bottom as a function of glass composition, time, and temperature. When coupled with an associated operating limit (e.g., the maximum tolerable thickness of an accumulated layer of crystals), this model could then be integrated into the process control algorithms to formulate crystal tolerant high level waste (HLW) glasses targeting higher waste loadings while still meeting process related limits and melter lifetime expectancies. This report provides a review of the scaled melter testing that was completed in support of the Defense Waste Processing Facility (DWPF) melter. Testing with scaled melters provided the data to define the DWPF operating limits to avoid bulk (volume) crystallization in the un-agitated DWPF melter and provided the data to distinguish between spinels generated by K-3 refractory corrosion versus spinels that precipitated from the HLW glass melt pool. This report includes a review of the crystallization observed with the scaled melters and the full scale DWPF melters (DWPF Melter 1 and DWPF Melter 2). Examples of actual DWPF melter attainment with Melter 2 are given. The intent is to provide an overview of lessons learned, including some example data, that can be used to advance the development and implementation of an empirical model and operating limit for crystal accumulation for WTP. Operation of the first and second (current) DWPF melters has demonstrated that the strategy of using a liquidus temperature predictive model combined with a 100 °C offset from the normal melter operating temperature of 1150 °C (i.e., the predicted liquidus temperature (TL) of the glass must be 1050 °C or less) has been successful in preventing any detrimental accumulation of spinel in the DWPF melt pool, and spinel has not been

  19. Glass formulation development and testing for the vitrification of DWPF HLW sludge coupled with crystalline silicotitanate (CST)

    Andrews, M.K.; Workman, P.J.

    1997-01-01

    An alternative to the In Tank Precipitation and sodium titanate processes at the Savannah River Site is the removal of cesium, strontium, and plutonium from the tank supernate by ion exchange using crystalline silicotitanate (CST). This inorganic material has been shown to effectively and selectively sorb these elements from supernate. The loaded CST could then be immobilized with High-Level Waste (HLW) sludge during vitrification. Initial efforts on the development of a glass formulation for a coupled waste stream indicate that reasonable loadings of both sludge and CST can be achieved in glass

  20. Final Report - Melt Rate Enhancement for High Aluminum HLW Glass Formulation, VSL-08R1360-1, Rev. 0, dated 12/19/08

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

    2013-11-13

    The principal objective of the work reported here was to develop and identify HLW glass compositions that maximize waste processing rates for the aluminum limted waste composition specified by ORP while maintaining high waste loadings and acceptable glass properties. This was accomplished through a combination of crucible-scale tests, confirmation tests on the DM100 melter system, and demonstration at pilot scale (DM1200). The DM100-BL unit was selected for these tests since it was used previously with the HLW waste streams evaluated in this study, was used for tests on HLW glass compositions to support subsequent tests on the HLW Pilot Melter, conduct tests to determine the effect of various glass properties (viscosity and conductivity) and oxide concentrations on glass production rates with HLW feed streams, and to assess the volatility of cesium and technetium during the vitrification of an HLW AZ-102 composition. The same melter was selected for the present tests in order to maintain comparisons between the previously collected data. These tests provide information on melter processing characteristics and off-gas data, including formation of secondary phases and partitioning. Once DM100 tests were completed, one of the compositions was selected for further testing on the DM1200; the DM1200 system has been used for processing a variety of simulated Hanford waste streams. Tests on the larger melter provide processing data at one third of the scale of the actual WTP HLW melter and, therefore, provide a more accurate and reliable assessment of production rates and potential processing issues. The work focused on maximizing waste processing rates for high aluminum HLW compositions. In view of the diversity of forms of aluminum in the Hanford tanks, tests were also conducted on the DM100 to determine the effect of changes in the form of aluminum on feed properties and production rate. In addition, the work evaluated the effect on production rate of modest increases

  1. Advances in Glass Formulations for Hanford High-Alumimum, High-Iron and Enhanced Sulphate Management in HLW Streams-13000

    Kruger, Albert A.

    2013-01-01

    The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet Hanford Tank Waste Treatment and Immobilization Plant (WTP) Contract terms. The WTP's overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulphur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings and higher throughput efficiencies. Results of this work have demonstrated the feasibility of increases in waste loading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. In view of the importance of aluminum limited waste streams at Hanford (and also Savannah River), the ability to achieve high waste loadings without adversely impacting melt rates has the potential for enormous cost savings from reductions in canister count and the potential for schedule acceleration. Consequently, the potential return on the investment made in the development of these enhancements is extremely favorable. Glass composition development for one of the latest Hanford HLW projected compositions with sulphate concentrations high enough to limit waste loading have been successfully tested and show tolerance for previously unreported tolerance for sulphate. Though a significant increase in waste loading for high-iron wastes has been achieved, the magnitude of the increase is not as substantial as those achieved for high-aluminum, high-chromium, high-bismuth or sulphur

  2. Development Of High Waste-Loading HLW Glasses For High Bismuth Phosphate Wastes, VSL-12R2550-1, Rev 0

    Kruger, A. A.; Pegg, Ian L.; Gan, Hao; Kot, Wing K.

    2012-01-01

    This report presents results from tests with new glass formulations that have been developed for several high Bi-P HLW compositions that are expected to be processed at the WTP that have not been tested previously. WTP HLW feed compositions were reviewed to select waste batches that are high in Bi-P and that are reasonably distinct from the Bi-limited waste that has been tested previously. Three such high Bi-P HLW compositions were selected for this work. The focus of the present work was to determine whether the same type of issues as seen in previous work with high-Bi HLW will be seen in HLW with different concentrations of Bi, P and Cr and also whether similar glass formulation development approaches would be successful in mitigating these issues. New glass compositions were developed for each of the three representative Bi-P HLW wastes and characterized with respect to key processing and product quality properties and, in particular, those relating to crystallization and foaming tendency

  3. Development Of High Waste-Loading HLW Glasses For High Bismuth Phosphate Wastes, VSL-12R2550-1, Rev 0

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Gan, Hao [The Catholic University of America, Washington, DC (United States); Kot, Wing K. [The Catholic University of America, Washington, DC (United States)

    2012-12-13

    This report presents results from tests with new glass formulations that have been developed for several high Bi-P HLW compositions that are expected to be processed at the WTP that have not been tested previously. WTP HLW feed compositions were reviewed to select waste batches that are high in Bi-P and that are reasonably distinct from the Bi-limited waste that has been tested previously. Three such high Bi-P HLW compositions were selected for this work. The focus of the present work was to determine whether the same type of issues as seen in previous work with high-Bi HLW will be seen in HLW with different concentrations of Bi, P and Cr and also whether similar glass formulation development approaches would be successful in mitigating these issues. New glass compositions were developed for each of the three representative Bi-P HLW wastes and characterized with respect to key processing and product quality properties and, in particular, those relating to crystallization and foaming tendency.

  4. Advances in Glass Formulations for Hanford High-Aluminum, High-Iron and Enhanced Sulphate Management in HLW Streams - 13000

    Kruger, Albert A. [WTP Engineering Division, United States Department of Energy, Office of River Protection, Post Office Box 450, Richland, Washington 99352 (United States)

    2013-07-01

    The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet Hanford Tank Waste Treatment and Immobilization Plant (WTP) Contract terms. The WTP's overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulphur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings and higher throughput efficiencies. Results of this work have demonstrated the feasibility of increases in waste loading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. In view of the importance of aluminum limited waste streams at Hanford (and also Savannah River), the ability to achieve high waste loadings without adversely impacting melt rates has the potential for enormous cost savings from reductions in canister count and the potential for schedule acceleration. Consequently, the potential return on the investment made in the development of these enhancements is extremely favorable. Glass composition development for one of the latest Hanford HLW projected compositions with sulphate concentrations high enough to limit waste loading have been successfully tested and show tolerance for previously unreported tolerance for sulphate. Though a significant increase in waste loading for high-iron wastes has been achieved, the magnitude of the increase is not as substantial as those achieved for high-aluminum, high-chromium, high-bismuth or

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

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

    2008-01-01

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

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

    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

  7. JSS project phase 4: Experimental and modelling studies of HLW glass dissolution in repository environments

    1987-10-01

    A goal of the JSS project was to develop a scientific basis for understanding the effects of waste package components, groundwater chemistry, and other repository conditions on glass dissolution behaviour, and to develop and refine a model for the processes governing glass dissolution. The fourth phase of the project, which was performed by the Hahn-Meitner-Institut, Berlin, FRG, dealt specifically with model development and application. Phase 4 also adressed whether basaltic glasses could serve as natural analogues for nuclear waste glasses, thus providing a means to test the capability of the model for long-term predictions. Additional experiments were performed in order to complete the data base necessary to model interactions between the glass and bentonite and between glass and steel corrosion products. More data on temperature, S/V, and pH dependence of the glass/water reaction were also collected. In this report, the data acquired during phase 4 are presented and discussed. (orig./DG)

  8. DATA SUMMARY REPORT SMALL SCALE MELTER TESTING OF HLW ALGORITHM GLASSES MATRIX1 TESTS VSL-07S1220-1 REV 0 7/25/07

    KRUGER AA; MATLACK KS; PEGG IL

    2011-12-29

    Eight tests using different HLW feeds were conducted on the DM100-BL to determine the effect of variations in glass properties and feed composition on processing rates and melter conditions (off-gas characteristics, glass processing, foaming, cold cap, etc.) at constant bubbling rate. In over seven hundred hours of testing, the property extremes of glass viscosity, electrical conductivity, and T{sub 1%}, as well as minimum and maximum concentrations of several major and minor glass components were evaluated using glass compositions that have been tested previously at the crucible scale. Other parameters evaluated with respect to glass processing properties were +/-15% batching errors in the addition of glass forming chemicals (GFCs) to the feed, and variation in the sources of boron and sodium used in the GFCs. Tests evaluating batching errors and GFC source employed variations on the HLW98-86 formulation (a glass composition formulated for HLW C-106/AY-102 waste and processed in several previous melter tests) in order to best isolate the effect of each test variable. These tests are outlined in a Test Plan that was prepared in response to the Test Specification for this work. The present report provides summary level data for all of the tests in the first test matrix (Matrix 1) in the Test Plan. Summary results from the remaining tests, investigating minimum and maximum concentrations of major and minor glass components employing variations on the HLW98-86 formulation and glasses generated by the HLW glass formulation algorithm, will be reported separately after those tests are completed. The test data summarized herein include glass production rates, the type and amount of feed used, a variety of measured melter parameters including temperatures and electrode power, feed sample analysis, measured glass properties, and gaseous emissions rates. More detailed information and analysis from the melter tests with complete emission chemistry, glass durability, and

  9. Degradation of glass in the soil

    Romich, H.; Gerlach, S.; Mottner, P. [Fraunhofer-Institut fur Silicatforschung (ISC), Wertheim-Bronnbach (Germany)

    2004-07-01

    Full text of publication follows: Glass has been produced and used in Europe for over 2000 years. Glass objects from the Roman period onwards have been excavated during the last centuries. In general, Roman glass is chemically quite stable, and often the only sign of chemical alteration is an iridescent surface, caused by the leaching of cations, which leads to the formation of a hydrated silica-rich layer. Medieval potash glasses are much less durable, and their surfaces are often found deeply leached, sometimes to a point that no unaltered glass remains. These surfaces may be coherent, though fragile, or they are laminar, with no cohesion between the layers at all. In this study an analytical examination of a series of fragments of archaeological glass retrieved from different sites near Cologne and Stuttgart (Germany) has been carried out. Samples of corroded glasses were analysed by optical microscopy and SEM/EDX (surface and cross sections) in order to obtain information about the chemical composition of the bulk glass and the weathered layers. Since the environmental parameters have constantly varied for archaeological objects, mechanistic studies have to rely on laboratory experiments under controlled conditions. For an extensive exposure programme standardised soil or natural garden earth was used, for which the pH was modified. Several corrosion sensitive potash-lime silicate glasses have been designed to study the effect of glass composition. A model glass consisting of SiO{sub 2} (54.2), CaO (28.8) and K{sub 2}O (17.0 weight-%) mostly lead to the formation of a crust on the leached layer, with a total thickness of 100 micrometer (for soil with pH 7 to 8, 12 months exposure). Model glasses also containing Al, Mg and P have built up preferably laminated structures (total thickness up to 200 micrometer). This presentation will give an overview about the variety of degradation phenomena observed on originals and compare the results with controlled laboratory

  10. In situ corrosion tests on HLW glass as part of a larger approach

    Van Iseghem, P.

    1997-01-01

    In-situ corrosion tests were performed on various candidate high-level waste glasses in the underground laboratory in clay underneath SCK x CEN. The tests exposed the glass samples directly to the Boom clay rock, for maximum durations of 7.5 years. We succeeded to interpret the corrosion data at 90 deg C in terms of dissolution mechanisms, and we concluded that the glass composition has a determining effect on the corrosion stability. The data from our in-situ tests were of high relevance for estimating the long-term behaviour of the glasses. The long-term in-situ tests provide corrosion data which show different trends than other corrosion tests, e.g. shorter duration tests in Boom clay, or tests in deionized water. The initial dissolution rate using MCC1 test at 90 deg C is about the same for the three glasses discussed, but the longest duration in Boom clay at 90 deg C shows a difference in mass loss of about 25 times. We finally present some ideas on how the corrosion tests can meet the needs, such as the modelling of the glass corrosion or providing input in the performance assessment. (author)

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

    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

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

    Neumann, W.

    1980-01-01

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

  13. Transuranium elements leaching from simulated HLW glasses in synthetic interstitial claywater

    Wang, L.

    1992-08-01

    The main objective of this Master Thesis is to measure the steady-state concentrations of Pu, Np, and Am upon the leaching of High-Level Waste Glass in two types of synthetic claywater: humic acid free and humic acid containing synthetic claywater. The synthetic claywater has a composition that is representative for the in-situ interstitial groundwater of the Boom clay formation, a potential geological repository of radioactive waste in Belgium. The steady-state concentrations of transuranium elements were measured by leaching experiments with a typical duration of 400 days. Five main conclusions are drawn from the experimental data. (1) The transuranium elements that are released from simulated High Level Waste Glass are dominantly present in the synthetic claywater solutions as colloids. These colloids are smaller than 2 nm in absence of humic acids. In the presence of humic acids however, the colloids interact with actinides (adsorb or coagulate) and form particles larger than 2 nm. Np and Am are associated with inorganic and organic colloids in the synthetic interstitial claywater solution whereas Pu forms only inorganic colloids. (2) The steady-state concentration of Pu is in good agreement with the solubility of the Pu compound PuO 2 .xH 2 O. It is therefore concluded that PuO 2 .xH 2 O is the solubility controlling phase. (3) The Pu(IV)-species are dominant in the leaching solutions. Carbonate and humic acid complexes are negligible. (4) The steady-state concentrations of Np and Am in leaching solutions were much lower than the values calculated on the basis of known thermodynamic data. This indicates that the solubility controlling phases for Np and Am were not correctly identified or that the measured Np and Am concentrations were not steady-state values. (5) Non-active glass leaching tests have indicated that no organic colloids were formed as a result of glass dissolution. (A.S.)

  14. TESTS WITH HIGH-BISMUTH HLW GLASSES FINAL REPORT VSL-10R1780-1, Rev. 0; 12/13/10

    Matlack, K.S.; Kruger, A.A.; Joseph, I.; Gan, H.; Kot, W.K.; Chaudhuri, M.; Mohr, R.K.; Mckeown, D.A.; Bardakei, T.; Gong, W.; Buecchele, A.C.; Pegg, I.L.

    2011-01-01

    This Final Report describes the testing of glass formulations developed for Hanford High Level Waste (HLW) containing high concentrations of bismuth. In previous work on high-bismuth HLW streams specified by the Office of River Protection (ORP), fully compliant, high waste loading compositions were developed and subjected to melter testing on the DM100 vitrification system. However, during heat treatment according to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW canister centerline cooling (CCC) curves, crucible melts of the high-bismuth glasses were observed to foam. Clearly, such an occurrence during cooling of actual HLW canisters would be highly undesirable. Accordingly, the present work involves larger-scale testing to determine whether this effect occurs under more prototypical conditions, as well as crucible-scale tests to determine the causes and potentially remediate the observed foaming behavior. The work included preparation and characterization of crucible melts designed to determine the underlying causes of the foaming behavior as well as to assess potential mitigation strategies. Testing was also conducted on the DM1200 HLW Pilot melter with a composition previously tested on the DM100 and shown to foam during crucible-scale CCC heat treatment. The DM1200 tests evaluated foaming of glasses over a range of bismuth concentrations poured into temperature-controlled, 55-gallon drums which have a diameter that is close to that of the full-scale WTP HLW canisters. In addition, the DM1200 tests provided the first large-scale melter test data on high-bismuth WTP HLW compositions, including information on processing rates, cold cap behavior and off-gas characteristics, and data from this waste composition on the prototypical DM1200 off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for ORP on the same waste composition. The scope of this study was outlined in a Test Plan that was

  15. TESTS WITH HIGH-BISMUTH HLW GLASSES FINAL REPORT VSL-10R1780-1 REV 0 12/13/10

    MATLACK KS; KRUGER AA; JOSEPH I; GAN H; KOT WK; CHAUDHURI M; MOHR RK; MCKEOWN DA; BARDAKEI T; GONG W; BUECCHELE AC; PEGG IL

    2011-01-05

    This Final Report describes the testing of glass formulations developed for Hanford High Level Waste (HLW) containing high concentrations of bismuth. In previous work on high-bismuth HLW streams specified by the Office of River Protection (ORP), fully compliant, high waste loading compositions were developed and subjected to melter testing on the DM100 vitrification system. However, during heat treatment according to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW canister centerline cooling (CCC) curves, crucible melts of the high-bismuth glasses were observed to foam. Clearly, such an occurrence during cooling of actual HLW canisters would be highly undesirable. Accordingly, the present work involves larger-scale testing to determine whether this effect occurs under more prototypical conditions, as well as crucible-scale tests to determine the causes and potentially remediate the observed foaming behavior. The work included preparation and characterization of crucible melts designed to determine the underlying causes of the foaming behavior as well as to assess potential mitigation strategies. Testing was also conducted on the DM1200 HLW Pilot melter with a composition previously tested on the DM100 and shown to foam during crucible-scale CCC heat treatment. The DM1200 tests evaluated foaming of glasses over a range of bismuth concentrations poured into temperature-controlled, 55-gallon drums which have a diameter that is close to that of the full-scale WTP HLW canisters. In addition, the DM1200 tests provided the first large-scale melter test data on high-bismuth WTP HLW compositions, including information on processing rates, cold cap behavior and off-gas characteristics, and data from this waste composition on the prototypical DM1200 off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for ORP on the same waste composition. The scope of this study was outlined in a Test Plan that was

  16. Structure and Degradation Behaviour of Calcium Phosphate Glasses

    Silva, A M B; Correia, R N; Fernandes, M H V; Oliveira, J M M

    2011-01-01

    Some studies have shown a relationship between glass structure and in vitro mineralization, generally associated with the rate of glass degradation, nature of released ions and subsequent Ca-P precipitation on glass surfaces when immersed in a Simulated Body Fluid (SBF). The knowledge of the ionic species distribution in glasses and of the involved bond strengths can be used to assess the in vitro behaviour of a glass. The role of ions such as silicon or titanium is of major importance for the development of new compositions and also for the control of glass degradation behaviour. A comparative study with two calcium phosphate glasses series was performed: Both glasses series - one with Si and another with Ti - include P 2 O 5 and alkaline earth ions in their compositions. Surface reactivity of glasses from the SiO 2 -containing system have been studied in SBF showing the precipitation of a Ca-P surface layer that increases with increasing MgO/CaO ratio. In glasses from the TiO 2 -containing series it is shown that the increase of TiO 2 contributes for the stabilization of the glass network thus allowing the control of their degradation rate when immersed in SBF. The relationship between structural features of these calcium-phosphate glasses and their degradation behaviour in SBF is discussed in terms of the structural role of Si and Ti ions. It is concluded that glasses with less interconnected species favour the Ca-P surface precipitation. The understanding of this relationship in synthetic physiological fluids is expected to allow the tailoring of glass degradation rates in complex biological systems.

  17. Use of natural and archaeological analogs to validate long - term behaviour of HLW glass in geological disposal conditions

    Gin, S.; Verney-Carron, A.; Libourel, G.

    2008-01-01

    Some old basaltic and Roman glasses have been studied in order to validate the predictive models developed for assessing the long-term behaviour of nuclear glass in geological repository conditions. Leaching behaviour of basaltic glass altered in both laboratory and natural environment conditions allows to validate the key mechanisms that control glass dissolution kinetics and the order of magnitude of glass packages lifetime In a stable clayey formation (French reference concept for a geological disposal of high level waste). The study of Roman glass blocks (with the same geometry as nuclear glass package) altered during 1800 years in a marine environment gives new insight on the basic mechanisms involved in confined media (fractures and small cracks). Results show the importance of the coupling between transport of reactive species and chemical reactions. This study, still in progress, would allow to validate the modelling of such a complex system. (author)

  18. R7T7-type HLW glass alteration under irradiation. Study of the residual alteration rate regime

    Rolland, Severine

    2012-01-01

    In France, fission products and minor actinides remaining after reprocessing of spent nuclear fuel are confined in a borosilicate glass matrix, named R7T7, for disposal in a geological repository. However, in these conditions, after several thousand years, water could arrive in contact with glass and be radio-lysed. In this work, we investigated the irradiation influence and especially the influence of the energy deposition on the residual glass alteration rate regime in pure water. Two types of leaching tests have been carried out. The first were performed on radioactive glass and the second on a SON68 glass (nonradioactive surrogate of R7T7 glass) under external irradiation γ. (author) [fr

  19. Alumina particle degradation during solid particle impact on glass

    Slikkerveer, P.J.; Veld, in 't H.; Verspui, M.A.; With, de G.; Reefman, D.

    2000-01-01

    Particle degradation limits the reuse of powders in industrial powder-blast processes. In this paper the degradation behavior of Al2O3 powder is studied during erosion of glass substrates. Three techniques were used on original and multiply used powders: particle size measurements, single particle

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

    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

  1. Degradation of glass artifacts: application of modern surface analytical techniques.

    Melcher, Michael; Wiesinger, Rita; Schreiner, Manfred

    2010-06-15

    A detailed understanding of the stability of glasses toward liquid or atmospheric attack is of considerable importance for preserving numerous objects of our cultural heritage. Glasses produced in the ancient periods (Egyptian, Greek, or Roman glasses), as well as modern glass, can be classified as soda-lime-silica glasses. In contrast, potash was used as a flux in medieval Northern Europe for the production of window panes for churches and cathedrals. The particular chemical composition of these potash-lime-silica glasses (low in silica and rich in alkali and alkaline earth components), in combination with increased levels of acidifying gases (such as SO(2), CO(2), NO(x), or O(3)) and airborne particulate matter in today's urban or industrial atmospheres, has resulted in severe degradation of important cultural relics, particularly over the last century. Rapid developments in the fields of microelectronics and computer sciences, however, have contributed to the development of a variety of nondestructive, surface analytical techniques for the scientific investigation and material characterization of these unique and valuable objects. These methods include scanning electron microscopy in combination with energy- or wavelength-dispersive spectrometry (SEM/EDX or SEM/WDX), secondary ion mass spectrometry (SIMS), and atomic force microscopy (AFM). In this Account, we address glass analysis and weathering mechanisms, exploring the possibilities (and limitations) of modern analytical techniques. Corrosion by liquid substances is well investigated in the glass literature. In a tremendous number of case studies, the basic reaction between aqueous solutions and the glass surfaces was identified as an ion-exchange reaction between hydrogen-bearing species of the attacking liquid and the alkali and alkaline earth ions in the glass, causing a depletion of the latter in the outermost surface layers. Although mechanistic analogies to liquid corrosion are obvious, atmospheric

  2. Final Report - Glass Formulation Development and Testing for DWPF High AI2O3 HLW Sludges, VSL-10R1670-1, Rev. 0, dated 12/20/10

    Kruger, Albert A.; Pegg, I. L.; Kot, W. K.; Gan, H.; Matlack, K. S.

    2013-11-13

    The principal objective of the work described in this Final Report is to develop and identify glass frit compositions for a specified DWPF high-aluminum based sludge waste stream that maximizes waste loading while maintaining high production rate for the waste composition provided by ORP/SRS. This was accomplished through a combination of crucible-scale, vertical gradient furnace, and confirmation tests on the DM100 melter system. The DM100-BL unit was selected for these tests. The DM100-BL was used for previous tests on HLW glass compositions that were used to support subsequent tests on the HLW Pilot Melter. It was also used to process compositions with waste loadings limited by aluminum, bismuth, and chromium, to investigate the volatility of cesium and technetium during the vitrification of an HLW AZ-102 composition, to process glass formulations at compositional and property extremes, and to investigate crystal settling on a composition that exhibited one percent crystals at 963{degrees}C (i.e., close to the WTP limit). The same melter was selected for the present tests in order to maintain comparisons between the previously collected data. The tests provide information on melter processing characteristics and off-gas data, including formation of secondary phases and partitioning. Specific objectives for the melter tests are as follows: Determine maximum glass production rates without bubbling for a simulated SRS Sludge Batch 19 (SB19). Demonstrate a feed rate equivalent to 1125 kg/m{sup 2}/day glass production using melt pool bubbling. Process a high waste loading glass composition with the simulated SRS SB19 waste and measure the quality of the glass product. Determine the effect of argon as a bubbling gas on waste processing and the glass product including feed processing rate, glass redox, melter emissions, etc.. Determine differences in feed processing and glass characteristics for SRS SB19 waste simulated by the co-precipitated and direct

  3. Degradation of partially immersed glass: A new perspective

    Chinnam, R. K.; Fossati, P. C. M.; Lee, W. E.

    2018-05-01

    The International Simple Glass (ISG) is a six-component borosilicate glass which was developed as a reference for international collaborative studies on high level nuclear waste encapsulation. Its corrosion behaviour is typically examined when it is immersed in a leaching solution, or when it is exposed to water vapour. In this study, an alternative situation is considered in which the glass is only partially immersed for 7 weeks at a temperature of 90 °C. In this case, half of the glass sample is directly in the solution itself, and the other half is in contact with a water film formed by condensation of water vapour that evaporated from the solution. This results in a different degradation behaviour compared to standard tests in which the material is fully immersed. In particular, whilst in standard tests the system reaches a steady state with a very low alteration rate thanks to the formation of a protective gel layer, in partially-immersed tests this steady state could not be reached because of the continuous alteration from the condensate water film. The constant input of ions from the emerged part of the sample caused a supersaturation of the solution, which resulted in early precipitation of secondary crystalline phases. This setup mimics storage conditions once small amounts of water have entered a glass waste form containing canister. It offers a more realistic outlook of corrosion mechanisms happening in such situations than standard fully-immersed corrosion tests.

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

    Vandiver, P.B.

    1994-01-01

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

  5. Glass fibres reinforced polyester composites degradation monitoring by surface analysis

    Croitoru, Catalin [“Transilvania” University of Brasov, Materials Engineering and Welding Department, Eroilor 29 Str., 500036 Brasov (Romania); Patachia, Silvia, E-mail: st.patachia@unitbv.ro [“Transilvania” University of Brasov, Product Design Environment and Mechatronics Department, Eroilor 29 Str., 500036 Brasov (Romania); Papancea, Adina [“Transilvania” University of Brasov, Product Design Environment and Mechatronics Department, Eroilor 29 Str., 500036 Brasov (Romania); Baltes, Liana; Tierean, Mircea [“Transilvania” University of Brasov, Materials Engineering and Welding Department, Eroilor 29 Str., 500036 Brasov (Romania)

    2015-12-15

    Highlights: • Glass fibre-reinforced polyester composites surface analysis by photographic method. • The composites are submitted to accelerated ageing by UV irradiation at 254 nm. • The UV irradiation promotes differences in the surface chemistry of the composites. • MB dye is differently adsorbed on surfaces with different degradation degrees. • Good correlation between the colouring degree and surface chemistry. - Abstract: The paper presents a novel method for quantification of the modifications that occur on the surface of different types of gel-coated glass fibre-reinforced polyester composites under artificial UV-ageing at 254 nm. The method implies the adsorption of an ionic dye, namely methylene blue, on the UV-aged composite, and computing the CIELab colour space parameters from the photographic image of the coloured composite's surface. The method significantly enhances the colour differences between the irradiated composites and the reference, in contrast with the non-coloured ones. The colour modifications that occur represent a good indicative of the surface degradation, alteration of surface hydrophily and roughness of the composite and are in good correlation with the ATR-FTIR spectroscopy and optical microscopy results. The proposed method is easier, faster and cheaper than the traditional ones.

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

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

    1998-09-01

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

  7. Memo, "Incorporation of HLW Glass Shell V2.0 into the Flowsheets," to ED Lee, CCN: 184905, October 20, 2009

    Gimpel, Rodney F.; Kruger, Albert A.

    2013-12-18

    Efforts are being made to increase the efficiency and decrease the cost of vitrifying radioactive waste stored in tanks at the U.S. Department of Energy Hanford Site. The compositions of acceptable and processable high-level waste (HL W) glasses need to be optimized to minimize the waste-form volume and, hence, to reduce cost. A database of glass properties of waste glass and associated simulated waste glasses was collected and documented in PNNL 18501, Glass Property Data and Models for Estimating High-Level Waste Glass Volume and glass property models were curve-fitted to the glass compositions. A routine was developed that estimates HL W glass volumes using the following glass property models: II Nepheline, II One-Percent Crystal Temperature (T1%), II Viscosity (11) II Product Consistency Tests (PCT) for boron, sodium, and lithium, and II Liquidus Temperature (TL). The routine, commonly called the HL W Glass Shell, is presented in this document. In addition to the use of the glass property models, glass composition constraints and rules, as recommend in PNNL 18501 and in other documents (as referenced in this report) were incorporated. This new version of the HL W Glass Shell should generally estimate higher waste loading in the HL W glass than previous versions.

  8. Characterization of HLW glass samples Task 3 Characterization of radioactive waste forms a series of final reports (1985-89) No 20

    Malow, G.; Behrend, U.; Schubert, P.

    1991-01-01

    Due to a delay in the melting of the highly radioactive SON68 glass, a short-term post-investigation of the highly radioactive glass from the Pamela plant in Mol (Belgium) has been carried out, the aim being a check-up of the active LEWC glass SM 513 LW11. The results were compared with those obtained for non-radioactive glass samples. The final report of the present CEC programme shortly describes the planned investigations of the glass R7T7 for the whole period of the research contract and the results of the short-term post-investigation of the Pamela glass. 11 refs.; 9 figs.; 4 tabs

  9. MIIT: International in-situ testing of simulated HLW forms--preliminary analyses of SRL 165/TDS waste glass and metal systems

    Wicks, G.G.; Lodding, A.R.; Macedo, P.B.; Molecke, M.A.

    1989-01-01

    The first in-situ tests involving burial of simulated high-level waste (HLW) forms conducted in the United States were started on July 22, 1986. This effort, called the Materials Interface Interactions Tests (MIIT), comprises the largest, most cooperative field testing venture in the international waste management community. Included in the study are over 900 waste form samples comprising 15 different systems supplied by seven countries. Also included are almost 300 potential canister or overpack metal samples of 11 different metals along with more than 500 geologic and backfill specimens. There are a total of 1926 relevant interactions that characterize this effort which is being conducted in the bedded salt site at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico

  10. COMSOL Multiphysics Model for HLW Canister Filling

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

    2016-04-11

    The U.S. Department of Energy (DOE) is building a Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is being temporarily stored in 177 underground tanks. Efforts are being made to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. Wastes containing high concentrations of Al2O3 and Na2O can contribute to nepheline (generally NaAlSiO4) crystallization, which can sharply reduce the chemical durability of high level waste (HLW) glass. Nepheline crystallization can occur during slow cooling of the glass within the stainless steel canister. The purpose of this work was to develop a model that can be used to predict temperatures of the glass in a WTP HLW canister during filling and cooling. The intent of the model is to support scoping work in the laboratory. It is not intended to provide precise predictions of temperature profiles, but rather to provide a simplified representation of glass cooling profiles within a full scale, WTP HLW canister under various glass pouring rates. These data will be used to support laboratory studies for an improved understanding of the mechanisms of nepheline crystallization. The model was created using COMSOL Multiphysics, a commercially available software. The model results were compared to available experimental data, TRR-PLT-080, and were found to yield sufficient results for the scoping nature of the study. The simulated temperatures were within 60 ºC for the centerline, 0.0762m (3 inch) from centerline, and 0.2286m (9 inch) from centerline thermocouples once the thermocouples were covered with glass. The temperature difference between the experimental and simulated values reduced to 40 ºC, 4 hours after the thermocouple was covered, and down to 20 ºC, 6 hours after the thermocouple was covered

  11. Technetium Chemistry in HLW

    Hess, Nancy J.; Felmy, Andrew R.; Rosso, Kevin M.; Xia Yuanxian

    2005-01-01

    Tc contamination is found within the DOE complex at those sites whose mission involved extraction of plutonium from irradiated uranium fuel or isotopic enrichment of uranium. At the Hanford Site, chemical separations and extraction processes generated large amounts of high level and transuranic wastes that are currently stored in underground tanks. The waste from these extraction processes is currently stored in underground High Level Waste (HLW) tanks. However, the chemistry of the HLW in any given tank is greatly complicated by repeated efforts to reduce volume and recover isotopes. These processes ultimately resulted in mixing of waste streams from different processes. As a result, the chemistry and the fate of Tc in HLW tanks are not well understood. This lack of understanding has been made evident in the failed efforts to leach Tc from sludge and to remove Tc from supernatants prior to immobilization. Although recent interest in Tc chemistry has shifted from pretreatment chemistry to waste residuals, both needs are served by a fundamental understanding of Tc chemistry

  12. MIIT: International in-situ testing of simulated HLW forms - performance of SRS simulated waste glass after 6 mos., 1 yr., 2 yrs. and 5 yrs. of burial at WIPP

    Wicks, G.G.; Lodding, A.R.; Macedo, P.B.; Clark, D.E.

    1991-01-01

    The first field test, involving burial of simulated high-level waste (HLW) forms and package components, to be conducted in the United States, was begun in July of 1986. This program, called the Materials Interface Interactions Test or MIIT, comprises the largest cooperative field-testing venture in the international waste management community. Included in the study are over 900 waste form samples comprising 15 different systems supplied by 7 countries. Also included are about 300 potential canister or overpack metal samples along with more than 500 geologic and backfill specimens. There are almost 2000 relevant interactions that characterize this effort which is being conducted in the bedded salt site at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico. The MIIT program represents a joint effort managed by Sandia National Laboratories in Albuquerque, N.M., and Savannah River Laboratory in Aiken, S.C. and sponsored by the US Department of Energy. Also involved in MIIT are participants from various laboratories and universities in France, Germany, Belgium, Canada, Japan, Sweden, the United Kingdom, and the United States. In July of 1991, the experimental portion of the 5-yr. MIIT program was completed. Although only about 5% of all MIIT samples have been assessed thus far, there are already interesting findings that have emerged. The present paper will discuss results obtained for SRS 165/TDS waste glass after burial of 6 mo., 1 yr. and 2 yrs., along with initial analyses of 5 yr. samples

  13. Industrial scale-plant for HLW partitioning in Russia

    Dzekun, E.G.; Glagolenko, Y.V.; Drojko, E.G.; Kurochkin, A.I.

    1996-01-01

    Radiochemical plant of PA > at Ozersk, which was come on line in December 1948 originally for weapon plutonium production and reoriented on the reprocessing of spent fuel, till now keeps on storage HLW of the military program. Application of the vitrification method since 1986 has not essentially reduced HLW volumes. So, as of September 1, 1995 vitrification installations had been processed 9590 m 3 HLW and 235 MCi of radionuclides was included in glass. However only 1100 m 3 and 20.5 MCi is part of waste of the military program. The reason is the fact, that the technology and equipment of vitrification were developed for current waste of Purex-process, for which low contents of corrosion-dangerous impurity to materials of vitrification installation is characteristic of. With reference to HLW, which are growing at PA > in the course of weapon plutonium production, the program of Science-Research Works includes the following main directions of work. Development of technology and equipment of installations for immobilising HLW with high contents of impurity into a solid form at induction melter. Application of High-temperature Adsorption Method for sorption of radionuclides from HLW on silica gel. Application of Partitioning Method of radionuclides from HLW, based on extraction cesium and strontium into cobalt dicarbollyde or crown-ethers, but also on recovery of cesium radionuclides by sorption on inorganic sorbents. In this paper the results of work on creation of first industrial scale-plant for partitioning HLW by the extraction and sorption methods are reported

  14. Source term measurements on vitrified HLW

    Hough, A.; Marples, J.A.C.

    1988-01-01

    The equilibrium concentrations of Tc-99, Np-237, Pu-239/240 and Am-241 have been measured in the presence of materials likely to be present in a vitrified HLW repository: glass, iron, backfill and rock. Results were measured under both oxidising and reducing conditions and at pH values set by the backfill bentonite and cement. Under reducing conditions and with cementitious backfills, the equilibrium concentrations ranged from three to 30 times allowed drinking water levels for the four isotopes. (author)

  15. HLW Disposal System Development

    Choi, J. W.; Choi, H. J.; Lee, J. Y. (and others)

    2007-06-15

    A KRS is suggested through design requirement analysis of the buffer and the canister which are the constituent of disposal system engineered barrier and HLW management plans are proposed. In the aspect of radionuclide retention capacity, the thickness of the buffer is determined 0.5m, the shape to be disc and ring and the dry density to be 1.6 g/cm{sup 3}. The maximum temperature of the buffer is below 100 .deg. which meets the design requirement. And bentonite blocks with 5 wt% of graphite showed more than 1.0 W/mK of thermal conductivity without the addition of sand. The result of the thermal analysis for proposed double-layered buffer shows that decrease of 7 .deg. C in maximum temperature of the buffer. For the disposal canister, the copper for the outer shell material and cast iron for the inner structure material is recommended considering the results analyzed in terms of performance of the canisters and manufacturability and the geochemical properties of deep groundwater sampled from the research area with granite, salt water intrusion, and the heavy weight of the canister. The results of safety analysis for the canister shows that the criticality for the normal case including uncertainty is the value of 0.816 which meets subcritical condition. Considering nation's 'Basic Plan for Electric Power Demand and Supply' and based on the scenario of disposing CANDU spent fuels in the first phase, the disposal system that the repository will be excavated in eight phases with the construction of the Underground Research Laboratory (URL) beginning in 2020 and commissioning in 2040 until the closure of the repository is proposed. Since there is close correlation between domestic HLW management plans and front-end/back-end fuel cycle plans causing such a great sensitivity of international environment factor, items related to assuring the non-proliferation and observing the international standard are showed to be the influential factor and acceptability

  16. Safety of HLW shipments

    1998-01-01

    The third shipment back to Japan of vitrified high-level radioactive waste (HLW) produced through reprocessing in France is scheduled to take place in early 1998. A consignment last March drew protest from interest groups and countries along the shipping route. Requirements governing the shipment of cargoes of this type and concerns raised by Greenpeace that were assessed by an international expert group, were examined in a previous article. A further report prepared on behalf of Greenpeace Pacific has been released. The paper: Transportation accident of a ship carrying vitrified high-level radioactive waste, Part 1 Impact on the Federated States of Micronesia by Resnikoff and Champion, is dated 31 July 1997. A considerable section of the report is given over to discussion of the economic situation of the Federated Statess of Micronesia, and lifestyle and dietary factors which would influence radiation doses arising from a release. It postulates a worst case accident scenario of a collision between the HLW transport ship and an oil tanker 1 km off Pohnpei with the wind in precisely the direction to result in maximum population exposure, and attempts to assess the consequences. In summary, the report postulates accident and exposure scenarios which are conceivable but not credible. It combines a series of worst case scenarios and attempts to evaluate the consequences. Both the combined scenario and consequences have probabilities of occurrence which are negligible. The shipment carried by the 'Pacific Swan' left Cherbourgon 21 January 1998 and comprised 30 tonnes of reprocessed vitrified waste in 60 stainless steel canisters loaded into three shipping casks. (author)

  17. Mathematical modeling of photoinitiated coating degradation: Effects of coating glass transition temperature and light stabilizers

    Kiil, Søren; G.de With, R.A.T.M.Van Benthem

    2013-01-01

    A mathematical model, describing coating degradation mechanisms of thermoset coatings exposed to ultraviolet radiation and humidity at constant temperature, was extended to simulate the behavior of a coating with a low glass transition temperature. The effects of adding light stabilizers (a UV...

  18. Mechanical behaviour of degradable phosphate glass fibres and composites-a review.

    Colquhoun, R; Tanner, K E

    2015-12-23

    Biodegradable materials are potentially an advantageous alternative to the traditional metallic fracture fixation devices used in the reconstruction of bone tissue defects. This is due to the occurrence of stress shielding in the surrounding bone tissue that arises from the absence of mechanical stimulus to the regenerating bone due to the mismatch between the elastic modulus of bone and the metal implant. However although degradable polymers may alleviate such issues, these inert materials possess insufficient mechanical properties to be considered as a suitable alternative to current metallic devices at sites of sufficient mechanical loading. Phosphate based glasses are an advantageous group of materials for tissue regenerative applications due to their ability to completely degrade in vivo at highly controllable rates based on the specific glass composition. Furthermore the release of the glass's constituent ions can evoke a therapeutic stimulus in vivo (i.e. osteoinduction) whilst also generating a bioactive response. The processing of these materials into fibres subsequently allows them to act as reinforcing agents in degradable polymers to simultaneously increase its mechanical properties and enhance its in vivo response. However despite the various review articles relating to the compositional influences of different phosphate glass systems, there has been limited work summarising the mechanical properties of different phosphate based glass fibres and their subsequent incorporation as a reinforcing agent in degradable composite materials. As a result, this review article examines the compositional influences behind the development of different phosphate based glass fibre compositions intended as composite reinforcing agents along with an analysis of different potential composite configurations. This includes variations in the fibre content, matrix material and fibre architecture as well as other novel composites designs.

  19. High-Level Waste Glass Formulation Model Sensitivity Study 2009 Glass Formulation Model Versus 1996 Glass Formulation Model

    Belsher, J.D.; Meinert, F.L.

    2009-01-01

    This document presents the differences between two HLW glass formulation models (GFM): The 1996 GFM and 2009 GFM. A glass formulation model is a collection of glass property correlations and associated limits, as well as model validity and solubility constraints; it uses the pretreated HLW feed composition to predict the amount and composition of glass forming additives necessary to produce acceptable HLW glass. The 2009 GFM presented in this report was constructed as a nonlinear optimization calculation based on updated glass property data and solubility limits described in PNNL-18501 (2009). Key mission drivers such as the total mass of HLW glass and waste oxide loading are compared between the two glass formulation models. In addition, a sensitivity study was performed within the 2009 GFM to determine the effect of relaxing various constraints on the predicted mass of the HLW glass.

  20. Laser induced breakdown spectroscopy for analysis and characterization of degradation pathologies of Roman glasses

    Palomar, T. [Instituto de Historia, Centro de Ciencias Humanas y Sociales, CSIC, C/Albasanz 26-28, 28037 Madrid (Spain); Oujja, M., E-mail: m.oujja@iqfr.csic.es [Instituto de Química Física Rocasolano, CSIC, C/Serrano 119, 28006 Madrid (Spain); García-Heras, M.; Villegas, M.A. [Instituto de Historia, Centro de Ciencias Humanas y Sociales, CSIC, C/Albasanz 26-28, 28037 Madrid (Spain); Castillejo, M. [Instituto de Química Física Rocasolano, CSIC, C/Serrano 119, 28006 Madrid (Spain)

    2013-09-01

    The feasibility and possibilities of laser induced breakdown spectroscopy (LIBS) in the full study of non-destructible historic glasses have been explored in the present work. Thirteen Roman glass samples, including seven entire glass beads, from the ancient town of Augusta Emerita (SW Spain) were characterized by LIBS in combination with other conventional techniques, such as scanning electron microscopy/energy dispersive X-ray spectrometry, X-ray fluorescence and ultraviolet–visible spectrophotometry. LIBS stratigraphic analysis, carried out by the application of successive laser pulses on the same spot, has been mainly targeted at characterizing particular features of non-destructible historic glasses, such as bulk chemical composition, surface degradation pathologies (dealkalinization layers and deposits), chromophores, and opacifying elements. The obtained data demonstrate that LIBS can be a useful and alternative technique for spectroscopic studies of historical glasses, especially for those conserved under burial conditions and when it deals with studying non-destructible samples. - Highlights: • Determination of chromophores and opacifiers in non-destructible glass by LIBS • Manganese is determined as principal component of dark deposits. • Antimony appears in all decorations while lead is only present in yellow ones. • Stratigraphic analysis enables the identification of dealkalinization layers.

  1. Laser induced breakdown spectroscopy for analysis and characterization of degradation pathologies of Roman glasses

    Palomar, T.; Oujja, M.; García-Heras, M.; Villegas, M.A.; Castillejo, M.

    2013-01-01

    The feasibility and possibilities of laser induced breakdown spectroscopy (LIBS) in the full study of non-destructible historic glasses have been explored in the present work. Thirteen Roman glass samples, including seven entire glass beads, from the ancient town of Augusta Emerita (SW Spain) were characterized by LIBS in combination with other conventional techniques, such as scanning electron microscopy/energy dispersive X-ray spectrometry, X-ray fluorescence and ultraviolet–visible spectrophotometry. LIBS stratigraphic analysis, carried out by the application of successive laser pulses on the same spot, has been mainly targeted at characterizing particular features of non-destructible historic glasses, such as bulk chemical composition, surface degradation pathologies (dealkalinization layers and deposits), chromophores, and opacifying elements. The obtained data demonstrate that LIBS can be a useful and alternative technique for spectroscopic studies of historical glasses, especially for those conserved under burial conditions and when it deals with studying non-destructible samples. - Highlights: • Determination of chromophores and opacifiers in non-destructible glass by LIBS • Manganese is determined as principal component of dark deposits. • Antimony appears in all decorations while lead is only present in yellow ones. • Stratigraphic analysis enables the identification of dealkalinization layers

  2. Interference of different ionic species on the analysis of phosphate in HLW using spectrophotometer

    Mishra, P.K.; Ghongane, D.E.; Valsala, T.P.; Sonavane, M.S.; Kulkarni, Y.; Changrani, R.D.

    2010-01-01

    During reprocessing of spent nuclear fuel by PUREX process different categories of radioactive liquid wastes like High Level (HL), Intermediate Level (IL) and Low Level (LL) are generated. Different methodologies are adopted for management of these wastes. Since PUREX solvent (30% Tri butyl phosphate-70% Normal Paraffin Hydrocarbon) undergoes chemical degradation in the highly acidic medium of dissolver solution, presence of phosphate in the waste streams is inevitable. Since higher concentrations of phosphate in the HLW streams will affect its management by vitrification, knowledge about the concentration of phosphate in the waste is essential before finalising the glass composition. Since a large number of anionic and cationic species are present in the waste, these species may interfere phosphate analysis using spectrophotometer. In the present work, the interference of different anionic and cationic species on the analysis of phosphate in waste solutions using spectrophotometer was studied

  3. The basic corrosion mechanisms of HLW glasses

    Conradt, R.; Roggendorf, H.; Ostertag, R.

    1986-01-01

    During the years 1975 to 1984, the Commission of the European Communities organized and promoted an R and D programme on the testing and evaluation of solidified high-level waste forms with the purpose of providing a scientific basis for the management and storage of radioactive waste. A fair number of materials were tested under a broad variation of experimental data. The Fraunhofer-Institut fuer Silicatforschung, Wuerzburg, has undertaken to perform a synoptic evaluation of the above data. The purpose of this evaluation is: - to compile the data from the individual national contributors (as presented in the joint annual reports of the EC) with respect to: the materials, or the experimental parameters, or further aspects, and to harmonize them with respect to their presentation, choice of units, etc., - to compare the results to the international state of information, - to elaborate and demonstrate common features of the diverse materials, e.g. common patterns of the corrosion behaviour, - to check the validity of present models, - to define shortcomings and questions that are still open

  4. Redox Control For Hanford HLW Feeds VSL-12R2530-1, REV 0

    Kruger, A. A.; Matlack, Keith S.; Pegg, Ian L.; Kot, Wing K.; Joseph, Innocent

    2012-01-01

    The principal objectives of this work were to investigate the effects of processing simulated Hanford HLW at the estimated maximum concentrations of nitrates and oxalates and to identify strategies to mitigate any processing issues resulting from high concentrations of nitrates and oxalates. This report provides results for a series of tests that were performed on the DM10 melter system with simulated C-106/AY-102 HLW. The tests employed simulated HLW feeds containing variable amounts of nitrates and waste organic compounds corresponding to maximum concentrations proj ected for Hanford HLW streams in order to determine their effects on glass production rate, processing characteristics, glass redox conditions, melt pool foaming, and the tendency to form secondary phases. Such melter tests provide information on key process factors such as feed processing behavior, dynamic effects during processing, processing rates, off-gas amounts and compositions, foaming control, etc., that cannot be reliably obtained from crucible melts

  5. Redox Control For Hanford HLW Feeds VSL-12R2530-1, REV 0

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Kot, Wing K. [The Catholic University of America, Washington, DC (United States); Joseph, Innocent [The Catholic University of America, Washington, DC (United States)

    2012-12-13

    The principal objectives of this work were to investigate the effects of processing simulated Hanford HLW at the estimated maximum concentrations of nitrates and oxalates and to identify strategies to mitigate any processing issues resulting from high concentrations of nitrates and oxalates. This report provides results for a series of tests that were performed on the DM10 melter system with simulated C-106/AY-102 HLW. The tests employed simulated HLW feeds containing variable amounts of nitrates and waste organic compounds corresponding to maximum concentrations proj ected for Hanford HLW streams in order to determine their effects on glass production rate, processing characteristics, glass redox conditions, melt pool foaming, and the tendency to form secondary phases. Such melter tests provide information on key process factors such as feed processing behavior, dynamic effects during processing, processing rates, off-gas amounts and compositions, foaming control, etc., that cannot be reliably obtained from crucible melts.

  6. Safety assessment of HLW geological disposal system

    Naito, Morimasa

    2006-01-01

    In accordance with the Japanese nuclear program, the liquid waste with a high level of radioactivity arising from reprocessing is solidified in a stable glass matrix (vitrification) in stainless steel fabrication containers. The vitrified waste is referred to as high-level radioactive waste (HLW), and is characterized by very high initial radioactivity which, even though it decreases with time, presents a potential long-term risk. It is therefore necessary to thoroughly manage HLW from human and his environment. After vitrification, HLW is stored for a period of 30 to 50 years to allow cooling, and finally disposed of in a stable geological environment at depths greater than 300 m below surface. The deep underground environment, in general, is considered to be stable over geological timescales compared with surface environment. By selecting an appropriate disposal site, therefore, it is considered to be feasible to isolate the waste in the repository from man and his environment until such time as radioactivity levels have decayed to insignificance. The concept of geological disposal in Japan is similar to that in other countries, being based on a multibarrier system which combines the natural geological environment with engineered barriers. It should be noted that geological disposal concept is based on a passive safety system that does not require any institutional control for assuring long term environmental safety. To demonstrate feasibility of safe HLW repository concept in Japan, following technical steps are essential. Selection of a geological environment which is sufficiently stable for disposal (site selection). Design and installation of the engineered barrier system in a stable geological environment (engineering measures). Confirmation of the safety of the constructed geological disposal system (safety assessment). For site selection, particular consideration is given to the long-term stability of the geological environment taking into account the fact

  7. Melter Throughput Enhancements for High-Iron HLW

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Gan, Hoa [The Catholic University of America, Washington, DC (United States); Joseph, Innocent [The Catholic University of America, Washington, DC (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States); Chaudhuri, Malabika [The Catholic University of America, Washington, DC (United States); Kot, Wing [The Catholic University of America, Washington, DC (United States)

    2012-12-26

    This report describes work performed to develop and test new glass and feed formulations in order to increase glass melting rates in high waste loading glass formulations for HLW with high concentrations of iron. Testing was designed to identify glass and melter feed formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts to assess melt rate using a vertical gradient furnace system and to develop new formulations with enhanced melt rate. Testing evaluated the effects of waste loading on glass properties and the maximum waste loading that can be achieved. The results from crucible-scale testing supported subsequent DuraMelter 100 (DM100) tests designed to examine the effects of enhanced glass and feed formulations on waste processing rate and product quality. The DM100 was selected as the platform for these tests due to its extensive previous use in processing rate determination for various HLW streams and glass compositions.

  8. Mechanical, degradation and cytocompatibility properties of magnesium coated phosphate glass fibre reinforced polycaprolactone composites.

    Liu, Xiaoling; Hasan, Muhammad S; Grant, David M; Harper, Lee T; Parsons, Andrew J; Palmer, Graham; Rudd, Chris D; Ahmed, Ifty

    2014-11-01

    Retention of mechanical properties of phosphate glass fibre reinforced degradable polyesters such as polycaprolactone and polylactic acid in aqueous media has been shown to be strongly influenced by the integrity of the fibre/polymer interface. A previous study utilising 'single fibre' fragmentation tests found that coating with magnesium improved the fibre and matrix interfacial shear strength. Therefore, the aim of this study was to investigate the effects of a magnesium coating on the manufacture and characterisation of a random chopped fibre reinforced polycaprolactone composite. Short chopped strand non-woven phosphate glass fibre mats were sputter coated with degradable magnesium to manufacture phosphate glass fibre/polycaprolactone composites. The degradation behaviour (water uptake, mass loss and pH change of the media) of these polycaprolactone composites as well as of pure polycaprolactone was investigated in phosphate buffered saline. The Mg coated fibre reinforced composites revealed less water uptake and mass loss during degradation compared to the non-coated composites. The cations released were also explored and a lower ion release profile for all three cations investigated (namely Na(+), Mg(2+) and Ca(2+)) was seen for the Mg coated composite samples. An increase of 17% in tensile strength and 47% in tensile modulus was obtained for the Mg coated composite samples. Both flexural and tensile properties were investigated and a higher retention of mechanical properties was obtained for the Mg coated fibre reinforced composite samples up to 10 days immersion in PBS. Cytocompatibility study showed both composite samples (coated and non-coated) had good cytocompatibility with human osteosarcoma cell line. © The Author(s) 2014 Reprints and permissions: sagepub.co.uk/journalsPermissions.nav.

  9. Analysis and interpretation of the performance degradation of glass Resistive Plate Chambers operated in streamer mode

    Calcaterra, A; Patteri, P; Piccolo, M; Della Mea, G; Restello, S; Ferri, F; Musella, P; Redaelli, N; Tabarelli de Fatis, T; Tinti, G; Mannocchi, G; Trinchero, G

    2007-01-01

    The long-term stability of Resistive Plate Chambers (RPCs) with glass electrodes was studied for one year with a dedicated test station hosting about 10 m2 of detectors. RPCs were operated in streamer mode with a ternary gas mixture containing argon (27%), isobutane (9%) and tetrafluoroethane (64%). Environmental conditions were kept under control and, in particular, the water pollution in the gas, deemed responsible for the degradation of glass RPC performance, was monitored never to exceed 30 ppm in the exhaust line. Evidence for a substantial aging of the detectors was observed, resulting in a loss of efficiency correlated to an increased rate of spurious streamers. This can be ascribed to the chemical attack of the glass surface by hydrofluoric acid formed in the streamer process, as confirmed by detailed morphological and chemical analyses of the electrode surface. Our results strengthen the indication that the instability of glass RPCs in the long term is related to the use of fluorocarbons as quenching...

  10. HLW Tank Space Management, Final Report

    Sessions, J.

    1999-01-01

    The HLW Tank Space Management Team (SM Team) was chartered to select and recommend an HLW Tank Space Management Strategy (Strategy) for the HLW Management Division of Westinghouse Savannah River Co. (WSRC) until an alternative salt disposition process is operational. Because the alternative salt disposition process will not be available to remove soluble radionuclides in HLW until 2009, the selected Strategy must assure that it safely receives and stores HLW at least until 2009 while continuing to supply sludge slurry to the DWPF vitrification process

  11. Effects of PV Module Soiling on Glass Surface Resistance and Potential-Induced Degradation: Preprint

    Hacke, Peter; Burton, Patrick; Hendrickson, Alex; Spartaru, Sergiu; Glick, Stephen; Terwilliger, Kent

    2015-12-03

    The sheet resistance of three soil types (Arizona road dust, soot, and sea salt) on glass were measured by the transmission line method as a function of relative humidity (RH) between 39% and 95% at 60 degrees C. Sea salt yielded a 3.5 order of magnitude decrease in resistance on the glass surface when the RH was increased over this RH range. Arizona road dust showed reduced sheet resistance at lower RH, but with less humidity sensitivity over the range tested. The soot sample did not show significant resistivity change compared to the unsoiled control. Photovoltaic modules with sea salt on their faces were step-stressed between 25% and 95% RH at 60 degrees C applying -1000 V bias to the active cell circuit. Leakage current from the cell circuit to ground ranged between two and ten times higher than that of the unsoiled controls. Degradation rate of modules with salt on the surface increased with increasing RH and time.

  12. Degradation of glass-fiber reinforced plastics by low temperature irradiation

    Nishijima, S.; Nishiura, T.; Ueno, S.; Tsukazaki, Y.; Okada, T.; Okada, T.M.; Miyata, K.; Kodaka, H.

    1998-01-01

    Low-temperature irradiation effects of glass-fiber reinforced plastics (GFRP) have been investigated in terms of mechanical properties such as interlaminar shear strength and creep, in order to obtain the selection standard of insulating materials of superconducting magnets used for fusion reactor. It was revealed that the degradation of interlaminar shear strength was strongly dependent of characteristics of matrix and/or glass/epoxy interface. Especially, the research has been carried out towards the creep behaviour of epoxy which is the matrix of GFRP, by both experimental and simulation method. It was suggested that the synergistic effects was observed in creep test. From the molecular dynamics simulation it was found that the cage effects was the one of the main reason of the stress effects of creep behavior under irradiation. (author)

  13. RECENT PROCESS IMPROVEMENTS TO INCREASE HLW THROUGHPUT AT THE DWPF

    Herman, C

    2007-01-01

    The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF), the world's largest operating high level waste (HLW) vitrification plant, began stabilizing about 35 million gallons of SRS liquid radioactive waste by-product in 1996. The DWPF has since filled over 2000 canisters with about 4000 pounds of radioactive glass in each canister. In the past few years there have been several process and equipment improvements at the DWPF to increase the rate at which the waste can be stabilized. These improvements have either directly increased waste processing rates or have desensitized the process and therefore minimized process upsets and thus downtime. These improvements, which include glass former optimization, increased waste loading of the glass, the melter heated bellows liner, and glass surge protection software, will be discussed in this paper

  14. The Influence of Na and Ti on the In Vitro Degradation and Bioactivity in 58S Sol-Gel Bioactive Glass

    Shirong Ni

    2012-01-01

    Full Text Available The aim of this study was to investigate the effect of Na and Ti on the in vitro degradation and bioactivity in the 58S bioactive glass. The degradation was evaluated through the activation energy of Si ion release from bioactive glasses and the weight loss of bioactive glasses in Tris-HCl buffer solution. The in vitro bioactivity of the bioactive glasses was investigated by analysis of apatite-formation ability in the simulated body fluid (SBF. The results showed that Na in the 58S glass accelerated the dissolution rate of the glass, whereas Ti in the 58S glass slowed down the rate of glass solubility. Bioactivity tests showed that Na in glass increased the apatite-forming ability in SBF. In contrast, Ti in glass retards the apatite formation at the initial stage of SBF soaking but does not affect the growth of apatite after long periods of soaking.

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

    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

  16. In vitro degradation of chitosan composite foams for biomedical applications and effect of bioactive glass as a crosslinker

    Martins Talita; Moreira Cheisy D. F.; Costa-Júnior Ezequiel S.; Pereira Marivalda M.

    2018-01-01

    In tissue engineering applications, 3D scaffolds with adequate structure and composition are required to provide durability that is compatiblewith the regeneration of native tissue. In the present study, the degradation of novel flexible 3D composite foams of chitosan (CH) combined with bioactive glass (BG)was evaluated, focusing on the role of BG as a physical crosslinker in the composites, and its effect on the degradation process. Highly porous CH/BG composite foams were obtained, and an e...

  17. Silicate glasses

    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

  18. In Vitro Evaluation the Influence of Glass-Ceramic Degradation Products on Osteoblast Cells.

    Israa K. Sabree

    2016-03-01

    Full Text Available Regenerative medicine focuses on using biomaterials as three-dimensional (3D porous scaffolds, specifically designed to mimic the nature of host tissue and hence to promote cell growth and tissue regeneration. 3D bioactive glass-ceramic scaffolds are one of the most frequently studied types of scaffolds for bone tissue engineering because of their excellent bioactivity and potential for stimulating osteogenesis and angiogenesis. For such purposes, porous 3D 70%SiO2-30%CaO bioactive glass-ceramic scaffolds with three different pore sizes and identical porosity are used in present study to investigate In vitro, the effect of pore size on the degradation rate of scaffold which is achieved through examining changes in the composition of the immersion solution(SBF, simulated body fluid, and to investigate the action of released ions from the bioactive glass-ceramic scaffold during soaking process on osteoblast cells The results confirmed that all three scaffolds behaved in a similar manner and the ions release from the three scaffolds were of comparable concentration, which may be attributable to the identical porosity for all the scaffolds in addition to the using static immersion which delays ions diffusion. The pH of culture media increased from 7.6 to 8.2 after one day soaking. The optical microscopy images demonstrated that high ion concentration (Si, Ca, P in the culture medium could have a negative effect on the cells and induce cell death, while low concentration of ionic dissolution products induces osteoblast proliferation in dilute culture medium.

  19. Degradation of basalt fibre and glass fibre/epoxy resin composites in seawater

    Wei Bin; Cao Hailin; Song Shenhua

    2011-01-01

    Research highlights: → BFRP degradation process in seawater environment was first investigated. → The mass gain change includes two effects: absorption and extraction. → The interfacial adhesion of BFRP is bigger than GFRP. → After treated, the bending strength of BFRP is lower than GFRP. → Reducing the Fe 2+ in the basalt fibre could lead to a higher stability of BFRP. - Abstract: Epoxy resins reinforced, respectively, by basalt fibres and glass fibres were treated with a seawater solution for different periods of time. Both the mass gain ratio and the strength maintenance ratio of the composites were examined after the treatment. The fracture surfaces were characterized using scanning electron microscopy. The tensile and bending strengths of the seawater treated samples showed a decreasing trend with treating time. In general, the anti-seawater corrosion property of the basalt fibre reinforced composites was almost the same as that of the glass fibre reinforced ones. Based on the experimental results, possible corrosion mechanisms were explored, indicating that an effective lowering of the Fe 2+ content in the basalt fibre could lead to a higher stability for the basalt fibre reinforced composites in a seawater environment.

  20. HLW Melter Control Strategy Without Visual Feedback VSL-12R2500-1 Rev 0

    Kruger, A A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Joseph, Innocent [The Catholic University of America, Washington, DC (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States); Callow, Richard A. [The Catholic University of America, Washington, DC (United States); Abramowitz, Howard [The Catholic University of America, Washington, DC (United States); Pegg, Ian L. [The Catholic University of America, Washington, DC (United States); Brandys, Marek [The Catholic University of America, Washington, DC (United States); Kot, Wing K. [The Catholic University of America, Washington, DC (United States)

    2012-11-13

    Plans for the treatment of high level waste (HL W) at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) are based upon the inventory of the tank wastes, the anticipated performance of the pretreatment processes, and current understanding of the capability of the borosilicate glass waste form [I]. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat and mass transfer and increase glass melting rates. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of ~ 1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HL W waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150°C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage.

  1. HLW Melter Control Strategy Without Visual Feedback VSL-12R2500-1 Rev 0

    Kruger, A A.; Joseph, Innocent; Matlack, Keith S.; Callow, Richard A.; Abramowitz, Howard; Pegg, Ian L.; Brandys, Marek; Kot, Wing K.

    2012-01-01

    Plans for the treatment of high level waste (HL W) at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) are based upon the inventory of the tank wastes, the anticipated performance of the pretreatment processes, and current understanding of the capability of the borosilicate glass waste form [I]. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat and mass transfer and increase glass melting rates. The WTP HLW melter has a glass surface area of 3.75 m 2 and depth of ∼ 1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HL W waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150°C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage

  2. 12 Flasktransport of vitrified High Level Waste (HLW)

    Verdier, A.; Lancelot, J. [COGEMA Logistics (AREVA Group) (France); Gisbertz, A.; Graf, W. [GNS (Germany); Bartagnon, O. [COGEMA (AREVA Group) (France)

    2004-07-01

    The return of HLW to Germany has started in 1996 with the first attribution of 28 glass canisters to German utilities by COGEMA. After several transports comprising 1, 2 and 6 flasks per shipment German and French Authorities requested to transport 12 flasks in a single shipment. The first of these 12-flask-transports was performed with the type CASTOR {sup registered} HAW 20/28 CG flask in 2002 and the second followed in 2003. COGEMA LOGISTICS is responsible for the overall transport assigned by GNS (Gesellschaft fuer Nuklear-Service mbH) being itself entrusted by the German utilities with the return of reprocessing residues.

  3. 12 Flasktransport of vitrified High Level Waste (HLW)

    Verdier, A.; Lancelot, J.; Gisbertz, A.; Graf, W.; Bartagnon, O.

    2004-01-01

    The return of HLW to Germany has started in 1996 with the first attribution of 28 glass canisters to German utilities by COGEMA. After several transports comprising 1, 2 and 6 flasks per shipment German and French Authorities requested to transport 12 flasks in a single shipment. The first of these 12-flask-transports was performed with the type CASTOR registered HAW 20/28 CG flask in 2002 and the second followed in 2003. COGEMA LOGISTICS is responsible for the overall transport assigned by GNS (Gesellschaft fuer Nuklear-Service mbH) being itself entrusted by the German utilities with the return of reprocessing residues

  4. Feasibility Study for Preparation and Use of Glass Grains as an Alternative to Glass Nodules for Vitrification of Nuclear Waste

    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)

  5. Feasibility Study for Preparation and Use of Glass Grains as an Alternative to Glass Nodules for Vitrification of Nuclear Waste

    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)

  6. INTEGRATED DM 1200 MELTER TESTING OF HLW C-106/AY-102 COMPOSITION USING BUBBLERS VSL-03R3800-1 REV 0 9/15/03

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of simulated HLW C-106/AY-102 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW C-106/AY-102 feed; determine the effect of bubbling rate on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post test inspections of system components.

  7. Integrated DM 1200 Melter Testing Of HLW C-106/AY-102 Composition Using Bubblers VSL-03R3800-1, Rev. 0, 9/15/03

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Bardakci, T.; Gong, W.; D'Angelo, N.A.; Pegg, I.L.

    2011-01-01

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of simulated HLW C-106/AY-102 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW C-106/AY-102 feed; determine the effect of bubbling rate on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post test inspections of system components.

  8. The In Vitro Bioactivity, Degradation, and Cytotoxicity of Polymer-Derived Wollastonite-Diopside Glass-Ceramics

    Amanda De Castro Juraski

    2017-04-01

    Full Text Available Ca-Mg silicates are receiving a growing interest in the field of bioceramics. In a previous study, wollastonite-diopside (WD glass-ceramics were successfully prepared by a new processing route, consisting of the heat treatment of a silicone resin embedding reactive oxide particles and a Ca/Mg-rich glass. The in vitro degradation, bioactivity, and cell response of these new WD glass-ceramics, fired at 900–1100 °C for 1 h, as a function of the Ca/Mg-rich glass content, are the aim of this investigation The results showed that WD glass-ceramics from formulations comprising different glass contents (70–100% at 900 °C, 30% at 1100 °C exhibit the formation of an apatite-like layer on their surface after immersion in SBF for seven days, thus confirming their surface bioactivity. The XRD results showed that these samples crystallized, mainly forming wollastonite (CaSiO3 and diopside (CaMgSi2O6, but combeite (Na2Ca2Si3O9 crystalline phase was also detected. Besides in vitro bioactivity, cytotoxicity and osteoblast adhesion and proliferation tests were applied after all characterizations, and the formulation comprising 70% glass was demonstrated to be promising for further in vivo studies.

  9. The In Vitro Bioactivity, Degradation, and Cytotoxicity of Polymer-Derived Wollastonite-Diopside Glass-Ceramics

    Juraski, Amanda De Castro; Dorion Rodas, Andrea Cecilia; Elsayed, Hamada; Bernardo, Enrico; Oliveira Soares, Viviane; Daguano, Juliana

    2017-01-01

    Ca-Mg silicates are receiving a growing interest in the field of bioceramics. In a previous study, wollastonite-diopside (WD) glass-ceramics were successfully prepared by a new processing route, consisting of the heat treatment of a silicone resin embedding reactive oxide particles and a Ca/Mg-rich glass. The in vitro degradation, bioactivity, and cell response of these new WD glass-ceramics, fired at 900–1100 °C for 1 h, as a function of the Ca/Mg-rich glass content, are the aim of this investigation The results showed that WD glass-ceramics from formulations comprising different glass contents (70–100% at 900 °C, 30% at 1100 °C) exhibit the formation of an apatite-like layer on their surface after immersion in SBF for seven days, thus confirming their surface bioactivity. The XRD results showed that these samples crystallized, mainly forming wollastonite (CaSiO3) and diopside (CaMgSi2O6), but combeite (Na2Ca2Si3O9) crystalline phase was also detected. Besides in vitro bioactivity, cytotoxicity and osteoblast adhesion and proliferation tests were applied after all characterizations, and the formulation comprising 70% glass was demonstrated to be promising for further in vivo studies. PMID:28772783

  10. In vitro degradation of chitosan composite foams for biomedical applications and effect of bioactive glass as a crosslinker

    Martins Talita

    2018-02-01

    Full Text Available In tissue engineering applications, 3D scaffolds with adequate structure and composition are required to provide durability that is compatiblewith the regeneration of native tissue. In the present study, the degradation of novel flexible 3D composite foams of chitosan (CH combined with bioactive glass (BGwas evaluated, focusing on the role of BG as a physical crosslinker in the composites, and its effect on the degradation process. Highly porous CH/BG composite foams were obtained, and an elevated degradation temperature and lower degradation rate compared with pure chitosan were observed, probably as a result of greater intermolecular interaction between CH and BG. The Fourier transform infrared spectroscopy (FTIR data suggest that hydrogen bonds were responsible for the physical crosslinking between CH and BG. The results confirm that CH/BG foams can combine controllable bioactivity and degradation behavior and, therefore, could be useful for tissue regeneration matrices.

  11. Waste Isolation Pilot Plant in situ experimental program for HLW

    Molecke, M.A.

    1977-01-01

    The Waste Isolation Pilot Plant (WIPP) will be a facility to demonstrate the environmental and operational safety of storing radioactive wastes in a deep geologic bedded salt facility. The WIPP will be located in southeastern New Mexico, approximately 30 miles east of the city of Carlsbad. The major focus of the pilot plant operation involves ERDA defense related low and intermediate-level transuranic wastes. The scope of the project also specifically includes experimentation utilizing commercially generated high-level wastes, or alternatively, spent unreprocessed fuel elements. WIPP HLW experiments are being conducted in an inter-related laboratory, bench-scale, and in situ mode. This presentation focuses on the planned in situ experiments which, depending on the availability of commercially reprocessed waste plus delays in the construction schedule of the WIPP, will begin in approximately 1985. Such experiments are necessary to validate preceding laboratory results and to provide actual, total conditions of geologic storage which cannot be adequately simulated. One set of planned experiments involves emplacing bare HLW fragments into direct contact with the bedded salt environment. A second set utilizes full-size canisters of waste emplaced in the salt in the same manner as planned for a future HLW repository. The bare waste experiments will study in an accelerated manner waste-salt bed-brine interactions including matrix integrity/degradation, brine leaching, system chemistry, and potential radionuclide migration through the salt bed. Utilization of full-size canisters of HLW in situ permits us to demonstrate operational effectiveness and safety. Experiments will evaluate corrosion and compatibility interactions between the waste matrix, canister and overpack materials, getter materials, stored energy, waste buoyancy, etc. Using full size canisters also allows us to demonstrate engineered retrievability of wastes, if necessary, at the end of experimentation

  12. Effect of magnesia on the degradability and bioactivity of sol-gel derived SiO2-CaO-MgO-P2O5 system glasses.

    Ma, J; Chen, C Z; Wang, D G; Jiao, Y; Shi, J Z

    2010-11-01

    Mesoporous 58SiO(2)-(38-x)CaO-xMgO-4P(2)O(5) glasses (where x=0, 5, 10 and 20 mol%) have been prepared by the sol-gel synthesis route. The effects of the substitution of MgO for CaO on glass degradation and bioactivity were studied in tris-(hydroxymethyl)-aminomethane and hydrochloric acid buffer solution (Tris-HCl) and simulated body fluid (SBF), respectively. It is observed that the synthesized glasses with various MgO contents possess the similar textural properties. The studies of in vitro degradability and bioactivity show that the rate of glass degradation gradually decreases with the increase of MgO and the formation of apatite layer on glass surface is retarded. The influences of the composition upon glass properties are explained in terms of their internal structures. Copyright (c) 2010 Elsevier B.V. All rights reserved.

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

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

    2015-09-01

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

  14. Glasses

    Dyre, Jeppe

    2004-01-01

    The temperature dependence of the viscosity of most glassforming liquids is known to depart significantly from the classical Arrhenius behaviour of simple fluids. The discovery of an unexpected correlation between the extent of this departure and the Poisson ratio of the resulting glass could lead...... to new understanding of glass ageing and viscous liquid dynamics....

  15. Support for HLW Direct Feed - Phase 2, VSL-15R3440-1

    Matlack, K. S. [The Catholic Univ. of America, Washington, DC (United States); Pegg, I. [The Catholic Univ. of America, Washington, DC (United States); Joseph, I. [EnergySolutions, Columbia, MD (United States); Kot, W. K. [The Catholic Univ. of America, Washington, DC (United States)

    2017-03-20

    This report describes work performed to develop and test new glass and feed formulations originating from a potential flow-sheet for the direct vitrification of High Level Waste (HLW) with minimal or no pretreatment. In the HLW direct feed option that is under consideration for early operations at the Hanford Tank Waste Treatment and Immobilization Plant (WTP), the pretreatment facility would be bypassed in order to support an earlier start-up of the vitrification facility. For HLW, this would mean that the ultrafiltration and caustic leaching operations that would otherwise have been performed in the pretreatment facility would either not be performed or would be replaced by an interim pretreatment function (in-tank leaching and settling, for example). These changes would likely affect glass formulations and waste loadings and have impacts on the downstream vitrification operations. Modification of the pretreatment process may result in: (i) Higher aluminum contents if caustic leaching is not performed; (ii) Higher chromium contents if oxidative leaching is not performed; (iii) A higher fraction of supernate in the HLW feed resulting from the lower efficiency of in-tank washing; and (iv) A higher water content due to the likely lower effectiveness of in-tank settling compared to ultrafiltration. The HLW direct feed option has also been proposed as a potential route for treating HLW streams that contain the highest concentrations of fast-settling plutoniumcontaining particles, thereby avoiding some of the potential issues associated with such particles in the WTP Pretreatment facility [1]. In response, the work presented herein focuses on the impacts of increased supernate and water content on wastes from one of the candidate source tanks for the direct feed option that is high in plutonium.

  16. Korean Reference HLW Disposal System

    Choi, Heui Joo; Lee, J. Y.; Kim, S. S. (and others)

    2008-03-15

    This report outlines the results related to the development of Korean Reference Disposal System for High-level radioactive wastes. The research has been supported around for 10 years through a long-term research plan by MOST. The reference disposal method was selected via the first stage of the research during which the technical guidelines for the geological disposal of HLW were determined too. At the second stage of the research, the conceptual design of the reference disposal system was made. For this purpose the characteristics of the reference spent fuels from PWR and CANDU reactors were specified, and the material and specifications of the canisters were determined in term of structural analysis and manufacturing capability in Korea. Also, the mechanical and chemical characteristics of the domestic Ca-bentonite were analyzed in order to supply the basic design parameters of the buffer. Based on these parameters the thermal and mechanical analysis of the near-field was carried out. Thermal-Hydraulic-Mechanical behavior of the disposal system was analyzed. The reference disposal system was proposed through the second year research. At the final third stage of the research, the Korean Reference disposal System including the engineered barrier, surface facilities, and underground facilities was proposed through the performance analysis of the disposal system.

  17. A dose of HLW reality

    Payne, J.

    1993-01-01

    What many people were sure they knew, and some others were fairly confident they knew, was acknowledged by the US Department of Energy in December: A monitored retrievable storage (MRS) facility will not be ready to accept spent fuel by January 31, 1998. A dose of reality has thus been added to the US high-level radioactive waste scene. Perhaps as important as the new reality is the practical, businesslike nature of the DOE's plan. The Department's proposal has the quality of a plan aimed at genuinely solving a problem rather just going through the motions. (In contrast, some readers are familiar with New York State's procedures for siting and licensing a low-level waste facility - procedures so labyrinthine that they are much more likely to protect political careers in that state than they are to achieve an LLW site). The DOE has received a lot of criticism - some justified, some not - about its handling of the HLW program. In this instance, it is proposing what many in the industry might have recommended: Make available storage capacity for spent nuclear fuel at existing federal government sites

  18. Effects of photovoltaic module soiling on glass surface resistance and potential-induced degradation

    Hacke, Peter; Burton, Patrick; Hendrickson, Alexander

    2015-01-01

    The sheet resistance of three soil types (Arizona road dust, soot, and sea salt) on glass were measured by the transmission line method as a function of relative humidity (RH) between 39% and 95% at 60°C. Sea salt yielded a 3.5 orders of magnitude decrease in resistance on the glass surface when ...

  19. MICROWAVE INDUCED DEGRADATION OF GLASS FIBER REINFORCED POLYESTER FOR FIBER AND RESIN RECOVERY

    Ucar, Hülya; Nielsen, Rudi Pankratz; Søgaard, Erik Gydesen

    A solvolysis process to depolymerize the resin in glass fiber reinforced composites and recover the glass fibers has been investigated using microwave induced irradiation. The depolymerization was carried out in HNO3 with concentrations in the range of 1M-7M and in KOH with concentrations ranging...

  20. Final Report - Testing of Optimized Bubbler Configuration for HLW Melter VSL-13R2950-1, Rev. 0, dated 6/12/2013

    Kruger, Albert A.; Pegg, I. L.; Callow, R. A.; Joseph, I.; Matlack, K. S.; Kot, W. K.

    2013-11-13

    The principal objective of this work was to determine the glass production rate increase and ancillary effects of adding more bubbler outlets to the current WTP HLW melter baseline. This was accomplished through testing on the HLW Pilot Melter (DM1200) at VSL. The DM1200 unit was selected for these tests since it was used previously with several HLW waste streams including the four tank wastes proposed for initial processing at Hanford. This melter system was also used for the development and optimization of the present baseline WTP HLW bubbler configuration for the WTP HLW melter, as well as for MACT testing for both HLW and LAW. Specific objectives of these tests were to: Conduct DM1200 melter testing with the baseline WTP bubbling configuration and as augmented with additional bubblers. Conduct DM1200 melter testing to differentiate the effects of total bubbler air flow and bubbler distribution on glass production rate and cold cap formation. Collect melter operating data including processing rate, temperatures at a variety of locations within the melter plenum space, melt pool temperature, glass melt density, and melter pressure with the baseline WTP bubbling configuration and as augmented with additional bubblers. Collect melter exhaust samples to compare particulate carryover for different bubbler configurations. Analyze all collected data to determine the effects of adding more bubblers to the WTP HLW melter to inform decisions regarding future lid re-designs. The work used a high aluminum HLW stream composition defined by ORP, for which an appropriate simulant and high waste loading glass formulation were developed and have been previously processed on the DM1200.

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

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

    2015-08-18

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

  2. Integrated HLW Conceptual Process Flowsheet(s) for the Crystalline Silicotitanate Process SRDF-98-04

    Jacobs, R.A.

    1998-01-01

    The Strategic Research and Development Fund (SRDF) provided funds to develop integrated conceptual flowsheets and material balances for a CST process as a potential replacement for, or second generation to, the ITP process. This task directly supports another SRDF task: Glass Form for HLW Sludge with CST, SRDF-98-01, by M. K. Andrews which seeks to further develop sludge/CST glasses that could be used if the ITP process were replaced by CST ion exchange. The objective of the proposal was to provide flowsheet support for development and evaluation of a High Level Waste Division process to replace ITP. The flowsheets would provide a conceptual integrated material balance showing the impact on the HLW division. The evaluation would incorporate information to be developed by Andrews and Harbour on CST/DWPF glass formulations and provide the bases for evaluating the economic impact of the proposed replacement process. Coincident with this study, the Salt Disposition Team began its evaluation of alternatives for disposition of the HLW salts in the SRS waste tanks. During that time, the CST IX process was selected as one of four alternatives (of eighteen Phase II alternatives) for further evaluation during Phase III

  3. Final Report Tests On The Duramelter 1200 HLW Pilot Melter System Using AZ-101 HLW Simulants VSL-02R0100-2, Rev. 1, 2/17/03

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Bardakci, T.; Gong, W.; D'Angelo, N.A.; Schatz, T.R.; Pegg, I.L.

    2011-01-01

    This document provides the final report on data and results obtained from a series of nine tests performed on the one-third scale DuraMelter(trademark) 1200 (DM1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part B1 (1). Both melters have similar melt surface areas (1.2 m 2 ) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plans. The nine tests reported here were preceded by an initial series of short-duration tests conducted to support the start-up and commissioning of this system. This report is a followup to the previously issued Preliminary Data Summary Reports. The DM1200 system was deployed for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. These tests include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The primary objective of the present series of tests was to determine the effects of a variety of parameters on the glass production rate in comparison to the RPP-WTP HL W design basis of 400 kg/m 2 /d. Previous testing on the DMIOOO system (1) concluded that achievement of that rate with simulants of projected WTP melter feeds (AZ-101 and C-106/AY-102) was unlikely without the use of bubblers. As part of those tests, the same feed that was used during the cold-commissioning of the West Valley Demonstration Project (WVDP) HLW vitrification system was run on the DM1000 system. The DM1000 tests reproduced the rates that were obtained at the larger

  4. FINAL REPORT TESTS ON THE DURAMELTER 1200 HLW PILOT MELTER SYSTEM USING AZ-101 HLW SIMULANTS VSL-02R0100-2 REV 1 2/17/03

    KRUGER AA; MATLACK KS; KOT WK; BARDAKCI T; GONG W; D' ANGELO NA; SCHATZ TR; PEGG IL

    2011-12-29

    This document provides the final report on data and results obtained from a series of nine tests performed on the one-third scale DuraMelter{trademark} 1200 (DM1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part B1 [1]. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plans. The nine tests reported here were preceded by an initial series of short-duration tests conducted to support the start-up and commissioning of this system. This report is a followup to the previously issued Preliminary Data Summary Reports. The DM1200 system was deployed for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. These tests include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The primary objective of the present series of tests was to determine the effects of a variety of parameters on the glass production rate in comparison to the RPP-WTP HL W design basis of 400 kg/m{sup 2}/d. Previous testing on the DMIOOO system [1] concluded that achievement of that rate with simulants of projected WTP melter feeds (AZ-101 and C-106/AY-102) was unlikely without the use of bubblers. As part of those tests, the same feed that was used during the cold-commissioning of the West Valley Demonstration Project (WVDP) HLW vitrification system was run on the DM1000 system. The DM1000 tests reproduced the rates that were obtained at the

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

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

    1997-01-01

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

  6. DM100 AND DM1200 MELTER TESTING WITH HIGH WASTE LOADING FORMULATIONS FOR HANFORD HIGH-ALUMINUM HLW STREAMS, TEST PLAN 09T1690-1

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Pegg, I.L.; Joseph, I.

    2009-01-01

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

  7. Monitoring chemical degradation of thermally cycled glass-fibre composites using hyperspectral imaging

    Papadakis, V.; Muller, B.; Hagenbeek, M.; Sinke, J.; Groves, R.M.; Yu, T.; Gyekenyesi, A.L.; Shull, P.J.; Wu, H.F.

    2016-01-01

    Nowadays, the application of glass-fibre composites in light-weight structures is growing. Although mechanical characterizations of those structures are commonly performed in testing, chemical changes of materials under stresses have not yet been well documented. In the present work coupon tests and

  8. Final Report - Effects of High Spinel and Chromium Oxide Crystal Contents on Simulated HLW Vitrification in DM100 Melter Tests, VSL-09R1520-1, Rev. 0, dated 6/22/09

    Kruger, Albert A.; Matlack, K. S.; Kot, W.; Pegg, I. L.; Chaudhuri, M.; Lutze, W.

    2013-11-13

    The principal objective of the work was to evaluate the effects of spinel and chromium oxide particles on WTP HLW melter operations and potential impacts on melter life. This was accomplished through a combination of crucible-scale tests, settling and rheological tests, and tests on the DM100 melter system. Crucible testing was designed to develop and identify HLW glass compositions with high waste loadings that exhibit formation of crystalline spinel and/or chromium oxide phases up to relatively high crystal contents (i.e., > 1 vol%). Characterization of crystal settling and the effects on melt rheology was performed on the HLW glass formulations. Appropriate candidate HLW glass formulations were selected, based on characterization results, to support subsequent melter tests. In the present work, crucible melts were formulated that exhibit up to about 4.4 vol% crystallization.

  9. Final Report - Management of High Sulfur HLW, VSL-13R2920-1, Rev. 0, dated 10/31/2013

    Kruger, Albert A.; Gan, H.; Pegg, I. L.; Feng, Z.; Gan, H; Joseph, I.; Matlack, K. S.

    2013-11-13

    The present report describes results from a series of small-scale crucible tests to determine the extent of corrosion associated with sulfur containing HLW glasses and to develop a glass composition for a sulfur-rich HLW waste stream, which was then subjected to small-scale melter testing to determine the maximum acceptable sulfate loadings. In the present work, a new glass formulation was developed and tested for a projected Hanford HLW composition with sulfate concentrations high enough to limit waste loading. Testing was then performed on the DM10 melter system at successively higher waste loadings to determine the maximum waste loading without the formation of a separate sulfate salt phase. Small scale corrosion testing was also conducted using the glass developed in the present work, the glass developed in the initial phase of this work [26], and a high iron composition, all at maximum sulfur concentrations determined from melter testing, in order to assess the extent of Inconel 690 and MA758 corrosion at elevated sulfate contents.

  10. A glass capillary based microfluidic electromembrane extraction of basic degradation products of nitrogen mustard and VX from water.

    Tak, Vijay; Kabra, Ankur; Pardasani, Deepak; Goud, D Raghavender; Jain, Rajeev; Dubey, D K

    2015-12-24

    In this work, a glass capillary based microfluidic electromembrane extraction (μ-EME) was demonstrated for the first time. The device was made by connecting an auxillary borosilicate glass tubing (O.D. 3mm, I.D. 2mm) perpendicular to main borosilicate glass capillary just below one end of the capillary (O.D. 8mm, I.D. 1.2mm). It generated the distorted T-shaped device with inlet '1' and inlet '2' for the introduction of sample and acceptor solutions, respectively. At one end of this device (inlet '2'), a microsyringe containing acceptor solution along with hollow fiber (O.D. 1000μm) was introduced. This configuration creates the micro-channel between inner wall of glass capillary and outer surface of hollow fiber. Sample solution was pumped into the system through another end of glass capillary (inlet '1'), with a micro-syringe pump. The sample was in direct contact with the supported liquid membrane (SLM), consisted of 20% (w/w) di-(2-ethylhexyl)phosphate in 2-nitrophenyl octyl ether immobilized in the pores of the hollow fiber. In the lumen of the hollow fiber, the acceptor phase was present. The driving force for extraction was direct current (DC) electrical potential sustained over the SLM. Highly polar (logP=-2.5 to 1.4) basic degradation products of nitrogen mustard and VX were selected as model analytes. The influence of chemical composition of SLM, extraction time, voltage and pH of donor and acceptor phase were investigated. The model analytes were extracted from 10μL of pure water with recoveries ranging from 15.7 to 99.7% just after 3min of operation time. Under optimized conditions, good limits of detection (2-50ngmL(-1)), linearity (from 5-1000 to 100-1000ngmL(-1)), and repeatability (RSDs below 11.9%, n=3) were achieved. Applicability of the proposed μ-EME was proved by recovering triethanolamine (31.3%) from 10μL of five times diluted original water sample provided by the Organization for the Prohibition of Chemical Weapons during 28th official

  11. Significant degradability enhancement in multilayer coating of polycaprolactone-bioactive glass/gelatin-bioactive glass on magnesium scaffold for tissue engineering applications

    Yazdimamaghani, Mostafa [School of Chemical Engineering, Oklahoma State University, Stillwater, OK 74078 (United States); School of Materials Science and Engineering, Helmerich Advanced Technology Research Center, Oklahoma State University, Tulsa, OK 74106 (United States); Razavi, Mehdi [School of Materials Science and Engineering, Helmerich Advanced Technology Research Center, Oklahoma State University, Tulsa, OK 74106 (United States); Vashaee, Daryoosh [Electrical and Computer Engineering Department, North Carolina State University, Raleigh, NC 27606 (United States); Pothineni, Venkata Raveendra [Biomaterials and Advanced Drug Delivery Laboratory, Stanford University, Palo Alto, CA 94305 (United States); Rajadas, Jayakumar [Biomaterials and Advanced Drug Delivery Laboratory, Stanford University, Palo Alto, CA 94305 (United States); Department of Neurology and Neurological Sciences, Stanford University School of Medicine, Palo Alto, CA 94305 (United States); Stanford Cardiovascular Institute, Stanford University School of Medicine, Palo Alto, CA 94305 (United States); Tayebi, Lobat, E-mail: lobat.tayebi@marquette.edu [School of Materials Science and Engineering, Helmerich Advanced Technology Research Center, Oklahoma State University, Tulsa, OK 74106 (United States); Biomaterials and Advanced Drug Delivery Laboratory, Stanford University, Palo Alto, CA 94305 (United States); Department of Developmental Sciences, Marquette University School of Dentistry, Milwaukee, WI 53233 (United States)

    2015-05-30

    Highlights: • PCL-BaG/Gel-BaG coating was applied on the surface of Mg scaffolds. • Mg scaffold/PCL-BaG/Gel-BaG presented improved biodegradation resistance. • Mg scaffold coated with the PCL-BaG layer indicated better bioactivity. - Abstract: Magnesium (Mg) is a promising candidate to be used in medical products especially as bone tissue engineering scaffolds. The main challenge for using Mg in biomedical applications is its high degradation rate in the body. For this reason, in this study, a multilayer polymeric layer composed of polycaprolactone (PCL) and gelatin (Gel) reinforced with bioactive glass (BaG) particles has been applied on the surface of Mg scaffolds. The materials characteristics of uncoated Mg scaffold, Mg scaffold coated only with PCL-BaG and Mg scaffold coated with PCL-BaG and Gel-BaG have been analyzed and compared in detail. Scanning electron microscope (SEM) equipped with energy dispersive spectroscopy (EDS), and Fourier transform infrared spectroscopy (FTIR) were utilized for microstructural studies. In vitro bioactivity and biodegradation evaluations were carried out by submerging the scaffolds in simulated body fluid (SBF) at pre-determined time points. The results demonstrated that Mg scaffold coated with PCL-BaG and Gel-BaG exhibited significant improvement in biodegradability.

  12. Significant degradability enhancement in multilayer coating of polycaprolactone-bioactive glass/gelatin-bioactive glass on magnesium scaffold for tissue engineering applications

    Yazdimamaghani, Mostafa; Razavi, Mehdi; Vashaee, Daryoosh; Pothineni, Venkata Raveendra; Rajadas, Jayakumar; Tayebi, Lobat

    2015-05-01

    Magnesium (Mg) is a promising candidate to be used in medical products especially as bone tissue engineering scaffolds. The main challenge for using Mg in biomedical applications is its high degradation rate in the body. For this reason, in this study, a multilayer polymeric layer composed of polycaprolactone (PCL) and gelatin (Gel) reinforced with bioactive glass (BaG) particles has been applied on the surface of Mg scaffolds. The materials characteristics of uncoated Mg scaffold, Mg scaffold coated only with PCL-BaG and Mg scaffold coated with PCL-BaG and Gel-BaG have been analyzed and compared in detail. Scanning electron microscope (SEM) equipped with energy dispersive spectroscopy (EDS), and Fourier transform infrared spectroscopy (FTIR) were utilized for microstructural studies. In vitro bioactivity and biodegradation evaluations were carried out by submerging the scaffolds in simulated body fluid (SBF) at pre-determined time points. The results demonstrated that Mg scaffold coated with PCL-BaG and Gel-BaG exhibited significant improvement in biodegradability.

  13. Significant degradability enhancement in multilayer coating of polycaprolactone-bioactive glass/gelatin-bioactive glass on magnesium scaffold for tissue engineering applications

    Yazdimamaghani, Mostafa; Razavi, Mehdi; Vashaee, Daryoosh; Pothineni, Venkata Raveendra; Rajadas, Jayakumar; Tayebi, Lobat

    2015-01-01

    Highlights: • PCL-BaG/Gel-BaG coating was applied on the surface of Mg scaffolds. • Mg scaffold/PCL-BaG/Gel-BaG presented improved biodegradation resistance. • Mg scaffold coated with the PCL-BaG layer indicated better bioactivity. - Abstract: Magnesium (Mg) is a promising candidate to be used in medical products especially as bone tissue engineering scaffolds. The main challenge for using Mg in biomedical applications is its high degradation rate in the body. For this reason, in this study, a multilayer polymeric layer composed of polycaprolactone (PCL) and gelatin (Gel) reinforced with bioactive glass (BaG) particles has been applied on the surface of Mg scaffolds. The materials characteristics of uncoated Mg scaffold, Mg scaffold coated only with PCL-BaG and Mg scaffold coated with PCL-BaG and Gel-BaG have been analyzed and compared in detail. Scanning electron microscope (SEM) equipped with energy dispersive spectroscopy (EDS), and Fourier transform infrared spectroscopy (FTIR) were utilized for microstructural studies. In vitro bioactivity and biodegradation evaluations were carried out by submerging the scaffolds in simulated body fluid (SBF) at pre-determined time points. The results demonstrated that Mg scaffold coated with PCL-BaG and Gel-BaG exhibited significant improvement in biodegradability

  14. Assessment of degradation concerns for spent fuel, high-level wastes, and transuranic wastes in monitored retrievalbe storage

    Guenther, R.J.; Gilbert, E.R.; Slate, S.C.; Partain, W.L.; Divine, J.R.; Kreid, D.K.

    1984-01-01

    It has been concluded that there are no significant degradation mechanisms that could prevent the design, construction, and safe operation of monitored retrievable storage (MRS) facilities. However, there are some long-term degradation mechanisms that could affect the ability to maintain or readily retrieve spent fuel (SF), high-level wastes (HLW), and transuranic wastes (TRUW) several decades after emplacement. Although catastrophic failures are not anticipated, long-term degradation mechanisms have been identified that could, under certain conditions, cause failure of the SF cladding and/or failure of TRUW storage containers. Stress rupture limits for Zircaloy-clad SF in MRS range from 300 to 440 0 C, based on limited data. Additional tests on irradiated Zircaloy (3- to 5-year duration) are needed to narrow this uncertainty. Cladding defect sizes could increase in air as a result of fuel density decreases due to oxidation. Oxidation tests (3- to 5-year duration) on SF are also needed to verify oxidation rates in air and to determine temperatures below which monitoring of an inert cover gas would not be required. Few, if any, changes in the physical state of HLW glass or canisters or their performance would occur under projected MRS conditions. The major uncertainty for HLW is in the heat transfer through cracked glass and glass devitrification above 500 0 C. Additional study of TRUW is required. Some fraction of present TRUW containers would probably fail within the first 100 years of MRS, and some TRUW would be highly degraded upon retrieval, even in unfailed containers. One possible solution is the design of a 100-year container. 93 references, 28 figures, 17 tables

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

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

    2015-08-01

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

  16. Optical Degradation of Colloidal Eu-Complex Embedded in Silica Glass Film Using Reprecipitation and Sol-Gel Methods.

    Fukuda, Takeshi; Kurabayashi, Tomokazu; Yamaki, Tatsuki

    2016-04-01

    A reprecipitation method has been investigated for fabricating colloidal nanoparticles using Eu-complex. Herein, we investigated optical degradation characteristics of (1,10-phenanthroline)tris [4,4,4-trifluoro-1-(2-thienyl)-1,3-butanedionato]europium(III) colloidal nanoparticles, which were embedded into a silica glass film fabricated by a conventional sol-gel process. At first, we tried several types of good solvents for the reprecipitation method, and dimethyl sulfoxide (DMSO) is found to be a suitable solvent for realizing the small diameter and the high long-term stability against the ultraviolet irradiation even though the boing point of DMSO is higher than that of water used as a poor solvent. By optimizing the good solvent and the concentration of Eu-complex, the relative photoluminescence intensity of 0.96 was achieved even though the ultraviolet light was continuously irradiated for 90 min. In addition, the average diameter of 106 nm was achieved when DMSO was used as a good solvent, resulting in the high transmittance at a visible wavelength region. Therefore, we can achieve the transparent emissive thin film with a center wavelength of 612 nm, and the optical degradation was drastically reduced by forming nanoparticles.

  17. Degradation of Zr-based bulk metallic glasses used in load-bearing implants: A tribocorrosion appraisal.

    Zhao, Guo-Hua; Aune, Ragnhild E; Mao, Huahai; Espallargas, Nuria

    2016-07-01

    Owing to the amorphous structure, Bulk Metallic Glasses (BMGs) have been demonstrating attractive properties for potential biomedical applications. In the present work, the degradation mechanisms of Zr-based BMGs with nominal compositions Zr55Cu30Ni5Al10 and Zr65Cu18Ni7Al10 as potential load-bearing implant material were investigated in a tribocorrosion environment. The composition-dependent micro-mechanical and tribological properties of the two BMGs were evaluated prior to the tribocorrosion tests. The sample Zr65-BMG with a higher Zr content exhibited increased plasticity but relatively reduced wear resistance during the ball-on-disc tests. Both BMGs experienced abrasive wear after the dry wear test under the load of 2N. The cross-sectional subsurface structure of the wear track was examined by Focused Ion Beam (FIB). The electrochemical properties of the BMGs in simulated body fluid were evaluated by means of potentiodynamic polarization and X-ray Photoelectron Spectroscopy (XPS). The spontaneous passivation of Zr-based BMGs in Phosphate Buffer Saline solution was mainly attributed to the highly concentrated zirconium cation (Zr(4+)) in the passive film. The tribocorrosion performance of the BMGs was investigated using a reciprocating tribometer equipped with an electrochemical cell. The more passive nature of the Zr65-BMG had consequently a negative influence on its tribocorrosion resistance, which induced the wear-accelerated corrosion and eventually speeded-up the degradation process. It has been revealed the galvanic coupling was established between the depassivated wear track and the surrounding passive area, which is the main degradation mechanism for the passive Zr65-BMG subjected to the tribocorrosion environment. Copyright © 2015 Elsevier Ltd. All rights reserved.

  18. FINAL REPORT DM1200 TESTS WITH AZ 101 HLW SIMULANTS VSL-03R3800-4 REV 0 2/17/04

    KRUGER AA; MATLACK KS; BARDAKCI T; D' ANGELO NA; GONG W; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM 1200 HLW Pilot Melter during processing of simulated HLW AZ-101 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW AZ-101 feed; determine the effect of bubbling rate and feed solids content on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post-test inspections of system components. The test objectives (including test success criteria), along with how they were met, are outlined in a table.

  19. Final Report DM1200 Tests With AZ 101 HLW Simulants VSL-03R3800-4, Rev. 0, 2/17/04

    Kruger, A.A.; Matlack, K.S.; Bardakci, T.; D'Angelo, N.A.; Gong, W.; Kot, W.K.; Pegg, I.L.

    2011-01-01

    This report documents melter and off-gas performance results obtained on the DM 1200 HLW Pilot Melter during processing of simulated HLW AZ-101 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW AZ-101 feed; determine the effect of bubbling rate and feed solids content on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post-test inspections of system components. The test objectives (including test success criteria), along with how they were met, are outlined in a table.

  20. Geochemistry Model Validation Report: Material Degradation and Release Model

    H. Stockman

    2001-09-28

    The purpose of this Analysis and Modeling Report (AMR) is to validate the Material Degradation and Release (MDR) model that predicts degradation and release of radionuclides from a degrading waste package (WP) in the potential monitored geologic repository at Yucca Mountain. This AMR is prepared according to ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 17). The intended use of the MDR model is to estimate the long-term geochemical behavior of waste packages (WPs) containing U. S . Department of Energy (DOE) Spent Nuclear Fuel (SNF) codisposed with High Level Waste (HLW) glass, commercial SNF, and Immobilized Plutonium Ceramic (Pu-ceramic) codisposed with HLW glass. The model is intended to predict (1) the extent to which criticality control material, such as gadolinium (Gd), will remain in the WP after corrosion of the initial WP, (2) the extent to which fissile Pu and uranium (U) will be carried out of the degraded WP by infiltrating water, and (3) the chemical composition and amounts of minerals and other solids left in the WP. The results of the model are intended for use in criticality calculations. The scope of the model validation report is to (1) describe the MDR model, and (2) compare the modeling results with experimental studies. A test case based on a degrading Pu-ceramic WP is provided to help explain the model. This model does not directly feed the assessment of system performance. The output from this model is used by several other models, such as the configuration generator, criticality, and criticality consequence models, prior to the evaluation of system performance. This document has been prepared according to AP-3.10Q, ''Analyses and Models'' (Ref. 2), and prepared in accordance with the technical work plan (Ref. 17).

  1. Geochemistry Model Validation Report: Material Degradation and Release Model

    Stockman, H.

    2001-01-01

    The purpose of this Analysis and Modeling Report (AMR) is to validate the Material Degradation and Release (MDR) model that predicts degradation and release of radionuclides from a degrading waste package (WP) in the potential monitored geologic repository at Yucca Mountain. This AMR is prepared according to ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 17). The intended use of the MDR model is to estimate the long-term geochemical behavior of waste packages (WPs) containing U. S . Department of Energy (DOE) Spent Nuclear Fuel (SNF) codisposed with High Level Waste (HLW) glass, commercial SNF, and Immobilized Plutonium Ceramic (Pu-ceramic) codisposed with HLW glass. The model is intended to predict (1) the extent to which criticality control material, such as gadolinium (Gd), will remain in the WP after corrosion of the initial WP, (2) the extent to which fissile Pu and uranium (U) will be carried out of the degraded WP by infiltrating water, and (3) the chemical composition and amounts of minerals and other solids left in the WP. The results of the model are intended for use in criticality calculations. The scope of the model validation report is to (1) describe the MDR model, and (2) compare the modeling results with experimental studies. A test case based on a degrading Pu-ceramic WP is provided to help explain the model. This model does not directly feed the assessment of system performance. The output from this model is used by several other models, such as the configuration generator, criticality, and criticality consequence models, prior to the evaluation of system performance. This document has been prepared according to AP-3.10Q, ''Analyses and Models'' (Ref. 2), and prepared in accordance with the technical work plan (Ref. 17)

  2. Molecular glasses for nuclear waste encapsulation

    Ropp, R.C.

    1982-01-01

    The use of a molecular glass based upon a polymerized phosphate of aluminum (PAP), indium or gallium overcomes all of the prior objections to use of glass as a high-level nuclear waste (HLW) encapsulation agent. This HLW glass product could not be made to devitrify, dissolved all of the oxides found in calcine, including the difficultly soluble ones, did not form microcrystallites in the melt or subsequent glass-casting, and possessed a hydrolytic etching rate to boiling water even lower than that of HLW-ZBS glass. A precursor compound, M(H 2 PO 4 ) 3 , is prepared, where M is a trivalent metal selected from the group consisting of aluminum, indium and gallium. The impurity level is carefully controlled so as not to exceed 300 ppm total. The precursor crystals may be washed to remove excess phosphoric acid as desired. HLW is added to the crystals and the mixture is then heated at a controlled heating rate to induce solid state polymerization and to form a melt at 1350 degrees C in which the HLW oxides dissolve rapidly

  3. Counter current decantation washing of HLW sludge

    Brooke, J.N.; Peterson, R.A.

    1997-01-01

    The Savannah River Site (SRS) has 51 High Level Waste (HLW) tanks with typical dimensions 25.9 meters (85 feet) diameter and 10 meters (33 feet) high. Nearly 114 million liters (30 M gallons) of HLW waste is stored in these tanks in the form of insoluble solids called sludge, crystallized salt called salt cake, and salt solutions. This waste is being converted to waste forms stable for long term storage. In one of the processes, soluble salts are washed from HLW sludge in preparation for vitrification. At present, sludge is batch washed in a waste tank with one or no reuse of the wash water. Sodium hydroxide and sodium nitrite are added to the wash water for tank corrosion protection; the large volumes of spent wash water are recycled to the evaporator system; additional salt cake is produced; and sodium carbonate is formed in the washed sludge during storage by reaction with CO 2 from the air. High costs and operational concerns with the current washing process prompts DOE and WSRC to seek an improved washing method. A new method should take full advantage of the physical/chemical properties of sludge, experience from other technical disciplines, processing rate requirements, inherent process safety, and use of proven processes and equipment. Counter current solids washing is a common process in the minerals processing and chemical industries. Washing circuits can be designed using thickeners, filters or centrifuges. Realizing the special needs of nuclear work and the low processing rates required, a Counter Current Decantation (CCD) circuit is proposed using small thickeners and fluidic pumps

  4. Strategic management of HLW repository projects

    Bartlett, J.W.

    1984-01-01

    This paper suggests an approach to strategic management of HLW repository projects based on the premise that a primary objective of project activities is resolution of issues. The approach would be implemented by establishing an issues management function with responsibility to define the issues agenda, develop and apply the tools for assessing progress toward issue resolution, and develop the issue resolution criteria. A principal merit of the approach is that it provides a defensible rationale for project plans and activities. It also helps avoid unnecessary costs and schedule delays, and it helps assure coordination between project functions that share responsibilities for issue resolution

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

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

    1996-04-01

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

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

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

    2009-10-05

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

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

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

    2009-01-01

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

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

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

    1997-07-01

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

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

    Weber, W.J.

    1997-01-01

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

  10. Grouping in partitioning of HLW for burning and/or transmutation with nuclear reactors

    Kitamoto, Asashi; Mulyanto.

    1995-01-01

    A basic concept on partitioning and transmutation treatment by neutron reaction was developed in order to improve the waste management and the disposal scenario of high level waste (HLW). The grouping in partitioning was important factor and closely linked with the characteristics of B/T (burning and/or transmutation) treatment. The selecting and grouping concept in partitioning of HLW was proposed herein, such as Group MA1 (Np, Am, and unrecovered U and Pu), Group MA2 (Cm, Cf etc.), Group A (Tc and I), Group B (Cs and Sr) and Group R (the partitioned remain of HLW), judging from the three criteria for B/T treatment proposed in this study, which is related to (1) the value of hazard index for long-term tendency based on ALI, (2) the relative dose factor related to the mobility or retardation in ground water penetrated through geologic layer, and (3) burning and/or transmutation characteristics for recycle B/T treatment and the decay acceleration ratio by neutron reaction. Group MA1 and Group A could be burned effectively by thermal B/T reactor. Group MA2 could be burned effectively by fast B/T reactor. Transmutation of Group B by neutron reaction is difficult, therefore the development of radiation application of Group B (Cs and Sr) in industrial scale may be an interesting option in the future. Group R, i.e. the partitioned remains of HLW, and also a part of Group B should be immobilized and solidified by the glass matrix. HI ALI , the hazard index based on ALI, due to radiotoxicity of Group R can be lower than HI ALI due to standard mill tailing (smt) or uranium ore after about 300 years. (author)

  11. X-ray absorption spectroscopy and high-energy XRD study of the local environment of copper in antibacterial copper-releasing degradable phosphate glasses

    Pickup, David M.; Ahmed, Ifty; Fitzgerald, Victoria; Moss, Rob M.; Wetherall, Karen; Knowles, Jonathan C.; Smith, Mark E.; Newport, Robert J.

    2006-01-01

    Phosphate-based glasses of the general formula Na2O-CaO-P2O5 are degradable in an aqueous environment, and therefore can act as antibacterial materials through the inclusion of ions such as copper. In this study, CuO and Cu2O were added to Na2O-CaO-P2O5 glasses (1-20 mol% Cu) and X-ray absorption spectroscopy (XAS) and high-energy X-ray diffraction (HEXRD) used to probe the local environment of the copper ions. Copper K-edge X-ray absorption near-edge structure (XANES) spectra confirm the oxi...

  12. In-situ polymerisation of fully bioresorbable polycaprolactone/phosphate glass fibre composites: In vitro degradation and mechanical properties.

    Chen, Menghao; Parsons, Andrew J; Felfel, Reda M; Rudd, Christopher D; Irvine, Derek J; Ahmed, Ifty

    2016-06-01

    Fully bioresorbable composites have been investigated in order to replace metal implant plates used for hard tissue repair. Retention of the composite mechanical properties within a physiological environment has been shown to be significantly affected due to loss of the integrity of the fibre/matrix interface. This study investigated phosphate based glass fibre (PGF) reinforced polycaprolactone (PCL) composites with 20%, 35% and 50% fibre volume fractions (Vf) manufactured via an in-situ polymerisation (ISP) process and a conventional laminate stacking (LS) followed by compression moulding. Reinforcing efficiency between the LS and ISP manufacturing process was compared, and the ISP composites revealed significant improvements in mechanical properties when compared to LS composites. The degradation profiles and mechanical properties were monitored in phosphate buffered saline (PBS) at 37°C for 28 days. ISP composites revealed significantly less media uptake and mass loss (pproperties of ISP composites were substantially higher (p<0.0001) than those of the LS composites, which showed that the ISP manufacturing process provided a significantly enhanced reinforcement effect than the LS process. During the degradation study, statistically higher flexural property retention profiles were also seen for the ISP composites compared to LS composites. SEM micrographs of fracture surfaces for the LS composites revealed dry fibre bundles and poor fibre dispersion with polymer rich zones, which indicated poor interfacial bonding, distribution and adhesion. In contrast, evenly distributed fibres without dry fibre bundles or polymer rich zones, were clearly observed for the ISP composite samples, which showed that a superior fibre/matrix interface was achieved with highly improved adhesion. Copyright © 2015 Elsevier Ltd. All rights reserved.

  13. Testing the efficacy and the potential effect on indoor air quality of a transparent self-cleaning TiO{sub 2}-coated glass through the degradation of a fluoranthene layer

    Romeas, V.; Pichat, P.; Guillard, C.; Chopin, T.; Lehaut, C.

    1999-10-01

    Self-cleaning glass can be obtained by coating glass with a transparent, thin layer of TiO{sub 2} nanoparticles. To test the self-cleaning properties, fluoranthene--the most abundant polycyclic aromatic hydrocarbon in the atmospheric particulate matter--was sprayed over the glass. Under solar-like UV light, not only was fluoranthene removed at a rate of ca. 0.73 nmol/h per cm{sup 2} of glass but also all fluoranthene degradation products were, and thus the coated-glass transparency was recovered, which did not occur with noncoated glass. The fluoranthene percentage converted to volatile carbonyl products released into ambient air was lower with than without TiO{sub 2} coating; i.e., the self-cleaning glass could have a positive influence on indoor air quality. Mechanisms are discussed to account for the main primary products among the 40 fluoranthene photocatalytic degradation intermediate products which the authors identified.

  14. Surface degradation of glass ceramics after exposure to acidulated phosphate fluoride

    CCAHUANA, Vanessa Zulema S.; ÖZCAN, Mutlu; MESQUITA, Alfredo Mikail Melo; NISHIOKA, Renato Sussumo; KIMPARA, Estevão Tomomitsu; BOTTINO, Marco Antonio

    2010-01-01

    Objective This study evaluated the surface degradation effect of acidulated phosphate fluoride (APF) gel exposure on the glassy matrix ceramics as a function of time. Material and methods Disc-shaped ceramic specimens (N = 120, 10/per ceramic material) were prepared in stainless steel molds (inner diameter: 5 mm, height: 2 mm) using 6 dental ceramics: 3 indicated for ceramic-fused-to-metal (Vita Omega 900, Carmen and Vita Titankeramik), 2 for all-ceramic (Vitadur Alpha and Finesse® Low Fusing) and 1 for both types of restorations (IPS d.SIGN). The specimens were wet ground finished, ultrasonically cleaned and auto-glazed. All specimens were subjected to calculation of percentage of mass loss, surface roughness analysis and topographical description by scanning electron microscopy (SEM) before (0 min) and after exposure to 1.23 % APF gel for 4 min and 60 min representing short- and long-term etching effect, respectively. The data were analyzed using two-way ANOVA with repeated measures and Tukey`s test (α=0.05). Results Significant effect of the type of the ceramics (p=0.0000, p=0.0031) and exposure time (p=0.0000) was observed in both surface roughness and percentage of mass loss values, respectively. The interaction factor between both parameters was also significant for both parameters (p=0.0904, p=0.0258). Both 4 min (0.44±0.1 - 0.81±0.2 µm) and 60 min (0.66±0.1 - 1.04±0.3 µm) APF gel exposure created significantly more surface roughness for all groups when compared to the control groups (0.33±0.2 - 0.68±0.2 µm) (p0.05) but at 60 min exposure, IPS d.SIGN showed the highest percentage of mass loss (0.1151±0.11). The mean surface roughness for Vita Titankeramik (0.84±0.2 µm) and Finesse® Low Fusing (0.74.±0.2 µm) was significantly higher than those of the other ceramics (0.59±0.1 µm - 0.49±0.1 µm) and Vita Titankeramik (pcorrosive attack on all of ceramics at varying degrees. Conclusions The ceramics indicated for either metal-ceramic or all

  15. HLW Long-term Management Technology Development

    Choi, Jong Won; Kang, C. H.; Ko, Y. K.

    2010-02-01

    Permanent disposal of spent nuclear fuels from the power generation is considered to be the unique method for the conservation of human being and nature in the present and future. In spite of spent nuclear fuels produced from power generation, based on the recent trends on the gap between supply and demand of energy, the advance on energy price and reduction of carbon dioxide, nuclear energy is expected to play a role continuously in Korea. It means that a new concept of nuclear fuel cycle is needed to solve problems on spent nuclear fuels. The concept of the advanced nuclear fuel cycle including PYRO processing and SFR was presented at the 255th meeting of the Atomic Energy Commission. According to the concept of the advanced nuclear fuel cycle, actinides and long-term fissile nuclides may go out of existence in SFR. And then it is possible to dispose of short term decay wastes without a great risk bearing. Many efforts had been made to develop the KRS for the direct disposal of spent nuclear fuels in the representative geology of Korea. But in the case of the adoption of Advanced nuclear fuel cycle, the disposal of PYRO wastes should be considered. For this, we carried out the Safety Analysis on HLW Disposal Project with 5 sub-projects such as Development of HLW Disposal System, Radwaste Disposal Safety Analysis, Feasibility study on the deep repository condition, A study on the Nuclide Migration and Retardation Using Natural Barrier, and In-situ Study on the Performance of Engineered Barriers

  16. Generic Degraded Configuration Probability Analysis for the Codisposal Waste Package

    S.F.A. Deng; M. Saglam; L.J. Gratton

    2001-01-01

    In accordance with the technical work plan, ''Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages'' (CRWMS M and O 2000c), this Analysis/Model Report (AMR) is developed for the purpose of screening out degraded configurations for U.S. Department of Energy (DOE) spent nuclear fuel (SNF) types. It performs the degraded configuration parameter and probability evaluations of the overall methodology specified in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000, Section 3) to qualifying configurations. Degradation analyses are performed to assess realizable parameter ranges and physical regimes for configurations. Probability calculations are then performed for configurations characterized by k eff in excess of the Critical Limit (CL). The scope of this document is to develop a generic set of screening criteria or models to screen out degraded configurations having potential for exceeding a criticality limit. The developed screening criteria include arguments based on physical/chemical processes and probability calculations and apply to DOE SNF types when codisposed with the high-level waste (HLW) glass inside a waste package. The degradation takes place inside the waste package and is long after repository licensing has expired. The emphasis of this AMR is on degraded configuration screening and the probability analysis is one of the approaches used for screening. The intended use of the model is to apply the developed screening criteria to each DOE SNF type following the completion of the degraded mode criticality analysis internal to the waste package

  17. HLW disposal in Germany - R and D achievements and outlook

    Steininger, W.

    2006-01-01

    The paper gives a brief overview of the status of R and D on HLW disposal. Shortly addressed is the current nuclear policy. After describing the responsibilities regarding R and D for disposing of heat-generating high-level (HLW) waste (vitrified waste and spent fuel), selected projects are mentioned to illustrate the state of knowledge in disposing of waste in rock salt. Participation in international projects and programs is described to illustrate the value for the German concepts and ideas for HLW disposal in different rock types. Finally, a condensed outlook on future activities is given. (author)

  18. Final Report Integrated DM1200 Melter Testing Using AZ-102 And C-106/AY-102 HLW Simulants: HLW Simulant Verification VSL-05R5800-1, Rev. 0, 6/27/05

    Kruger, A.A.; Matlack, K.S.; Gong, W.; Bardakci, T.; D'Angelo, N.A.; Brandys, M.; Kot, W.K.; Pegg, I.L.

    2011-01-01

    The principal objectives of the DM1200 melter tests were to determine the effects of feed rheology, feed solid content, and bubbler configuration on glass production rate and off-gas system performance while processing the HLW AZ-101 and C-106/AY-102 feed compositions; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components, as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and post test inspections of system components. The specific objectives (including test success criteria) of this testing, along with how each objective was met, are outlined in a table. The data provided in this Final Report address the impacts of HLW melter feed rheology on melter throughput and validation of the simulated HLW melter feeds. The primary purpose of this testing is to further validate/verify the HLW melter simulants that have been used for previous melter testing and to support their continued use in developing melter and off-gas related processing information for the Project. The primary simulant property in question is rheology. Simulants and melter feeds used in all previous melter tests were produced by direct addition of chemicals; these feed tend to be less viscous than rheological the upper-bound feeds made from actual wastes. Data provided here compare melter processing for the melter feed used in all previous DM100 and DM1200 tests (nominal melter feed) with feed adjusted by the feed vendor (NOAH Technologies) to be more viscous, thereby simulating more closely the upperbounding feed produced from actual waste. This report provides results of tests that are described in the Test Plan for this work. The Test Plan is responsive to one of several test objectives covered in the WTP Test Specification for this work; consequently, only part of the scope described in the Test Specification was addressed in this particular Test Plan. For the purpose of

  19. FINAL REPORT INTEGRATED DM1200 MELTER TESTING USING AZ 102 AND C 106/AY-102 HLW SIMULANTS: HLW SIMULANT VERIFICATION VSL-05R5800-1 REV 0 6/27/05

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    The principal objectives of the DM1200 melter tests were to determine the effects of feed rheology, feed solid content, and bubbler configuration on glass production rate and off-gas system performance while processing the HLW AZ-101 and C-106/AY-102 feed compositions; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components, as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and post test inspections of system components. The specific objectives (including test success criteria) of this testing, along with how each objective was met, are outlined in a table. The data provided in this Final Report address the impacts of HLW melter feed rheology on melter throughput and validation of the simulated HLW melter feeds. The primary purpose of this testing is to further validate/verify the HLW melter simulants that have been used for previous melter testing and to support their continued use in developing melter and off-gas related processing information for the Project. The primary simulant property in question is rheology. Simulants and melter feeds used in all previous melter tests were produced by direct addition of chemicals; these feed tend to be less viscous than rheological the upper-bound feeds made from actual wastes. Data provided here compare melter processing for the melter feed used in all previous DM100 and DM1200 tests (nominal melter feed) with feed adjusted by the feed vendor (NOAH Technologies) to be more viscous, thereby simulating more closely the upperbounding feed produced from actual waste. This report provides results of tests that are described in the Test Plan for this work. The Test Plan is responsive to one of several test objectives covered in the WTP Test Specification for this work; consequently, only part of the scope described in the Test Specification was addressed in this particular Test Plan. For the purpose of

  20. The borosilicate glass for 'PAMELA'

    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.)

  1. In Vitro Degradation of Borosilicate Bioactive Glass and Poly(l-lactide-co-ε-caprolactone Composite Scaffolds

    Jenna Tainio

    2017-11-01

    Full Text Available Composite scaffolds were obtained by mixing various amounts (10, 30 and 50 weight % [wt %] of borosilicate bioactive glass and poly(l-lactide-co-ε-caprolactone (PLCL copolymer. The composites were foamed using supercritical CO2. An increase in the glass content led to a decrease in the pore size and density. In vitro dissolution/reaction test was performed in simulated body fluid. As a function of immersion time, the solution pH increased due to the glass dissolution. This was further supported by the increasing amount of Ca in the immersing solution with increasing immersion time and glass content. Furthermore, the change in scaffold mass was significantly greater with increasing the glass content in the scaffold. However, only the scaffolds containing 30 and 50 wt % of glasses exhibited significant hydroxyapatite (HA formation at 72 h of immersion. The compression strength of the samples was also measured. The Young’s modulus was similar for the 10 and 30 wt % glass-containing scaffolds whereas it increased to 90 MPa for the 50 wt % glass containing scaffold. Upon immersion up to 72 h, the Young’s modulus increased and then remained constant for longer immersion times. The scaffold prepared could have great potential for bone and cartilage regeneration.

  2. Antibacterial properties of copper iodide-doped glass ionomer-based materials and effect of copper iodide nanoparticles on collagen degradation.

    Renné, Walter G; Lindner, Amanda; Mennito, Anthony S; Agee, Kelli A; Pashley, David H; Willett, Daniel; Sentelle, David; Defee, Michael; Schmidt, Michael; Sabatini, Camila

    2017-01-01

    This study investigated the antibacterial properties and micro-hardness of polyacrylic acid (PAA)-coated copper iodide (CuI) nanoparticles incorporated into glass ionomer-based materials, and the effect of PAA-CuI on collagen degradation. PAA-CuI nanoparticles were incorporated into glass ionomer (GI), Ionofil Molar AC, and resin-modified glass ionomer (RMGI), Vitrebond, at 0.263 wt%. The antibacterial properties against Streptococcus mutans (n = 6/group) and surface micro-hardness (n = 5/group) were evaluated. Twenty dentin beams were completely demineralized in 10 wt% phosphoric acid and equally divided in two groups (n = 10/group) for incubation in simulated body fluid (SBF) or SBF containing 1 mg/ml PAA-CuI. The amount of dry mass loss and hydroxyproline (HYP) released were quantified. Kruskal-Wallis, Student's t test, two-way ANOVA, and Mann-Whitney were used to analyze the antibacterial, micro-hardness, dry mass, and HYP release data, respectively (p glass ionomer matrix yielded significant reduction (99.999 %) in the concentration of bacteria relative to the control groups. While micro-hardness values of PAA-CuI-doped GI were no different from its control, PAA-CuI-doped RMGI demonstrated significantly higher values than its control. A significant decrease in dry mass weight was shown only for the control beams (10.53 %, p = 0.04). Significantly less HYP was released from beams incubated in PAA-CuI relative to the control beams (p glass ionomer-based materials as they greatly enhance their antibacterial properties and reduce collagen degradation without an adverse effect on their mechanical properties. The use of copper-doped glass ionomer-based materials under composite restorations may contribute to an increased longevity of adhesive restorations, because of their enhanced antibacterial properties and reduced collagen degradation.

  3. SNF/HLW Transfer System Description Document

    W. Holt

    2005-01-01

    The purpose of this system description document (SDD) is to establish requirements that drive the design of the spent nuclear fuel (SNF)/high-level radioactive waste (HLW) transfer system and associated bases, which will allow the design effort to proceed to license application. This SDD will be revised at strategic points as the design matures. This SDD identifies the requirements and describes the system design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This SDD is an engineering tool for design control. Accordingly, the primary audience and users are design engineers. This SDD is part of an iterative design process. It leads the design process with regard to the flowdown of upper tier requirements onto the system. Knowledge of these requirements is essential in performing the design process. The SDD follows the design with regard to the description of the system. The description provided in this SDD reflects the current results of the design process

  4. Stress analysis of HLW containers. Compas project

    1989-01-01

    This document reports the work carried out for the Compas project which looked at the performance of various computer codes in a selected benchmark exercise. This exercise consisted of several analyses on simplified models which have features typical of HLW containers. These analyses comprise two groups; one related to thick walled, stressed shell overpacks, the other related to thin walled, supported shell overpacks with a lead filler. The first set of analyses looked at an elastic-plastic behaviour and large deformation of a cylinder representative of the main body of thick walled containers). The second set looked at creep behaviour of the lead filler, and the shape the base of thin walled containers will take up, after hundreds of years in the repository. On the thick walled analyses with the cylinder subject to an external pressure all the codes gave consistent results in the elastic region and there is good agreement in the yield pressures. Once in the plastic region there is more divergence in the results although a consistent trend is predicted. One of the analyses predicted a non-axisymmetric mode of deformation as would be expected in reality. Fewer results were received for the creep analysis, however the transient creep results showed consistency, and were bounded by the final-state results

  5. HLW Flexible jumper materials compatibility evaluation

    Skidmore, T. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-11-13

    H-Tank Farm Engineering tasked SRNL/Materials Science & Technology (MS&T) to evaluate the compatibility of Goodyear Viper® chemical transfer hose with HLW solutions. The hose is proposed as a flexible Safety Class jumper for up to six months service. SRNL/MS&T performed various tests to evaluate the effects of radiation, high pH chemistry and elevated temperature on the hose, particularly the inner liner. Test results suggest an upper dose limit of 50 Mrad for the hose. Room temperature burst pressure values at 50 Mrad are estimated at 600- 800 psi, providing a safety factor of 4.0-5.3X over the anticipated operating pressure of 150 psi and a safety factor of 3.0-4.0X over the working pressure of the hose (200 psi), independent of temperature effects. Radiation effects are minimal at doses less than 10 Mrad. Doses greater than 50 Mrad may be allowed, depending on operating conditions and required safety factors, but cannot be recommended at this time. At 250 Mrad, burst pressure values are reduced to the hose working pressure. At 300 Mrad, burst pressures are below 150 psi. At a bounding continuous dose rate of 57,870 rad/hr, the 50 Mrad dose limit is reached within 1.2 months. Actual dose rates may be lower, particularly during non-transfer periods. Refined dose calculations are therefore recommended to justify longer service. This report details the tests performed and interpretation of the results. Recommendations for shelf-life/storage, component quality verification, and post-service examination are provided.

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

    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

  7. Technology for the long-term management of defense HLW at the Idaho Chemical Processing Plant

    Staples, B.A.; Berreth, J.R.; Knecht, D.A.

    1986-01-01

    The Defense Waste Management Plan of June 1983 includes a reference plan for the long-term management of Idaho Chemical Processing Plant (ICPP) high-level waste (HLW), with a goal of disposing of the annual output in 500 canisters a year by FY-2008. Based on the current vitrification technology, the ICPP base-glass case would produce 1700 canisters per year after FY-2007. Thus, to meet the DWMP goal processing steps including fuel dissolution, waste treatment, and waste immobilization are being studied as areas where potential modifications could result in HLW volume reductions for repository disposal. It has been demonstrated that ICPP calcined wastes can be densified by hot isostatic pressing to multiphase ceramic forms of high loading and density. Conversion of waste by hot isostatic pressing to these forms has the potential of reducing the annual ICPP waste production to volumes near those of the goal of the DWMP. This report summarizes the laboratory-scale information currently available on the development of these forms

  8. R and D on HLW Partitioning in Russia

    Khaperskaya, A.; Babain, V.; Alyapyshev, M.

    2015-01-01

    Results of more than thirty years investigations on high level radioactive waste (HLW) partitioning in Russia are described. The objectives of research and development is to assess HLW partitioning technical feasibility and its advantages compared to direct vitrification of long-lived radionuclides. Many technological flowsheets for long-lived nuclides (cesium, strontium and minor actinides) separation were developed and tested with simulated and actual HLW. Different classes of extractants, including carbamoyl-phosphine oxides, dialkyl-phosphoric acids, crown ethers and diamides of heterocyclic acids were studied. Some of these processes were tested at PA 'Mayak' and MCC. Many extraction systems based on chlorinated cobalt dicarbollide (CCD), including UNEX-extractant and its modifications, were also observed. Diamides of diglycolic acid and diamides of heterocyclic acids in polar diluents have shown promising properties for minor actinide-lanthanide extraction and separation. Comparison of different solvents and possible ways of implementing new flowsheets in radiochemical technology are also discussed. (authors)

  9. HLW Canister and Can-In-Canister Drop Calculation

    H. Marr

    1999-01-01

    The purpose of this calculation is to evaluate the structural response of the standard high-level waste (HLW) canister and the HLW canister containing the cans of immobilized plutonium (''can-in-canister'' throughout this document) to the drop event during the handling operation. The objective of the calculation is to provide the structure parameter information to support the canister design and the waste handling facility design. Finite element solution is performed using the commercially available ANSYS Version (V) 5.4 finite element code. Two-dimensional (2-D) axisymmetric and three-dimensional (3-D) finite element representations for the standard HLW canister and the can-in-canister are developed and analyzed using the dynamic solver

  10. Final Report Start-Up And Commissioning Tests On The Duramelter 1200 HLW Pilot Melter System Using AZ-101 HLW Simulants VSL-01R0100-2, Rev. 0, 1/20/03

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Brandys, M.; Wilson, C.N.; Schatz, T.R.; Gong, W.; Pegg, I.L.

    2011-01-01

    This document provides the final report on data and results obtained from commissioning tests performed on the one-third scale DuraMelter(trademark) 1200 (DM 1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part BI (1). Both melters have similar melt surface areas (1.2 m 2 ) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plan. This report is a followup to the previously issued Preliminary Data Summary Report. The DM1200 system will be used for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. This will include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The results presented in this report are from the initial series of short-duration tests that were conducted to support the start-up and commissioning of this system prior to conducting the main body of development tests that have been planned for this system. These tests were directed primarily at system 'debugging,' operator training, and procedure refinement. The AZ-101 waste simulant and glass composition that was used for previous testing was selected for these tests.

  11. FINAL REPORT START-UP AND COMMISSIONING TESTS ON THE DURAMELTER 1200 HLW PILOT MELTER SYSTEM USING AZ-101 HLW SIMULANTS VSL-01R0100-2 REV 0 1/20/03

    KRUGER AA; MATLACK KS; KOT WK; BRANDYS M; WILSON CN; SCHATZ TR; GONG W; PEGG IL

    2011-12-29

    This document provides the final report on data and results obtained from commissioning tests performed on the one-third scale DuraMelter{trademark} 1200 (DM 1200) HLW Pilot Melter system that has been installed at VSL with an integrated prototypical off-gas treatment system. That system has replaced the DM1000 system that was used for HLW throughput testing during Part BI [1]. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. These tests were performed under a corresponding RPP-WTP Test Specification and associated Test Plan. This report is a followup to the previously issued Preliminary Data Summary Report. The DM1200 system will be used for testing and confirmation of basic design, operability, flow sheet, and process control assumptions as well as for support of waste form qualification and permitting. This will include data on processing rates, off-gas treatment system performance, recycle stream compositions, as well as process operability and reliability. Consequently, this system is a key component of the overall HLW vitrification development strategy. The results presented in this report are from the initial series of short-duration tests that were conducted to support the start-up and commissioning of this system prior to conducting the main body of development tests that have been planned for this system. These tests were directed primarily at system 'debugging,' operator training, and procedure refinement. The AZ-101 waste simulant and glass composition that was used for previous testing was selected for these tests.

  12. Vitrification of HLW in cold crucible melter

    Bordier, G.

    2005-01-01

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel the CEA (French Atomic Energy Commission), COGEMA (Industrial Operator), and SGN (COGEMA's Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities: the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed remotely in one of the R7 vitrification

  13. Long-term storage or disposal of HLW-dilemma

    Ninkovic, M. M.; Raicevic, J.

    1995-01-01

    In this paper, a new concept approach to HLW management founded on deterministic safety philosophy - i.e. long-term storage with final objective of destroying was justified and proposed instead of multi barrier concept with final disposal in extra stable environmental conditions, which are founded on probabilistic safety approach model. As a support to this new concept some methods for destruction of waste which are now accessible, on scientific stage only, as transmutation in fast reactors and accelerators of heavy ions were briefly discussed . It is justified to believe that industrial technology for destruction of HLW would be developed in not so far future. (author).

  14. Active geothermal systems as natural analogs of HLW repositories

    Elders, W.A.; Williams, A.E.; Cohen, L.H.

    1988-01-01

    Geologic analogs of long-lived processes in high-level waste (HLW) repositories have been much studied in recent years. However, most of these occurrences either involve natural processes going on today at 25 degree C, or, if they are concerned with behavior at temperatures similar to the peak temperatures anticipated near HLW canisters, have long since ended. This paper points out the usefulness of studying modern geothermal systems as natural analogs, and to illustrate the concept with a dramatic example, the Salton Sea geothermal system (SSGS)

  15. Enhanced sludge processing of HLW: Hydrothermal oxidation of chromium, technetium, and complexants by nitrate. 1997 mid-year progress report

    Buelow, S.

    1997-01-01

    'Treatment of High Level Waste (HLW) is the second most costly problem identified by OEM. In order to minimize costs of disposal, the volume of HLW requiring vitrification and long term storage must be reduced. Methods for efficient separation of chromium from waste sludges, such as the Hanford Tank Wastes (HTW), are key to achieving this goal since the allowed level of chromium in high level glass controls waste loading. At concentrations above 0.5 to 1.0 wt.% chromium prevents proper vitrification of the waste. Chromium in sludges most likely exists as extremely insoluble oxides and minerals, with chromium in the plus III oxidation state [1]. In order to solubilize and separate it from other sludge components, Cr(III) must be oxidized to the more soluble Cr(VI) state. Efficient separation of chromium from HLW could produce an estimated savings of $3.4B[2]. Additionally, the efficient separation of technetium [3], TRU, and other metals may require the reformulation of solids to free trapped species as well as the destruction of organic complexants. New chemical processes are needed to separate chromium and other metals from tank wastes. Ideally they should not utilize additional reagents which would increase waste volume or require subsequent removal. The goal of this project is to apply hydrothermal processing for enhanced chromium separation from HLW sludges. Initially, the authors seek to develop a fundamental understanding of chromium speciation, oxidation/reduction and dissolution kinetics, reaction mechanisms, and transport properties under hydrothermal conditions in both simple and complex salt solutions. The authors also wish to evaluate the potential of hydrothermal processing for enhanced separations of technetium and TRU by examining technetium and TRU speciation at hydrothermal conditions optimal for chromium dissolution.'

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

    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)

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

    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)

  18. Degradation mechanisms of organic rubber and glass flake/vinyl ester linings in flue gas desulphurization plants

    Weber, R [Allianz-Centre for Technology GmbH, Ismaning (Germany)

    1996-12-01

    In recent years, there have been reports in numerous publications about damage to rubber and glass flake coatings in flue gas desulphurization plants. The pattern of damage has been described and attempts have frequently been made to determine and explain the cause of the damage. Oxidation/hydrolytic changes were generally observed as the damage mechanisms. In addition, blistering occurs in both the chloroprene coatings in the absorbers and in the glass flake coatings in clean gas ducts. This blistering may be considered as the end of the useful life and leads to cost-intensive and time-intensive repair and restoration measures. The present state of knowledge suggests that the blistering is mainly due to osmotic processes preceded by permeation processes and with permeation processes superimposed on them. Among other things, the reports describe the permeation behaviour of water and other flue gas constituents; the blistering in chloroprene rubber coatings and glass flake coatings is explained by means of the knowledge gained. (au) 16 refs.

  19. Mechanical properties of nuclear waste glasses

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

    2011-01-01

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

  20. Glass leaching performance

    Chick, L.A.; Turcotte, R.P.

    1983-05-01

    Current understanding of the leaching performance of high-level nuclear waste (HLW) glass is summarized. The empirical model of waste glass leaching behavior developed shows that at high water flow rates the glass leach rate is kinetically limited to a maximum value. At intermediate water flow rates, leaching is limited by the solution concentration of silica and decreases with decreasing water flow rates. Release of soluble elements is controlled by silica dissolution because silica forms the binding network of the glass. At low water flow rates, mass loss rates reach values controlled by formation rates of alteration minerals, or by diffusion of dissolution products through essentially stagnant water. The parameters reviewed with respect to their quantifiable influence on leaching behavior include temperature, pH, leachant composition, glass composition, thermal history, and radiation. Of these, temperature is most important since the rate of mass loss approximately doubles with each 10 0 C increase in dilute solutions. The pH has small effects within the 4 to 10 range. The chemical composition of the leachant is most important with regard to its influence on alteration product formation. Glass composition exhibits the largest effects at high flow rates where improved glasses leach from ten to thirty times slower than glass 76 to 68. The effects of the thermal history (devitrification) of the glass are not likely to be significant. Radiation effects are important primarily in that radiolysis can potentially drive pH values to less than 4. Radiation damage to the glass causes insignificant changes in leaching performance

  1. Micrometer and nanometer scale photopatterning of proteins on glass surfaces by photo-degradation of films formed from oligo(ethylene glycol) terminated silanes.

    Tizazu, Getachew; el Zubir, Osama; Patole, Samson; McLaren, Anna; Vasilev, Cvetelin; Mothersole, David J; Adawi, Ali; Hunter, C Neil; Lidzey, David G; Lopez, Gabriel P; Leggett, Graham J

    2012-12-01

    Exposure of films formed by the adsorption of oligo(ethylene glycol) (OEG) functionalized trichlorosilanes on glass to UV light from a frequency-doubled argon ion laser (244 nm) causes photodegradation of the OEG chain. Although the rate of degradation is substantially slower than for monolayers of OEG terminated thiolates on gold, it is nevertheless possible to form micrometer-scale patterns by elective adsorption of streptavidin to exposed regions. A low density of aldehyde functional groups is produced, and this enables derivatization with nitrilotriacetic acid via an amine linker. Complexation with nickel enables the site-specific immobilization of histidine-tagged yellow and green fluorescent proteins. Nanometer-scale patterns may be fabricated using a Lloyd's mirror interferometer, with a sample and mirror set at right angles to each other. At low exposures, partial degradation of the OEG chains does not remove the protein-resistance of the surface, even though friction force microscopy reveals the formation of patterns. At an exposure of ca. 18 J cm(-2), the modified regions became adhesive to proteins in a narrow region ca. 30 nm (λ/8) wide. As the exposure is increased further the lines quickly broaden to ca. 90 nm. Adjustment of the angle between the sample and mirror enables the fabrication of lines of His-tagged green fluorescent protein at a period of 340 nm that could be resolved using a confocal microscope.

  2. Initiating the Validation of CCIM Processability for Multi-phase all Ceramic (SYNROC) HLW Form: Plan for Test BFY14CCIM-C

    Maio, Vince [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    This plan covers test BFY14CCIM-C which will be a first–of–its-kind demonstration for the complete non-radioactive surrogate production of multi-phase ceramic (SYNROC) High Level Waste Forms (HLW) using Cold Crucible Induction Melting (CCIM) Technology. The test will occur in the Idaho National Laboratory’s (INL) CCIM Pilot Plant and is tentatively scheduled for the week of September 15, 2014. The purpose of the test is to begin collecting qualitative data for validating the ceramic HLW form processability advantages using CCIM technology- as opposed to existing ceramic–lined Joule Heated Melters (JHM) currently producing BSG HLW forms. The major objectives of BFY14CCIM-C are to complete crystalline melt initiation with a new joule-heated resistive starter ring, sustain inductive melting at temperatures between 1600 to 1700°C for two different relatively high conductive materials representative of the SYNROC ceramic formation inclusive of a HLW surrogate, complete melter tapping and pouring of molten ceramic material in to a preheated 4 inch graphite canister and a similar canister at room temperature. Other goals include assessing the performance of a new crucible specially designed to accommodate the tapping and pouring of pure crystalline forms in contrast to less recalcitrant amorphous glass, assessing the overall operational effectiveness of melt initiation using a resistive starter ring with a dedicated power source, and observing the tapped molten flow and subsequent relatively quick crystallization behavior in pans with areas identical to standard HLW disposal canisters. Surrogate waste compositions with ceramic SYNROC forming additives and their measured properties for inductive melting, testing parameters, pre-test conditions and modifications, data collection requirements, and sampling/post-demonstration analysis requirements for the produced forms are provided and defined.

  3. Small Scale Mixing Demonstration Batch Transfer and Sampling Performance of Simulated HLW - 12307

    Jensen, Jesse; Townson, Paul; Vanatta, Matt [EnergySolutions, Engineering and Technology Group, Richland, WA, 99354 (United States)

    2012-07-01

    The ability to effectively mix, sample, certify, and deliver consistent batches of High Level Waste (HLW) feed from the Hanford Double Shell Tanks (DST) to the Waste treatment Plant (WTP) has been recognized as a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. At the end of 2009 DOE's Tank Operations Contractor, Washington River Protection Solutions (WRPS), awarded a contract to EnergySolutions to design, fabricate and operate a demonstration platform called the Small Scale Mixing Demonstration (SSMD) to establish pre-transfer sampling capacity, and batch transfer performance data at two different scales. This data will be used to examine the baseline capacity for a tank mixed via rotational jet mixers to transfer consistent or bounding batches, and provide scale up information to predict full scale operational performance. This information will then in turn be used to define the baseline capacity of such a system to transfer and sample batches sent to WTP. The Small Scale Mixing Demonstration (SSMD) platform consists of 43'' and 120'' diameter clear acrylic test vessels, each equipped with two scaled jet mixer pump assemblies, and all supporting vessels, controls, services, and simulant make up facilities. All tank internals have been modeled including the air lift circulators (ALCs), the steam heating coil, and the radius between the wall and floor. The test vessels are set up to simulate the transfer of HLW out of a mixed tank, and collect a pre-transfer sample in a manner similar to the proposed baseline configuration. The collected material is submitted to an NQA-1 laboratory for chemical analysis. Previous work has been done to assess tank mixing performance at both scales. This work involved a combination of unique instruments to understand the three dimensional distribution of solids using a combination of Coriolis meter measurements, in situ chord length distribution

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

    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

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

    Jantzen, C.M.

    2000-01-01

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

  6. Solidification of HLLW by glass-ceramic process

    Oguino, N.; Masuda, S.; Tsunoda, N.; Yamanaka, T.; Ninomiya, M.; Sakane, T.; Nakamura, S.; Kawamura, S.

    1979-01-01

    The compositions of glass-ceramics for the solidification of HLLW were studied, and the glass-ceramics in the diopside system was concluded to be the most suitable. Compared with the properties of HLW borosilicate glasses, those of diopside glass-ceramic were thought to be almost equal in leach rate and superior in thermal stability and mechanical strength. It was also found that the glass in this system can be crystallized simply by pouring it into a thermally insulated canister and then allowing it to cool to room temperature. 2 figures, 5 tables

  7. Generic Degraded Congiguration Probability Analysis for DOE Codisposal Waste Package

    S.F.A. Deng; M. Saglam; L.J. Gratton

    2001-05-23

    In accordance with the technical work plan, ''Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages'' (CRWMS M&O 2000c), this Analysis/Model Report (AMR) is developed for the purpose of screening out degraded configurations for U.S. Department of Energy (DOE) spent nuclear fuel (SNF) types. It performs the degraded configuration parameter and probability evaluations of the overall methodology specified in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000, Section 3) to qualifying configurations. Degradation analyses are performed to assess realizable parameter ranges and physical regimes for configurations. Probability calculations are then performed for configurations characterized by k{sub eff} in excess of the Critical Limit (CL). The scope of this document is to develop a generic set of screening criteria or models to screen out degraded configurations having potential for exceeding a criticality limit. The developed screening criteria include arguments based on physical/chemical processes and probability calculations and apply to DOE SNF types when codisposed with the high-level waste (HLW) glass inside a waste package. The degradation takes place inside the waste package and is long after repository licensing has expired. The emphasis of this AMR is on degraded configuration screening and the probability analysis is one of the approaches used for screening. The intended use of the model is to apply the developed screening criteria to each DOE SNF type following the completion of the degraded mode criticality analysis internal to the waste package.

  8. TWRS HLW interim storage facility search and evaluation

    Calmus, R.B., Westinghouse Hanford

    1996-05-16

    The purpose of this study was to identify and provide an evaluation of interim storage facilities and potential facility locations for the vitrified high-level waste (HLW) from the Phase I demonstration plant and Phase II production plant. In addition, interim storage facilities for solidified separated radionuclides (Cesium and Technetium) generated during pretreatment of Phase I Low-Level Waste Vitrification Plant feed was evaluated.

  9. Management strategy for site characterization at candidate HLW repository sites

    Bartlett, J.W.

    1988-01-01

    This paper describes a management strategy for HLW repository site characterization which is aimed at producing an optimal characterization trajectory for site suitability and licensing evaluations. The core feature of the strategy is a matrix of alternative performance targets and alternative information-level targets which can be used to allocate and justify program effort. Strategies for work concerning evaluation of expected and disrupted repository performance are distinguished, and the need for issue closure criteria is discussed

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

    Kruger, A.A.; Jain, V.

    2009-01-01

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

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

    KRUGER AA; VIENNA JD; KIM DS; JAIN V

    2009-05-27

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

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

    Vinjamuri, K.

    1990-01-01

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

  13. R and D programme for HLW disposal in Japan

    Tsuboya, Takao

    1997-01-01

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) has been active in developing an R and D programme for high-level radioactive waste (HLW) disposal in accordance with the overall HLW management programme defined by the Atomic Energy Commission (AEC) of Japan. The aim of the R and D activities at the current stage is to provide a scientific and technical basis for the geological disposal of HLW in Japan, which is turn promotes understanding of the safety concept not only in the scientific and technical community but also by the general public. As a major milestone in the R and D programme, PNC submitted a first progress report, referred to as H3, in September 1992. H3 summarised the results of R and D activities up to March 1992 and identified priority issues for further study. The second progress report, scheduled to be submitted around 2000, and should demonstrated more rigorously and transparently the feasibility of the specified disposal concept. It should also provide input for the siting and regulatory processes, which will be set in motion after the year 2000. (author). 10 refs., 4 figs

  14. A poly(glycerol sebacate)-coated mesoporous bioactive glass scaffold with adjustable mechanical strength, degradation rate, controlled-release and cell behavior for bone tissue engineering.

    Lin, Dan; Yang, Kai; Tang, Wei; Liu, Yutong; Yuan, Yuan; Liu, Changsheng

    2015-07-01

    Various requirements in the field of tissue engineering have motivated the development of three-dimensional scaffold with adjustable physicochemical properties and biological functions. A series of multiparameter-adjustable mesoporous bioactive glass (MBG) scaffolds with uncrosslinked poly(glycerol sebacate) (PGS) coating was prepared in this article. MBG scaffold was prepared by a modified F127/PU co-templating process and then PGS was coated by a simple adsorption and lyophilization process. Through controlling macropore parameters and PGS coating amount, the mechanical strength, degradation rate, controlled-release and cell behavior of the composite scaffold could be modulated in a wide range. PGS coating successfully endowed MBG scaffold with improved toughness and adjustable mechanical strength covering the bearing range of trabecular bone (2-12MPa). Multilevel degradation rate of the scaffold and controlled-release rate of protein from mesopore could be achieved, with little impact on the protein activity owing to an "ultralow-solvent" coating and "nano-cavity entrapment" immobilization method. In vitro studies indicated that PGS coating promoted cell attachment and proliferation in a dose-dependent manner, without affecting the osteogenic induction capacity of MBG substrate. These results first provide strong evidence that uncrosslinked PGS might also yield extraordinary achievements in traditional MBG scaffold. With the multiparameter adjustability, the composite MBG/PGS scaffolds would have a hopeful prospect in bone tissue engineering. The design considerations and coating method of this study can also be extended to other ceramic-based artificial scaffolds and are expected to provide new thoughts on development of future tissue engineering materials. Copyright © 2015 Elsevier B.V. All rights reserved.

  15. Focusing on clay formation as host media of HLW geological disposal in China

    Zheng Hualing; Chen Shi; Sun Donghui

    2007-01-01

    Host medium is vitally important for safety for HLW geological disposal. Chinese HLW disposal effort in the past decades were mainly focused on granite formation. However, the granite formation has fatal disadvantage for HLW geological disposal. This paper reviews experiences gained and lessons learned in the international community and analyzes key factors affecting the site selection. It is recommended that clay formation should be taken into consideration and additional effort should be made before decision making of host media of HLW disposal in China. (authors)

  16. Final Report Determination Of The Processing Rate Of RPP-WTP HLW Simulants Using A Duramelter J 1000 Vitrification System VSL-00R2590-2, Rev. 0, 8/21/00

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Perez-Cardenas, F.; Pegg, I.L.

    2011-01-01

    This report provides data, analysis, and conclusions from a series of tests that were conducted at the Vitreous State Laboratory of The Catholic University of America (VSL) to determine the melter processing rates that are achievable with RPP-WTP HLW simulants. The principal findings were presented earlier in a summary report (VSL-00R2S90-l) but the present report provides additional details. One of the most critical pieces of information in determining the required size of the RPP-WTP HLW melter is the specific glass production rate in terms of the mass of glass that can be produced per unit area of melt surface per unit time. The specific glass production rate together with the waste loading (essentially, the ratio of waste-in to glass-out, which is determined from glass formulation activities) determines the melt area that is needed to achieve a given waste processing rate with due allowance for system availability. As a consequence of the limited amount of relevant information, there exists, for good reasons, a significant disparity between design-base specific glass production rates for the RPP-WTP LAW and HLW conceptual designs (1.0 MT/m 2 /d and 0.4 MT/m 2 /d, respectively); furthermore, small-scale melter tests with HLW simulants that were conducted during Part A indicated typical processing rates with bubbling of around 2.0 MT/m 2 /d. This range translates into more than a factor of five variation in the resultant surface area of the HLW melter, which is clearly not without significant consequence. It is clear that an undersized melter is undesirable in that it will not be able to support the required waste processing rates. It is less obvious that there are potential disadvantages associated with an oversized melter, over and above the increased capital costs. A melt surface that is consistently underutilized will have poor cold cap coverage, which will result in increased volatilization from the melt (which is generally undesirable) and increased plenum

  17. FINAL REPORT DETERMINATION OF THE PROCESSING RATE OF RPP WTP HLW SIMULANTS USING A DURAMELTER J 1000 VITRIFICATION SYSTEM VSL-00R2590-2 REV 0 8/21/00

    KRUGER AA; MATLACK KS; KOT WK; PEREZ-CARDENAS F; PEGG IL

    2011-12-29

    This report provides data, analysis, and conclusions from a series of tests that were conducted at the Vitreous State Laboratory of The Catholic University of America (VSL) to determine the melter processing rates that are achievable with RPP-WTP HLW simulants. The principal findings were presented earlier in a summary report (VSL-00R2S90-l) but the present report provides additional details. One of the most critical pieces of information in determining the required size of the RPP-WTP HLW melter is the specific glass production rate in terms of the mass of glass that can be produced per unit area of melt surface per unit time. The specific glass production rate together with the waste loading (essentially, the ratio of waste-in to glass-out, which is determined from glass formulation activities) determines the melt area that is needed to achieve a given waste processing rate with due allowance for system availability. As a consequence of the limited amount of relevant information, there exists, for good reasons, a significant disparity between design-base specific glass production rates for the RPP-WTP LAW and HLW conceptual designs (1.0 MT/m{sup 2}/d and 0.4 MT/m{sup 2}/d, respectively); furthermore, small-scale melter tests with HLW simulants that were conducted during Part A indicated typical processing rates with bubbling of around 2.0 MT/m{sup 2}/d. This range translates into more than a factor of five variation in the resultant surface area of the HLW melter, which is clearly not without significant consequence. It is clear that an undersized melter is undesirable in that it will not be able to support the required waste processing rates. It is less obvious that there are potential disadvantages associated with an oversized melter, over and above the increased capital costs. A melt surface that is consistently underutilized will have poor cold cap coverage, which will result in increased volatilization from the melt (which is generally undesirable) and

  18. Glass Property Models and Constraints for Estimating the Glass to be Produced at Hanford by Implementing Current Advanced Glass Formulation Efforts

    Vienna, John D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kim, Dong-Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Skorski, Daniel C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-07-01

    Recent glass formulation and melter testing data have suggested that significant increases in waste loading in HLW and LAW glasses are possible over current system planning estimates. The data (although limited in some cases) were evaluated to determine a set of constraints and models that could be used to estimate the maximum loading of specific waste compositions in glass. It is recommended that these models and constraints be used to estimate the likely HLW and LAW glass volumes that would result if the current glass formulation studies are successfully completed. It is recognized that some of the models are preliminary in nature and will change in the coming years. Plus the models do not currently address the prediction uncertainties that would be needed before they could be used in plant operations. The models and constraints are only meant to give an indication of rough glass volumes and are not intended to be used in plant operation or waste form qualification activities. A current research program is in place to develop the data, models, and uncertainty descriptions for that purpose. A fundamental tenet underlying the research reported in this document is to try to be less conservative than previous studies when developing constraints for estimating the glass to be produced by implementing current advanced glass formulation efforts. The less conservative approach documented herein should allow for the estimate of glass masses that may be realized if the current efforts in advanced glass formulations are completed over the coming years and are as successful as early indications suggest they may be. Because of this approach there is an unquantifiable uncertainty in the ultimate glass volume projections due to model prediction uncertainties that has to be considered along with other system uncertainties such as waste compositions and amounts to be immobilized, split factors between LAW and HLW, etc.

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

    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.

  20. Rheology of Savannah River site tank 42 HLW radioactive sludge

    Ha, B.C.

    1997-01-01

    Knowledge of the rheology of the radioactive sludge slurries at the Savannah River Site is necessary in order to ensure that they can be retrieved from waste tanks and processed for final disposal. At Savannah River Site, Tank 42 sludge represents on of the first HLW radioactive sludges to be vitrified in the Defense Waste Processing Facility. The rheological properties of unwashed Tank 42 sludge slurries at various solids concentrations were measured remotely in the Shielded Cells at the Savannah River Technology Center using a modified Haake Rotovisco viscometer

  1. Durable Glass For Thousands Of Years

    Jantzen, C.

    2009-01-01

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

  2. DURABLE GLASS FOR THOUSANDS OF YEARS

    Jantzen, C.

    2009-12-04

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

  3. Issues at stake when considering long term storage of HLW. A comprehensive approach to designing the facility

    Marvy, A.; Ochem, D.

    2002-01-01

    CEA has been conducting a comprehensive R and D program to identify and study key HLW storage design criteria to possibly meet the lifetime goal of a century and beyond. A novel approach is being used since such installations must be understood as a global system comprised of various materials and hardware components, canisters, concrete and steel structures and specific procedures covering engineering steps from construction to operation including monitoring, care and maintenance as well as licensing. The challenge set by such a lifetime design goal made the R and D people focus on issues at stake and relevant to long term HLW storage in particular heat management, the effect of time on materials and the sustainability of care and maintenance. This opened up the R and D field from fundamental research areas to more conventional and technical aspects. Two major guiding principles have been devised as key design goals for the storage concepts under consideration. One is the paramount function of retrievability, which must allow the safe retrieval of any HLW package from the facility at any given time. Next is the passive containment philosophy requiring that a two-barrier system be considered. In the case of spent fuel, CEA's early assessment of the long-term behaviour of cladding shows that it cannot qualify as a reliable barrier over a long period of time. Therefore, the overriding strategy of preventing corrosion and material degradation to achieve canister protection, and therefore containment of radioactive material throughout the time of period envisaged, is at the heart of the R and D program and several design alternatives are being studied to meet that objective. For instance available thermal power from SF is used to establish dry corrosion conditions within the storage facility. The paper reviews all of these different R and D and engineering aspects. (author)

  4. Immobilization of hazardous and radioactive waste into glass structures

    Wicks, G.G.

    1997-01-01

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

  5. The chemical stockpile intergovernmental consultation program: Lessons for HLW public involvement

    Feldman, D.L.

    1991-01-01

    This paper assesses the appropriateness of the US Army's Chemical Stockpile Disposal Program's (CSDP) Intergovernmental Consultation and Coordination Boards (ICCBs) as models for incorporating public concerns in the future siting of HLW repositories by DOE. ICCB structure, function, and implementation are examined, along with other issues relevant to the HLW context. 27 refs

  6. Comparison of risks due to HLW and SURF repositories in bedded salt

    Chu, M.S.Y.; Ortiz, N.R.; Wahi, K.K.

    1983-01-01

    A methodology was developed for use in the analysis of risks from geologic disposal of nuclear wastes. This methodology is applied to two conceptual nuclear waste repositories in bedded salt containing High-Level Waste (HLW) and Spent Un-Reprocessed Fuel (SURF), respectively. A comparison of the risk estimated from the HLW and SURF repositories is presented

  7. Preliminary formulation studies for a ''hydroceramic'' alternative waste form for INEEL HLW

    Siemer, D.D.; Gougar, M.L.D.; Grutzeck, M.W.; Scheetz, B.E.

    1999-01-01

    Herein the authors discuss scoping studies performed to develop an efficient way to prepare the Idaho National Engineering and Environmental Laboratory (INEEL) nominally high-level (∼40 W/m 3 ) calcined radioactive waste (HLW) and liquid metal (sodium) reactor coolants for disposal. The investigated approach implements the chemistry of Hanford's cancrinite-making clay reaction process via Oak Ridge National Laboratory's (ORNL's) formed-under-elevated-temperatures-and-pressures concrete monolith-making technology to make hydroceramics (HCs). The HCs differ from conventional Portland cement/blast furnace slag (PC/BFS) grouts in that the binder minerals formed during the curing process are hydrated alkali-aluminosilicates (feldspathoids-sodalites, cancrinites, and zeolites) rather than hydrated calcium silicates (CSH). This is desirable because (a) US defense-type radioactive wastes generally contain much more sodium and aluminum than calcium; (b) sodalites/cancrinites do a much better job of retaining the anionic components of real radioactive waste (e.g., nitrate) than do calcium silicates; (c) natural feldspathoids form from glasses (and therefore are more stable) in that region of the United States where a repository for this sort of waste could be sited; and (d) if eventually deemed necessary, feldspathoid-type concrete wasteforms could be hot-isostatically-pressed into even more durable materials without removing them from their original canisters

  8. Spent fuel and HLW transportation the French experience

    Giraud, J.P.; Charles, J.L.

    1995-01-01

    With 53 nuclear power plants in operation at EDF and a fuel cycle with recycling policy of the valuable materials, COGEMA is faced with the transport of a wide range of radioactive materials. In this framework, the transport activity is a key link in closing the fuel cycle. COGEMA has developed a comprehensive Transport Organization System dealing with all the sectors of the fuel cycle. The paper will describe the status of transportation of spent fuel and HLW in France and the experience gathered. The Transport Organization System clearly defines the role of all actors where COGEMA, acting as the general coordinator, specifies the tasks to be performed and brings technical and commercial support to its various subcontractors: TRANSNUCLEAIRE, specialized in casks engineering and transport operations, supplies packaging and performs transport operations, LEMARECHAL and CELESTIN operate transport by truck in the Vicinity of the nuclear sites while French Railways are in charge of spent fuel transport by train. HLW issued from the French nuclear program is stored for 30 years in an intermediate storage installation located at the La Hague reprocessing plant. Ultimately, these canisters will be transported to the disposal site. COGEMA has set up a comprehensive transport organization covering all operational aspects including adapted procedures, maintenance programs and personnel qualification

  9. NOx AND HETEROGENEITY EFFECTS IN HIGH LEVEL WASTE (HLW)

    Meisel, Dan; Camaioni, Donald M.; Orlando, Thom

    2000-01-01

    We summarize contributions from our EMSP supported research to several field operations of the Office of Environmental Management (EM). In particular we emphasize its impact on safety programs at the Hanford and other EM sites where storage, maintenance and handling of HLW is a major mission. In recent years we were engaged in coordinated efforts to understand the chemistry initiated by radiation in HLW. Three projects of the EMSP (''The NOx System in Nuclear Waste,'' ''Mechanisms and Kinetics of Organic Aging in High Level Nuclear Wastes, D. Camaioni--PI'' and ''Interfacial Radiolysis Effects in Tanks Waste, T. Orlando--PI'') were involved in that effort, which included a team at Argonne, later moved to the University of Notre Dame, and two teams at the Pacific Northwest National Laboratory. Much effort was invested in integrating the results of the scientific studies into the engineering operations via coordination meetings and participation in various stages of the resolution of some of the outstanding safety issues at the sites. However, in this Abstract we summarize the effort at Notre Dame

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

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

    1982-01-01

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

  11. Calculated leaching of certain fission products from a cylinder of French glass

    Blomqvist, G.

    1977-07-01

    The probable total leaching of the most important fission products and actinides have been tabulated for a cylinder of French HLW glass with approximately 9 percent fission products. The calculations cover the period between 30 and 10000 years after removal from the reactor. The cylinder is of the type planned for the introduction of the HLW into Swedish crystalline rocks. All the components are supposed to have the same leach rate. The calculations also include the probable thickness of eroded glass layer/year. (author)

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

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

    2016-10-15

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

  13. Computer Modeling Of High-Level Waste Glass Temperatures Within DWPF Canisters During Pouring And Cool Down

    Amoroso, J.

    2011-01-01

    This report describes the results of a computer simulation study to predict the temperature of the glass at any location inside a DWPF canister during pouring and subsequent cooling. These simulations are an integral part of a larger research focus aimed at developing methods to predict, evaluate, and ultimately suppress nepheline formation in HLW glasses. That larger research focus is centered on holistically understanding nepheline formation in HLW glass by exploring the fundamental thermal and chemical driving forces for nepheline crystallization with respect to realistic processing conditions. Through experimental work, the goal is to integrate nepheline crystallization potential in HLW glass with processing capability to ultimately optimize waste loading and throughput while maintaining an acceptable product with respect to durability. The results of this study indicated severe temperature gradients and prolonged temperature dwell times exist throughout different locations in the canister and that the time and temperatures that HLW glass is subjected to during processing is a function of pour rate. The simulations indicate that crystallization driving forces are not uniform throughout the glass volume in a DWPF (or DWPF-like) canister and illustrate the importance of considering overall kinetics (chemical and thermal driving forces) of nepheline formation when developing methods to predict and suppress its formation in HLW glasses. The intended path forward is to use the simulation data both as a driver for future experimental work and, as an investigative tool for evaluating the impact of experimental results. Simulation data will be used to develop laboratory experiments to more acutely evaluate nepheline formation in HLW glass by incorporating the simulated temperatures throughout the canister into the laboratory experiments. Concurrently, laboratory experiments will be performed to identify nepheline crystallization potential in HLW glass as a function of

  14. Demonstration of pyrometallurgical processing for metal fuel and HLW

    Tadafumi, Koyama; Kensuke, Kinoshita; Takatoshi, Hizikata; Tadashi, Inoue; Ougier, M.; Rikard, Malmbeck; Glatz, J.P.; Lothar, Koch

    2001-01-01

    CRIEPI and JRC-ITU have started a joint study on pyrometallurgical processing to demonstrate the capability of this type of process for separating actinide elements from spent fuel and HLW. The equipment dedicated for this experiments has been developed and installed in JRC-ITU. The stainless steel box equipped with tele-manipulators is operated under pure Ar atmosphere, and prepared for later installation in a hot cell. Experiments on pyro-processing of un-irradiated U-Pu-Zr metal alloy fuel by molten salt electrorefining has been carried out. Recovery of U and Pu from this type alloy fuel was first demonstrated with using solid iron cathode and liquid Cd cathode, respectively. (author)

  15. Development of gap filling technique in HLW repository

    Nakashima, Hitoshi; Saito, Akira; Ishii, Takashi; Toguri, Satohito; Okihara, Mitsunobu; Iwasa, Kengo

    2016-01-01

    HLW is supposed to be disposed underground at depths more than 300 m in Japan. Buffer is an artificial barrier that controls radionuclides migrating into the groundwater. The buffer would be made of a natural swelling clay, bentonite. Construction technology for the buffer has been studied for many years, but studies for the gaps surrounding the buffer are little. The proper handling of the gaps is important for guaranteeing the functions of the buffer. In this paper, gap filling techniques using bentonite pellets have been developed in order to the gap having the same performance as the buffer. A new method for manufacturing high-density spherical pellets has been developed to fill the gap higher density ever reported. For the bentonite pellets, the filling performance and how to use were determined. And full-scale filling tests provided availability of the bentonite pellets and filling techniques. (author)

  16. Historical fuel reprocessing and HLW management in Idaho

    Knecht, D.A.; Staiger, M.D.; Christian, J.D.

    1997-01-01

    This article review some of the key decision points in the historical development of spent fuel reprocessing and waste management practices at the Idaho Chemical Processing Plant that have helped ICPP to successfully accomplish its mission safely and with minimal impact on the environment. Topics include ICPP reprocessing development; batch aluminum-uranium dissolution; continuous aluminum uranium dissolution; batch zirconium dissolution; batch stainless steel dissolution; semicontinuous zirconium dissolution with soluble poison; electrolytic dissolution of stainless steel-clad fuel; graphite-based rover fuel processing; fluorinel fuel processing; ICPP waste management consideration and design decisions; calcination technology development; ICPP calcination demonstration and hot operations; NWCF design, construction, and operation; HLW immobilization technology development. 80 refs., 4 figs

  17. Comparing technical concepts for disposal of Belgian vitrified HLW

    Bel, J.; Bock, C. de; Boyazis, J.P.

    2004-01-01

    The choice of a suitable repository design for different categories of radioactive waste is an important element in the decisional process that will eventually lead to the waste disposal in geological ground layers during the next decades. Most countries are in the process of elaborating different technical solutions for their EBS '. Considering possible design alternatives offers more flexibility to cope with remaining uncertainties and allows optimizing some elements of the EBS in the future. However, it is not feasible to continue carrying out detailed studies for a large number of alternative design options. At different stages in the decisional process, choices, even preliminary ones, have to be made. Although the impact of different stakeholders (regulator, waste agencies, waste producers, research centers,...) in making these design choices can differ from one country to another, the choices should be based on sound, objective, clear and unambiguous justification grounds. Moreover, the arguments should be carefully reported and easy to understand by the decision makers. ONDRAF/NIRAS recently elaborated three alternative designs for the disposal of vitrified HLW. These three designs are briefly described in the next section. A first series of technological studies pointed out that the three options are feasible. It would however be unreasonable to continue R and D work on all three alternatives in parallel. It is therefore planned to make a preliminary choice of a reference design for the vitrified HLW in 2003. This selection will depend on the way the alternative design options can be evaluated against a number of criteria, mainly derived from general repository design requirements. The technique of multi-criteria analysis (MCA) will be applied as a tool for making the optimum selection, considering all selection criteria and considering different strategic approaches. This paper describes the used methodology. The decision on the actual selection will be

  18. Glass Composition Constraint Recommendations for Use in Life-Cycle Mission Modeling

    McCloy, John S.; Vienna, John D.

    2010-05-03

    The component concentration limits that most influence the predicted Hanford life-cycle HLW glass volume by HTWOS were re-evaluated. It was assumed that additional research and development work in glass formulation and melter testing would be performed to improve the understanding of component effects on the processability and product quality of these HLW glasses. Recommendations were made to better estimate the potential component concentration limits that could be applied today while technology development is underway to best estimate the volume of HLW glass that will eventually be produced at Hanford. The limits for concentrations of P2O5, Bi2O3, and SO3 were evaluated along with the constraint used to avoid nepheline formation in glass. Recommended concentration limits were made based on the current HLW glass property models being used by HTWOS (Vienna et al. 2009). These revised limits are: 1) The current ND should be augmented by the OB limit of OB ≤ 0.575 so that either the normalized silica (NSi) is less that the 62% limit or the OB is below the 0.575 limit. 2) The mass fraction of P2O5 limit should be revised to allow for up to 4.5 wt%, depending on CaO concentrations. 3) A Bi2O3 concentration limit of 7 wt% should be used. 4) The salt accumulation limit of 0.5 wt% SO3 may be increased to 0.6 wt%. Again, these revised limits do not obviate the need for further testing, but make it possible to more accurately predict the impact of that testing on ultimate HLW glass volumes.

  19. Improvement of database on glass dissolution

    Hayashi, Maki; Sasamoto, Hiroshi; Yoshikawa, Hideki

    2008-03-01

    In geological disposal system, high-level radioactive waste (HLW) glass is expected to retain radionuclide for the long term as the first barrier to prevent radionuclide release. The advancement of its performance assessment technology leads to the reliability improvement of the safety assessment of entire geological disposal system. For this purpose, phenomenological studies for improvement of scientific understanding of dissolution/alteration mechanisms, and development of robust dissolution/alteration model based on the study outcomes are indispensable. The database on glass dissolution has been developed for supporting these studies. This report describes improvement of the prototype glass database. Also, this report gives an example of the application of the database for reliability assessment of glass dissolution model. (author)

  20. Liquidus Temperature Data for DWPF Glass

    Piepel, G.F.; Vienna, J.D.; Crum, J.V.; Mika, M.; Hrma, P.

    1999-01-01

    This report provides new liquidus temperature (T L ) versus composition data that can be used to reduce uncertainty in T L calculation for DWPF glass. According to the test plan and test matrix design PNNL has measured T L for 53 glasses within and just outside of the current DWPF processing composition window. The T L database generated under this task will directly support developing and enhancing the current T L process-control model. Preliminary calculations have shown a high probability of increasing HLW loading in glass produced at the SRS and Hanford. This increase in waste loading will decrease the life-cycle tank cleanup costs by decreasing process time and the volume of waste glass produced

  1. Microcrack growing and long-term mechanical stability in a HLW deep-borehole repository in granite

    Biurrun, E.; Hahne, K.

    1989-01-01

    The long-term host rock integrity assessment of a deep borehole emplacement for HLW in granite has been addressed with a detailed new constitutive model considering temperature and pressure effects on microscale phenomena (as microcracking) under repository conditions. The results of these finite element calculations have been compared with results obtained using conventional, state-of-the-art constitutive modelling. While the results of conventional modelling did suggest the existence of an important safety margin before failure, the improved calculations with the new model predict a thin but very long region of degradated host rock along the waste canister column. The results obtained up to now may well be considered as safety relevant, because they suggest that the actual long-term granite strength lies well below the conventionally determined failure limits, thus challenging the barrier properties of this host rock if the actual strength is not properly considered in the repository design

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

    CAROL, JANTZEN

    2005-01-01

    atomic ratios is shown to represent the structural effects of the glass on the dissolution and the formation of activated complexes in the glass leached layer. This provides two different methods by which a linear glass durability model can be formulated. One based on the quasi- crystalline mineral species in a glass and one based on cation ratios in the glass: both are related to the activated complexes on the surface by the law of mass action. The former would allow a new Thermodynamic Hydration Energy Model to be developed based on the hydration of the quasi-crystalline mineral species if all the pertinent thermodynamic data were available. Since the pertinent thermodynamic data is not available, the quasi-crystalline mineral species and the activated complexes can be related to cation ratios in the glass by the law of mass action. The cation ratio model can, thus, be used by waste form producers to formulate durable glasses based on fundamental structural and activated complex theories. Moreover, glass durability model based on atomic ratios simplifies HLW glass process control in that the measured ratios of only a few waste components and glass formers can be used to predict complex HLW glass performance with a high degree of accuracy, e.g. an R 2 approximately 0.97

  3. New glass material oxidation and dissolution system facility: Direct conversion of surplus fissile materials, spent nuclear fuel, and other material to high-level-waste glass. Storage and disposition of weapons-usable fissile materials programmatic environmental impact statement data report: Predecisional draft

    Forsberg, C.W.; Elam, K.R.; Reich, W.J.

    1995-01-01

    With the end of the Cold War, countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. It has been recommended that these surplus fissile materials (SFMs) be processed so that they are no more accessible than plutonium in spent nuclear fuel (SNF). This SNF standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. This report provides for the PEIS the necessary input data on a new method for the disposition of SFMs: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptunium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal. The primary form of this SNF is Hanford-N SNF with preirradiation uranium enrichments between 0.95 and 1.08%. The final product is a plutonium, low-enriched-uranium, HLW, borosilicate glass for disposition in a geological repository. The proposed conversion process is the Glass Material Oxidation and Dissolution System (GMODS), which is a new process. The initial analysis of the GMODS process indicates that a MODS facility for this application would be similar in size and environmental impact to the Defense Waste Processing Facility (DWPF) at the Savannah River Site. Because of this, the detailed information available on DWPF was used as the basis for much of the GMODS input into the SFMs PEIS

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

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

    2015-07-01

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

  5. FINAL REPORT INTEGRATED DM1200 MELTER TESTING OF REDOX EFFECTS USING HLW AZ-101 AND C-106/AY-102 SIMULANTS VSL-04R4800-1 REV 0 5/6/

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; LUTZE W; BIZOT PM; CALLOW RA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 and C-106/AY-102 HLW simulants. The tests reported herein are a subset of three tests from a larger series of tests described in the Test Plan for the work; results from the remaining tests will be reported separately. Three nine day tests, one with AZ-101 and two with C-106/AY-102 feeds were conducted with variable amounts of added sugar to address the effects of redox. The test with AZ-101 included ruthenium spikes to also address the effects of redox on ruthenium volatility. One of tests addressed the effects of increased flow-sheet nitrate levels using C-106/AY-102 feeds. With high nitrate/nitrite feeds (such as WTP LAW feeds), reductants are required to prevent melt foaming and deleterious effects on glass production rates. Sugar is the baseline WTP reductant for this purpose. WTP HLW feeds typically have relatively low nitrate/nitrite content in comparison to the organic carbon content and, therefore, have typically not required sugar additions. However, HLW feed variability, particularly with respect to nitrate levels, may necessitate the use of sugar in some instances. The tests reported here investigate the effects of variable sugar additions to the melter feed as well as elevated nitrate levels in the waste. Variables held constant to the extent possible included melt temperature, bubbling rate, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW feeds with variable amounts of added sugar and increased nitrate levels; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and

  6. Final Report Integrated DM1200 Melter Testing Of Redox Effects Using HLW AZ-101 And C-106/AY-102 Simulants VSL-04R4800-1, Rev. 0, 5/6/04

    Kruger, A.A.; Matlack, K.S.; Gong, W.; Bardakci, T.; D'Angelo, N.A.; Lutze, W.; Bizot, P.M.; Callow, R.A.; Brandys, M.; Kot, W.K.; Pegg, I.L.

    2011-01-01

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 and C-106/AY-102 HLW simulants. The tests reported herein are a subset of three tests from a larger series of tests described in the Test Plan for the work; results from the remaining tests will be reported separately. Three nine day tests, one with AZ-101 and two with C-106/AY-102 feeds were conducted with variable amounts of added sugar to address the effects of redox. The test with AZ-101 included ruthenium spikes to also address the effects of redox on ruthenium volatility. One of tests addressed the effects of increased flow-sheet nitrate levels using C-106/AY-102 feeds. With high nitrate/nitrite feeds (such as WTP LAW feeds), reductants are required to prevent melt foaming and deleterious effects on glass production rates. Sugar is the baseline WTP reductant for this purpose. WTP HLW feeds typically have relatively low nitrate/nitrite content in comparison to the organic carbon content and, therefore, have typically not required sugar additions. However, HLW feed variability, particularly with respect to nitrate levels, may necessitate the use of sugar in some instances. The tests reported here investigate the effects of variable sugar additions to the melter feed as well as elevated nitrate levels in the waste. Variables held constant to the extent possible included melt temperature, bubbling rate, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW feeds with variable amounts of added sugar and increased nitrate levels; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and perform pre- and

  7. Conclusions on the two technical panels on HLW-disposal and waste treatment processes respectively

    Dinkespiller, J.A.; Dejonghe, P.; Feates, F.

    1986-01-01

    The paper reports the concluding panel session at the European Community Conference on radioactive waste management and disposal, Luxembourg 1985. The panel considered the conclusions of two preceeding technical panels on high level waste (HLW) disposal and waste treatment processes. Geological disposal of HLW, waste management, safety assessment of waste disposal, public opinion, public acceptance of the manageability of radioactive wastes, international cooperation, and waste management in the United States, are all discussed. (U.K.)

  8. Legal precedents regarding use and defensibility of risk assessment in Federal transportation of SNF and HLW

    Bentz, E.J. Jr.; Bentz, C.B.; O'Hora, T.D.; Chen, S.Y.

    1997-01-01

    Risk assessment has become an increasingly important and essential tool in support of Federal decision-making regarding the handling, storage, disposal, and transportation of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). This paper analyzes the current statutory and regulatory framework and related legal precedents with regard to SNF and HLW transportation. The authors identify key scientific and technical issues regarding the use and defensibility of risk assessment in Federal decision-making regarding anticipated shipments

  9. The experiment of affective web risk communication on HLW geological disposal

    Kugo, Akihide; Yoshikawa, Eiwa; Wakabayashi, Yasunaga; Shimoda, Hiroshi; Uda, Akinobu; Ito, Kyoko

    2006-01-01

    Dialog mode web contents regarding the HLW risk is effective to altruism. To make it more effectively, we introduced affective elements such as facial expression of character agents and sympathetic response on the BBS by experts, which brought us smooth risk communication. This paper describes the result of preliminary experiments surrounding the affective ways to communicate on the risk of HLW geological disposal, leading to enhance the social cooperation, and the public open experiment for one month on the Web. (author)

  10. Nuclide transport models for HLW repository safety assessment in Finland, Japan, Sweden, and Canada

    Lee, Young Myoung; Kang, Chul Hyung; Hwang, Yong Soo; Choi, Jong Won; Kim, Sung Gi; Koh, Won Il

    1997-10-01

    Disposal and design concepts in such countries as Sweden, Finland, Canada and Japan which have already published safety assessment reports for the HLW repositories have been reviewed mainly in view of nuclide transport models used in their assessment. This kind of review would be very helpful in doing similar research in Korea where research program regarding HLW has been just started. (author). 44 refs., 2 tabs., 30 figs

  11. Rheology of Savannah River site tank 42 and tank 51 HLW radioactive sludges

    Ha, B.C.; Bibler, N.E.

    1996-01-01

    Knowledge of the rheology of the radioactive sludge slurries at the Savannah River Site (SRS) is necessary in order to ensure that they can be retrieved from waste tanks and processed for final disposal. The high activity radioactive wastes stored as caustic slurries at SRS result from the neutralization of acid waste generated from production of nuclear defense materials. During storage, the wastes separate into a supernate layer and a sludge layer. In the Defense Waste Processing Facility (DWPF) at SRS, the radionuclides from the sludge and supernate will be immobilized into borosilicate glass for long term storage and eventual disposal. Before transferring the waste from a storage tank to the DWPF, a portion of the aluminum in the waste sludge will be dissolved and the sludge will be extensively washed to remove sodium. Tank 51 and Tank 42 radioactive sludges represent the first batch of HLW sludge to be processed in the DWPF. This paper presents results of rheology measurements of Tank 51 and Tank 42 at various solids concentrations. The rheologies of Tank 51 and Tank 42 radioactive slurries were measured remotely in the Shielded Cells Operations (SCO) at the Savannah River Technology Center (SRTC) using a modified Haake Rotovisco RV-12 with an M150 measuring drive unit and TI sensor system. Rheological properties of the Tank 51 and Tank 42 radioactive sludges were measured as a function of weight percent solids. The weight percent solids of Tank 42 sludge was 27, as received. Tank 51 sludge had already been washed. The weight percent solids were adjusted by dilution with water or by concentration through drying. At 12, 15, and 18 weight percent solids, the yield stresses of Tank 51 sludge were 5, 11, and 14 dynes/cm2, respectively. The apparent viscosities were 6, 10, and 12 centipoises at 300 sec-1 shear rate, respectively

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

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

    2006-01-01

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

  13. Public Perspectives in the Japanese HLW Disposal Program

    Inatsugu, Shigefumi; Takeuchi, Mitsuo; Kato, Toshiaki [Nuclear Waste Management Organization of Japan (NUNIO), Tokyo (Japan)

    2006-09-15

    Following legislation entitled the 'Specified Radioactive Waste Final Disposal Act', the Nuclear Waste Management Organization of Japan (NUMO) was established in October 2000 as the implementing organization for geological disposal of vitrified high-level waste (HLW). Implementation of NUMO's disposal project will be based on three principles: 1) respecting public initiative and opinion, 2) adopting a stepwise approach and 3) ensuring transparency in information disclosure. NUMO has decided to adopt an open solicitation approach to finding volunteer municipalities for Preliminary Investigation Areas (PIAs). The official announcement of the start of the open solicitation program was made in 2002. Although no official applications had been received from volunteer municipalities by the end of 2005, NUMO has been continuing to carry out various activities aimed specifically at public communication and encouraging dialogue about the deep geological disposal project This paper summarizes the results obtained and lessons learned so far and identifies the issues that NUMO must tackle immediately in the areas of communication and dialogue.

  14. Stress analysis of HLW containers advanced test work Compas project

    Ove Arup and Partners

    1990-01-01

    The Compas project is concerned with the structural performance of metal overpacks which may be used to encapsulate vitrified high-level waste forms before disposal in deep geological repositories. This document describes the activities performed between June and August 1989 forming the advanced test work phase of this project. This is the culmination of two years' analysis and test work to demonstrate whether the analytical ability exists to model containers subjected to realistic loads. Three mild steel containers were designed and manufactured to be one-third scale models of a realistic HLW container, modified to represent the effect of anisotropic loading and to facilitate testing. The containers were tested under a uniform external pressure and all failed by buckling in the mid-body region. The outer surface of each container was comprehensively strain-gauged to provide strain history data at all positions of interest. In parallel with the test work, Compas project partners, from five different European countries, independently modelled the behaviour of each of the containers using their computer codes to predict the failure pressure and produce strain history data at a number of specified locations. The first axisymmetric container was well modelled but predictions for the remaining two non-axisymmetric containers were much more varied, with differences of up to 50% occurring between failure predictions and test data

  15. Technical and economic optimization study for HLW waste management

    Deffes, A.

    1989-01-01

    This study was conducted to assess the technical and economic aspects of high level waste (HLW) management with the objective of optimizing the interim storage duration and the dimensions of the underground repository site. The procedure consisted in optimizing the economic criterion under specified constraints. The results are intended to identify trends and guide the choice from among available options; simple and highly flexible models were therefore used in this study, and only nearfield thermal constraints were taken into consideration. Because of the present uncertainty on the physicochemical properties of the repository environment and on the unit cost figures, this study focused on developing a suitable method rather than on obtaining definitive results. With the physical and economic data bases used for the two media investigated (granite and salt) the optimum values found show that it is advisable to minimize the interim storage time, and that the geological repository should feature a high degree of spatial dilution. These results depend to a considerable extent on the assumption of high interim storage costs

  16. Public Perspectives in the Japanese HLW Disposal Program

    Inatsugu, Shigefumi; Takeuchi, Mitsuo; Kato, Toshiaki

    2006-01-01

    Following legislation entitled the 'Specified Radioactive Waste Final Disposal Act', the Nuclear Waste Management Organization of Japan (NUMO) was established in October 2000 as the implementing organization for geological disposal of vitrified high-level waste (HLW). Implementation of NUMO's disposal project will be based on three principles: 1) respecting public initiative and opinion, 2) adopting a stepwise approach and 3) ensuring transparency in information disclosure. NUMO has decided to adopt an open solicitation approach to finding volunteer municipalities for Preliminary Investigation Areas (PIAs). The official announcement of the start of the open solicitation program was made in 2002. Although no official applications had been received from volunteer municipalities by the end of 2005, NUMO has been continuing to carry out various activities aimed specifically at public communication and encouraging dialogue about the deep geological disposal project This paper summarizes the results obtained and lessons learned so far and identifies the issues that NUMO must tackle immediately in the areas of communication and dialogue

  17. Thermal analysis of the vertical disposal for HLW

    Zhao Honggang; Wang Ju; Liu Yuemiao; Su Rui

    2013-01-01

    The temperature on the canister surface is set to be no more than 100℃ in the high level radioactive waste (HLW) repository, it is a criterion to dictate the thermal dimension of the repository. The factors that affect the temperature on the canister surface include the initial power of the canister, the thermal properties of material as the engineered barrier system (EBS), the gaps around the canister in the EBS, the initial ground temperature and thermal properties of the host rock, the repository layout, etc. This article examines the thermal properties of the material in host rock and the EBS, the thermal conductivity properties of the different gaps in the EBS, the temperature evolution around the single canister by using the analysis method and the numerical method. The findings are as follows: 1) The most important and the sensitive parameter is the initial disposal power of the canister; 2) The two key factors that affect the highest temperature on the canister surface are the parameter of uncertainty and nature variability of material as the host rock and the EBS, and the gaps around the canister in the EBS; 3) The temperature difference between the canister and bentonite is no more than 10℃ , and the bigger the inner gaps are, the bigger the temperature difference will be; when the gap between the bentonite and the host rock is filled with water, the temperature difference becomes small, but it will be 1∼3℃ higher than the gaps filled will air. (authors)

  18. Biosphere modelling for a HLW repository - scenario and parameter variations

    Grogan, H.

    1985-03-01

    In Switzerland high-level radioactive wastes have been considered for disposal in deep-lying crystalline formations. The individual doses to man resulting from radionuclides entering the biosphere via groundwater transport are calculated. The main recipient area modelled, which constitutes the base case, is a broad gravel terrace sited along the south bank of the river Rhine. An alternative recipient region, a small valley with a well, is also modelled. A number of parameter variations are performed in order to ascertain their impact on the doses. Finally two scenario changes are modelled somewhat simplistically, these consider different prevailing climates, namely tundra and a warmer climate than present. In the base case negligibly low doses to man in the long term, resulting from the existence of a HLW repository have been calculated. Cs-135 results in the largest dose (8.4E-7 mrem/y at 6.1E+6 y) while Np-237 gives the largest dose from the actinides (3.6E-8 mrem/y). The response of the model to parameter variations cannot be easily predicted due to non-linear coupling of many of the parameters. However, the calculated doses were negligibly low in all cases as were those resulting from the two scenario variations. (author)

  19. Compas project stress analysis of HLW containers intermediate testwork

    Ove Arup and Partners London

    1990-01-01

    The Compas project is concerned with the structural performance of metal overpacks which may be used to encapsulate vitrified high-level waste forms before disposal in deep geological repositories. This document describes the series of experiments and associated calculations performed in the Intermediate testwork phase of this project. Seven mild steel, one-third scale simplified models of HLW containers were manufactured in a variety of configurations of geometry and weld type. The effects of reducing the wall thickness, corroding the external surface of the container, and using different welding methods were all investigated. The containers were tested under the action of a uniform external pressure up to their respective failure points. All containers failed by buckling at pressures of between 42 and 87 MPa dependent upon the particular geometric and weld configuration. The outer surface of each container was comprehensively strain-gauged in order to provide strain histories at positions of interest. The Compas project partners, from five different European countries, independently modelled the behaviour of three of the five different containers. Test results and computer predictions were compared and an assessment of the overall performance of the codes demonstrated good agreement in the initial loading of each container. However once stresses exceeded the material yield point there was a considerable spread in the predicted container behaviour

  20. Rheology of Savannah River Site Tank 51 HLW radioactive sludge

    Ha, B.C.

    1993-01-01

    Savannah River Site (SRS) Tank 51 HLW radioactive sludge represents a major portion of the first batch of sludge to be vitrified in the Defense Waste Processing Facility (DWPF) at SRS. The rheological properties of Tank 51 sludge will determine if the waste sludge can be pumped by the current DWPF process cell pump design and the homogeneity of melter feed slurries. The rheological properties of Tank 51 sludge and sludge/frit slurries at various solids concentrations were measured remotely in the Shielded Cells Operations (SCO) at the Savannah River Technology Center (SRTC) using a modified Haake Rotovisco viscometer system. Rheological properties of Tank 51 radioactive sludge/Frit 202 slurries increased drastically when the solids content was above 41 wt %. The yield stresses of Tank 51 sludge and sludge/frit slurries fall within the limits of the DWPF equipment design basis. The apparent viscosities also fall within the DWPF design basis for sludge consistency. All the results indicate that Tank 51 waste sludge and sludge/frit slurries are pumpable throughout the DWPF processes based on the current process cell pump design, and should produce homogeneous melter feed slurries

  1. Tc Chemistry in HLW: Role of Organic Complexants

    Hess, Nancy S.; Conradsen, Steven D.

    2003-01-01

    Tc complexation with organic compounds in tank waste plays a significant role in the redox chemistry of Tc and the partitioning of Tc between the supernatant and sludge components in waste tanks. These processes need to be understood so that strategies to effectively remove Tc from high-level nuclear waste prior to waste immobilization can be developed and so that long-term consequences of Tc remaining in residual waste after sludge removal can be evaluated. Only limited data on the stability of Tc-organic complexes exists and even less thermodynamic data on which to develop predictive models of Tc chemical behavior is available. To meet these challenges we are conducting a research program to study to develop thermodynamic data on Tc-organic complexation over a wide range of chemical conditions. We will attempt to characterize Tc-speciation in actual tank waste using state-of-the-art analytical organic chemistry, separations, and speciation techniques to validate our model. On the basis of such studies we will develop credible model of Tc chemistry in HLW that will allow prediction of Tc speciation in tank waste and Tc behavior during waste pretreatment processing and in waste tank residuals

  2. 'Practicality' as a key constraint to HLW repository design

    Kitayama, Kazumi; Sakabe, Yasushi; Ishiguro, Katsuhiko

    2007-01-01

    Designs of repositories in Japan for HLW have focused very much on demonstration of post-closure safety. Safety can be assured using very simple assessment techniques, which make many conservative simplifications. Such a situation is reasonable for the early stages of generic concept demonstration, but becomes less appropriate as NUMO moves towards siting, where a number of issues involved with construction and operation of a repository - generally grouped together as 'practicality'. The engineering logistics and conventional safety of repository construction and operation have been relatively little studied and present major challenges. Current designs emphasise a minimum of infrastructure in the emplacement tunnels and remote-handled operation. This would be difficult enough, but such operations need to be carried out to strict quality limits and need to be robust in the event of equipment failure or disruptive events. The paper will first examine how designs can be modified from the viewpoint of logistics. The implications of such modifications on operational robustness and associated safety in case of perturbation scenarios are then considered. (author)

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

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

    2011-10-17

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

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

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

    2011-01-01

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

  5. Survey of glass plutonium contents and poison selection

    Plodinec, M.J.; Ramsey, W.G. [Westinghouse Savannah River Company, Aiken, SC (United States); Ellison, A.J.G.; Shaw, H. [Lawrence Livermore National Laboratory, CA (United States)

    1996-05-01

    If plutonium and other actinides are to be immobilized in glass, then achieving high concentrations in the glass is desirable. This will lead to reduced costs and more rapid immobilization. However, glasses with high actinide concentrations also bring with them undersirable characteristics, especially a greater concern about nuclear criticality, particularly in a geologic repository. The key to achieving a high concentration of actinide elements in a glass is to formulate the glass so that the solubility of actinides is high. At the same time, the glass must be formulated so that the glass also contains neutron poisons, which will prevent criticality during processing and in a geologic repository. In this paper, the solubility of actinides, particularly plutonium, in three types of glasses are discussed. Plutonium solubilities are in the 2-4 wt% range for borosilicate high-level waste (HLW) glasses of the type which will be produced in the US. This type of glass is generally melted at relatively low temperatures, ca. 1150{degrees}C. For this melting temperature, the glass can be reformulated to achieve plutonium solubilities of at least 7 wt%. This low melting temperature is desirable if one must retain volatile cesium-137 in the glass. If one is not concerned about cesium volatility, then glasses can be formulated which can contain much larger amounts of plutonium and other actinides. Plutonium concentrations of at least 15 wt% have been achieved. Thus, there is confidence that high ({ge}5 wt%) concentrations of actinides can be achieved under a variety of conditions.

  6. Leaching behavior of simulated high-level waste glass

    Kamizono, Hiroshi

    1987-03-01

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

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

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

    2016-01-01

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

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

    Weber, W.J.; Ewing, R.C.

    1997-01-01

    A key challenge in the permanent disposal of high-level waste (HLW), plutonium residues/scraps, and excess weapons plutonium in glass waste forms is the development of predictive models of long-term performance that are based on a sound scientific understanding of relevant phenomena. Radiation effects from β-decay and α-decay can impact the performance of glasses for HLW and Pu disposition through the interactions of the α-particles, β-particles, recoil nuclei, and γ-rays with the atoms in the glass. Recently, a scientific panel convened under the auspices of the DOE Council on Materials Science to assess the current state of understanding, identify important scientific issues, and recommend directions for research in the area of radiation effects in glasses for HLW and Pu disposition. The overall finding of the panel was that there is a critical lack of systematic understanding on radiation effects in glasses at the atomic, microscopic, and macroscopic levels. The current state of understanding on radiation effects in glass waste forms and critical scientific issues are presented

  9. Clad Degradation - FEPs Screening Arguments

    E. Siegmann

    2004-01-01

    The purpose of this report is to document the screening of the cladding degradation features, events, and processes (FEPs) for commercial spent nuclear fuel (CSNF). This report also addresses the effect of some FEPs on both the cladding and the CSNF, DSNF, and HLW waste forms where it was considered appropriate to address the effects on both materials together. This report summarizes the work of others to screen clad degradation FEPs in a manner consistent with, and used in, the Total System Performance Assessment-License Application (TSPA-LA). This document was prepared according to ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA'' (BSC 2004a [DIRS 167796])

  10. Corrosion resistance of metal materials for HLW canister

    Furuya, Takashi; Muraoka, Susumu; Tashiro, Shingo

    1982-02-01

    In order to verify the materials as an important artificial barrier for canister of vitrified high-level waste from spent fuel reprocessing, data and reports were researched on corrosion resistance of the materials under conditions from glass form production to final disposal. Then, in this report, investigated subjects, improvement methods and future subjects are reviewed. It has become clear that there would be no problem on the inside and outside corrosion of the canister during glass production, but long term corrosion and radiation effect tests and the vitrification methods would be subjects in future on interim storage and final disposal conditions. (author)

  11. Novel waste forms for HLW and ILW immobilisation

    Lee, William E.; Milestone, Neil B.; Ojovan, Michael I.; Hyatt, Neil C.; Stennett, Martin C.; Setiadi, Anthony; Zhou, Qizhi

    2006-01-01

    The complex nature and heterogeneity of legacy wastes means that a toolbox of different host systems must be developed in which to immobilize them. New zirconolite ceramic, glass composite materials and novel cement systems including calcium sulpho aluminate cements and alkali activated slags being examined in the Immobilisation Science Laboratory at the University of Sheffield are described. (authors)

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

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

    2017-09-26

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

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

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

    2017-01-01

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

  14. Effect of Feed Melting, Temperature History and Minor Component Addition on Spinel Crystallization in High-Level Waste Glass

    Izak, Pavel; Hrma, Pavel R.; Arey, Bruce W.; Plaisted, Trevor J.

    2001-01-01

    This study was undertaken to help design mathematical models for high-level waste (HLW) glass melter that simulate spinel behavior in molten glass. Spinel, (Fe,Ni,Mn) (Fe,Cr)2O4, is the primary solid phase that precipitates from HLW glasses containing Fe and Ni in sufficient concentrations. Spinel crystallization affects the anticipated cost and risk of HLW vitrification. To study melting reactions, we used simulated HLW feed, prepared with co-precipitated Fe, Ni, Cr, and Mn hydroxides. Feed samples were heated up at a temperature-increase rate (4C/min) close to that which the feed experiences in the HLW glass melter. The decomposition, melting, and dissolution of feed components (such as nitrates, carbonates, and silica) and the formation of intermediate crystalline phases (spinel, sodalite (Na8(AlSiO4)6(NO2)2), and Zr-containing minerals) were characterized using evolved gas analysis, volume-expansion measurement, optical microscope, scanning electron microscope, thermogravimetric analysis, differential scanning calorimetry, and X-ray diffraction. Nitrates and quartz, the major feed components, converted to a glass-forming melt by 880C. A chromium-free spinel formed in the nitrate melt starting from 520C and Sodalite, a transient product of corundum dissolution, appeared above 600C and eventually dissolved in glass. To investigate the effects of temperature history and minor components (Ru,Ag, and Cu) on the dissolution and growth of spinel crystals, samples were heated up to temperatures above liquidus temperature (TL), then subjected to different temperature histories, and analyzed. The results show that spinel mass fraction, crystals composition, and crystal size depend on the chemical and physical makeup of the feed and temperature history

  15. Development of thermal analysis method for the near field of HLW repository using ABAQUS

    Kuh, Jung Eui; Kang, Chul Hyung; Park, Jeong Hwa [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    An appropriate tool is needed to evaluate the thermo-mechanical stability of high level radioactive waste (HLW) repository. In this report a thermal analysis methodology for the near field of HLW repository is developed to use ABAQUS which is one of the multi purpose FEM code and has been used for many engineering area. The main contents of this methodology development are the structural and material modelling to simulate a repository, setup of side conditions, e.g., boundary and load conditions, and initial conditions, and the procedure to selection proper material parameters. In addition to these, the interface programs for effective production of input data and effective change of model size for sensitivity analysis for disposal concept development are developed. The results of this work will be apply to evaluate the thermal stability and to use as main input data for mechanical analysis of HLW repository. (author). 20 refs., 15 figs., 5 tabs.

  16. Applicability of thermodynamic database of radioactive elements developed for the Japanese performance assessment of HLW repository

    Yui, Mikazu; Shibata, Masahiro; Rai, Dhanpat; Ochs, Michael

    2003-01-01

    In 1999 Japan Nuclear Cycle Development Institute (JNC) published a second progress report (also known as H12 report) on high-level radioactive waste (HLW) disposal in Japan (JNC 1999). This report helped to develop confidence in the selected HLW disposal system and to establish the implementation body in 2000 for the disposal of HLW. JNC developed an in-house thermodynamic database for radioactive elements for performance analysis of the engineered barrier system (EBS) and the geosphere for H12 report. This paper briefly presents the status of the JNC's thermodynamic database and its applicability to perform realistic analyses of the solubilities of radioactive elements, evolution of solubility-limiting solid phases, predictions of the redox state of Pu in the neutral pH range under reducing conditions, and to estimate solubilities of radioactive elements in cementitious conditions. (author)

  17. The Results of HLW Processing Using Zirconium Salt of Dibutyl phosphoric Acid in Hot Cell

    Fedorov, Yu.S.; Zilberman, B.Ya.; Shmidt, O.V. [Khlopin Radium Institute, 2nd Murinsky Ave., 28, Saint-Petersburg, 194021 (Russian Federation)

    2008-07-01

    Zirconium salt of dibutyl phosphoric acid (ZS HDBP), is an effective solvent for liquid HLW and ILW (high and intermediate level wastes) processing with radionuclide partitioning into different groups for further immobilization according to radiotoxicity. The rig trials in mixer-settles in hot cells were carried out using 30 L of real HLW containing transplutonium (TPE), rare earths (RE), Sr and Cs in 2 mol/L HNO{sub 3}, characterized by total specific activity 520 MBk/L. The recovery factor for TPE and RE was as high as 10{sup 4}, but only 10 for Sr. Purification factor of TPE and RE from Cs and Sr was 10{sup 4}, and that of Sr from TPE and Cs was 10{sup 3}. Almost all Cs was localized in the second cycle raffinate. So Zr salt of HDBP can be used in HLW processing with radionuclide partitioning with respect to the categories of radiotoxicity. (authors)

  18. Regulatory status on the safety assessment of a HLW repository in other countries

    Lee, Sung Ho; Hwang, Yong Soo

    2008-12-01

    To construct a HLW repository, it is essential to meet the requirements on the regulation for a deep geological disposal. Even if the construction of a HLW repository is determined positively, technical standards which assert the performance of a repository will be needed. Among various technical standards, safety assessment based on the repository evolution in the future will play an important role in the licensing process. The foreign countries' technical standards on the safety assessment of a HLW repository may be an indicator to carry out the R and D activities on geological disposal effectively. In this report, assessment period, limit of radiation dose and uncertainty related to the safety assessment are investigated and analyzed in detail. Especially, the technical reviews of USA regulation bodies seems to be reasonable in the point of the intrinsic attribute of safety assessment

  19. Study on evaluation method for potential effect of natural phenomena on a HLW disposal system

    Kawamura, Makoto; Makino, Hitoshi; Umeda, Koji; Osawa, Hideaki; Seo, Toshihiro; Ishimaru, Tsuneaki

    2005-01-01

    Evaluation for the potential effect of natural phenomena on a HLW disposal system is an important issue in safety assessment. A scenario construction method for the effects on a HLW disposal system condition and performance has been developed for two purposes: the first being effective elicitation and organization of information from investigators of natural phenomena and performance assessor and the second being, maintenance of traceability of scenario construction processes with suitable records. In this method, a series of works to construct scenarios is divided into pieces to facilitate and to elicit the features of potential effect of natural phenomena on a HLW disposal system and is organized to create reasonable scenarios with consistency, traceability and adequate conservativeness within realistic view. (author)

  20. Cold-Crucible Design Parameters for Next Generation HLW Melters

    Gombert, D.; Richardson, J.; Aloy, A.; Day, D.

    2002-01-01

    The cold-crucible induction melter (CCIM) design eliminates many materials and operating constraints inherent in joule-heated melter (JHM) technology, which is the standard for vitrification of high-activity wastes worldwide. The cold-crucible design is smaller, less expensive, and generates much less waste for ultimate disposal. It should also allow a much more flexible operating envelope, which will be crucial if the heterogeneous wastes at the DOE reprocessing sites are to be vitrified. A joule-heated melter operates by passing current between water-cooled electrodes through a molten pool in a refractory-lined chamber. This design is inherently limited by susceptibility of materials to corrosion and melting. In addition, redox conditions and free metal content have exacerbated materials problems or lead to electrical short-circuiting causing failures in DOE melters. In contrast, the CCIM design is based on inductive coupling of a water-cooled high-frequency electrical coil with the glass, causing eddycurrents that produce heat and mixing. A critical difference is that inductance coupling transfers energy through a nonconductive solid layer of slag coating the metal container inside the coil, whereas the jouleheated design relies on passing current through conductive molten glass in direct contact with the metal electrodes and ceramic refractories. The frozen slag in the CCIM design protects the containment and eliminates the need for refractory, while the corrosive molten glass can be the limiting factor in the JH melter design. The CCIM design also eliminates the need for electrodes that typically limit operating temperature to below 1200 degrees C. While significant marketing claims have been made by French and Russian technology suppliers and developers, little data is available for engineering and economic evaluation of the technology, and no facilities are available in the US to support testing. A currently funded project at the Idaho National Engineering

  1. Modelling spent fuel and HLW behaviour in repository conditions

    Esparza, A M; Esteban, J A

    2003-07-01

    The aim of this report is to give the reader an overall insight of the different models, which are used to predict the long-term behaviour of the spent fuels and HLW disposed in a repository. The models must be established on basic data and robust kinetics describing the mechanisms controlling spent fuel alteration/dissolution in a repository. The UO2 matrix, or source term, contains embedded in it the , majority of radionuclides of the spent fuel (some are in the gap cladding). For this reason the SF radionuclides release models play a significant role in the performance assessment of radioactive waste disposal. The differences existing between models published in the literature are due to the conceptual understanding of the processes and the degree of the conservatism used with the parameter values, and the boundary conditions. They mainly differ in their level of simplification and their final objective. Sometimes are focused the show compliance with regulatory requirements, other to support decision making, to increase the level of confidence of public and scientific community, could be empirical, semi-empirical or analytical. The models take into account the experimental results from radionuclides releases and their extrapolation to the very long term. Its necessary a great statistics for have a representative dissolution rate, due at the number of experimental results is not very high and many of them show a great scatter, independently of theirs different compositions by axial and radial variations, due to linear power or local burnup. On the other hand, it is difficult to predict the spent fuel behaviour over the long term, based in short term experiments. In this report is given a little description of the radionuclides distribution in the spent fuel and also in the cladding/pellet gap, grain boundary, cracks and rim zones (the matrix rim zone can be considered with an especial characteristics very different to the rest of the spent fuel), and structural

  2. Modelling spent fuel and HLW behaviour in repository conditions

    Esparza, A. M.; Esteban, J. A.

    2003-01-01

    The aim of this report is to give the reader an overall insight of the different models, which are used to predict the long-term behaviour of the spent fuels and HLW disposed in a repository. The models must be established on basic data and robust kinetics describing the mechanisms controlling spent fuel alteration/dissolution in a repository. The UO2 matrix, or source term, contains embedded in it the , majority of radionuclides of the spent fuel (some are in the gap cladding). For this reason the SF radionuclides release models play a significant role in the performance assessment of radioactive waste disposal. The differences existing between models published in the literature are due to the conceptual understanding of the processes and the degree of the conservatism used with the parameter values, and the boundary conditions. They mainly differ in their level of simplification and their final objective. Sometimes are focused the show compliance with regulatory requirements, other to support decision making, to increase the level of confidence of public and scientific community, could be empirical, semi-empirical or analytical. The models take into account the experimental results from radionuclides releases and their extrapolation to the very long term. Its necessary a great statistics for have a representative dissolution rate, due at the number of experimental results is not very high and many of them show a great scatter, independently of theirs different compositions by axial and radial variations, due to linear power or local burnup. On the other hand, it is difficult to predict the spent fuel behaviour over the long term, based in short term experiments. In this report is given a little description of the radionuclides distribution in the spent fuel and also in the cladding/pellet gap, grain boundary, cracks and rim zones (the matrix rim zone can be considered with an especial characteristics very different to the rest of the spent fuel), and structural

  3. Cesium and strontium fractionation from HLW for thermal-stress reduction in a geologic repository

    McKee, R.W.

    1983-02-01

    Results are described for a study to assess the benefits and costs of fractionating the cesium and strontium components in commercial high-level waste (HLW) to a separate waste stream for the purpose of reducing geologic repository thermal stresses. System costs are developed for a broad range of conditions comparing the Cs/Sr fractionation concept with disposal of 10-year old vitrified HLW and vitrified HLW aged to achieve (through decay) the same heat output as the fractionated high-level waste (FHLW). All comparisons are based on a 50,000 metric ton equivalent (MTE) system. The FHLW and the Cs/Sr waste are both disposed of a vitrified waste but emplaced in separate areas of a basalt repository. The FHLW is emplaced in high-integrity packages at relatively high waste loading but low heat loading, while the Cs/Sr waste is emplaced in minimum integrity packages at relatively high heat loading. System cost comparisons are based on minimum cost combinations of canister diameter, waste concentration, and canister spacing in a basalt repository for each waste type. The effects on both long- and near-term safety considerations are also addressed. The major conclusion is that the Cs/Sr fractionation concept offers, potentially, a substantial total system cost advantage for HLW disposal if reduced HLW package temperatures in a basalt repository are desired. However, there is no cost advantage if currently designated maximum design temperatures are acceptable. Aging the HLW for 50 to 100 years can accomplish similar results at equivalent or loser costs

  4. Glass sealing

    Brow, R.K.; Kovacic, L.; Chambers, R.S. [Sandia National Labs., Albuquerque, NM (United States)

    1996-04-01

    Hernetic glass sealing technologies developed for weapons component applications can be utilized for the design and manufacture of fuel cells. Design and processing of of a seal are optimized through an integrated approach based on glass composition research, finite element analysis, and sealing process definition. Glass sealing procedures are selected to accommodate the limits imposed by glass composition and predicted calculations.

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

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

    1996-02-01

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

  6. Electrochromic Glasses.

    1980-07-31

    this glass and that dipole-dipole correlations contribute to the "ferroelectric-like" character of this amorphous system. The TeO2 -W03 glasses can only...shows the dielectric constant and Fig. I(b) glass from pure TeO2 ot pure WO. In addition, glass the tan 8 of the WO glass as a function of temperature... glasses containing WO, in various glass forming nitworks of LifO-B1O0, Na:O-BzO,, and TeO2 were prepared from reagent grade oxides at 800 C - 9SO C in

  7. Processes for consensus building and role sharing. Lessons learned from HLW policies in European countries

    Nagano, Koji

    2003-01-01

    This report attempts to obtain lessons in implementation of HLW management policies for Japan by reviewing past experiences and present status of policy formulation and implementation as well as reflection of public opinions and consensus building of selected European countries, such as Finland, Sweden and others. After examining the situations of those countries, the author derives four key aspects that need to be addressed; separation of nuclear energy policies and HLW policies, fundamental support shared among national public, sense of controllability, and proper scheme of responsibility sharing. (author)

  8. Fabrication of solar light induced Fe-TiO{sub 2} immobilized on glass-fiber and application for phenol photocatalytic degradation

    Lin, Shaohua, E-mail: linsh75@163.com [School of Civil Engineering, Nanjing Forestry University, Nanjing, Jiangsu Province 210037 (China); Zhang, Xiwang [School of Applied Sciences and Engineering, Monash University Gippsland Campus, Churchill, Victoria 3842 (Australia); Sun, Qinju; Zhou, Tingting; Lu, Jingjing [School of Civil Engineering, Nanjing Forestry University, Nanjing, Jiangsu Province 210037 (China)

    2013-11-15

    Graphical abstract: - Highlights: • Fe-doped TiO{sub 2} immobilized on glass-fiber net were prepared by sol–gel method. • Fe inhibited the phase transition of TiO{sub 2} from anatase to rutile. • The optimal Fe doping dose was around 0.005 wt%. • The optimal calcination temperature was around 600 °C. - Abstract: Iron-doped anatase titanium dioxide catalysts coated on glass-fiber were successfully synthesized by a dip-coating sol–gel method. The prepared catalysts were characterized by scanning electron microscopy (SEM) with energy dispersive X-ray (EDX) analysis, X-ray diffraction (XRD), UV-Vis diffuse reflectance spectroscopy to understand the synthesis mechanism, and their photocatalytic activities were evaluated by photodegradation of phenol under simulated solar irradiation. EDX analysis confirmed the existence of iron in the immobilized catalysts. XRD suggested that the phase transition of the catalysts from anatase to rutile were restrained, and almost pure anatase TiO{sub 2} could retain even the calcination temperature reached 800 °C. The UV-Vis diffuse reflectance spectroscopy of the catalysts showed a red shift and increased photoabsorbance in the visible range for all the doped samples. Iron loading and calcination temperature have obvious influences on photocatalytic activity. In this study, the optimal doping dose and calcination temperature were around 0.005 wt% and 600 °C, respectively.

  9. Development of methodology to construct a generic conceptual model of river-valley evolution for performance assessment of HLW geological disposal

    Kawamura, Makoto; Tanikawa, Shin-ichi; Yasue, Ken-ichi; Niizato, Tadafumi

    2011-01-01

    In order to assess the long-term safety of a geological disposal system for high-level radioactive waste (HLW), it is important to consider the impact of uplift and erosion, which cannot be precluded on a timescale in the order of several hundred thousand years for many locations in Japan. Geomorphic evolution, caused by uplift and erosion and coupled to climatic and sea-level changes, will impact the geological disposal system due to resulting spatial and temporal changes in the disposal environment. Degradation of HLW barrier performance will be particularly significant when the remnant repository structures near, and are eventually exposed at, the ground surface. In previous studies, fluvial erosion was densified as the key concern in most settings in Japan. Interpretation of the impact of the phenomena at relevant locations in Japan has led to development of a generic conceptual model which contains the features typical at middle reach of rivers. Here, therefore, we present a methodology for development of a generic conceptual model based on best current understanding of fluvial erosion in Japan, which identifies the simplifications and uncertainties involved and assesses their consequences in the context of repository performance. (author)

  10. Cobalt ferrite nano-composite coated on glass by Doctor Blade method for photo-catalytic degradation of an azo textile dye Reactive Red 4: XRD, FESEM and DRS investigations.

    Habibi, Mohammad Hossein; Parhizkar, Janan

    2015-11-05

    Cobalt ferrite nano-composite was prepared by hydrothermal route using cobalt nitrate, iron nitrate and ethylene glycol as chelating agent. The nano-composite was coated on glass by Doctor Blade method and annealed at 300 °C. The structural, optical, and photocatalytic properties have been studied by powder X-ray diffraction (XRD), field emission scanning electron microscopy (FESEM) and UV-visible spectroscopy (UV-Vis DRS). Powder XRD analysis confirmed formation of CoFe2O4 spinel phase. The estimated particle size from FESEM data was 50 nm. The calculated energy band gaps, obtained by Tauc relation from UV-Vis absorption spectra was 1.3 eV. Photocatalytic degradation of Reactive Red 4 as an azo textile was investigated in aqueous solution under irradiation showed 68.0% degradation of the dye within 100 min. The experimental enhanced activity compare to pure Fe2O3 can be ascribed to the formation of composite, which was mainly attributable to the transfer of electron and hole to the surface of composite and hinder the electron hole recombination. Copyright © 2015 Elsevier B.V. All rights reserved.

  11. Bioactive glass in tissue engineering

    Rahaman, Mohamed N.; Day, Delbert E.; Bal, B. Sonny; Fu, Qiang; Jung, Steven B.; Bonewald, Lynda F.; Tomsia, Antoni P.

    2011-01-01

    This review focuses on recent advances in the development and use of bioactive glass for tissue engineering applications. Despite its inherent brittleness, bioactive glass has several appealing characteristics as a scaffold material for bone tissue engineering. New bioactive glasses based on borate and borosilicate compositions have shown the ability to enhance new bone formation when compared to silicate bioactive glass. Borate-based bioactive glasses also have controllable degradation rates, so the degradation of the bioactive glass implant can be more closely matched to the rate of new bone formation. Bioactive glasses can be doped with trace quantities of elements such as Cu, Zn and Sr, which are known to be beneficial for healthy bone growth. In addition to the new bioactive glasses, recent advances in biomaterials processing have resulted in the creation of scaffold architectures with a range of mechanical properties suitable for the substitution of loaded as well as non-loaded bone. While bioactive glass has been extensively investigated for bone repair, there has been relatively little research on the application of bioactive glass to the repair of soft tissues. However, recent work has shown the ability of bioactive glass to promote angiogenesis, which is critical to numerous applications in tissue regeneration, such as neovascularization for bone regeneration and the healing of soft tissue wounds. Bioactive glass has also been shown to enhance neocartilage formation during in vitro culture of chondrocyte-seeded hydrogels, and to serve as a subchondral substrate for tissue-engineered osteochondral constructs. Methods used to manipulate the structure and performance of bioactive glass in these tissue engineering applications are analyzed. PMID:21421084

  12. Colloid formation during waste glass corrosion

    Mertz, C.J.; Buck, E.C.; Fortner, J.A.; Bates, J.K.

    1996-01-01

    The long-term behavior of nuclear waste glass in a geologic repository may require a technical consideration of the role of colloids in the release and transport of radionuclides. The neglect of colloidal properties in assessing the near- and far-field migration behavior of actinides may lead to significant underestimates and poor predictions of biosphere exposure from high-level waste (HLW) disposal. Existing data on colloid-facilitated transport suggests that radionuclide migration may be enhanced, but the importance of colloids is not adequately assessed. Indeed, the occurrence of radionuclide transport, attributed to colloidal species, has been reported at Mortandad Canyon, Los Alamos and at the Nevada Test Site; both unsaturated regions are similar to the proposed HLW repository at Yucca Mountain. Although some developments have been made on understanding the transport characteristics of colloids, the characterization of colloids generated from the corrosion of the waste form has been limited. Colloids are known to incorporate radionuclides either from hydrolysis of dissolved species (real colloids) or from adsorption of dissolved species onto existing groundwater colloids (pseudocolloids); however, these colloids may be considered secondary and solubility limited when compared to the colloids generated during glass alteration

  13. Key Factors to Determine the Borehole Spacing in a Deep Borehole Disposal for HLW

    Lee, Jongyoul; Choi, Heuijoo; Lee, Minsoo; Kim, Geonyoung; Kim, Kyeongsoo

    2015-01-01

    Deep fluids also resist vertical movement because they are density stratified and reducing conditions will sharply limit solubility of most dose critical radionuclides at the depth. Finally, high ionic strengths of deep fluids will prevent colloidal transport. Therefore, as an alternative disposal concept, i.e., deep borehole disposal technology is under consideration in number of countries in terms of its outstanding safety and cost effectiveness. In this paper, the general concept for deep borehole disposal of spent fuels or high level radioactive wastes which has been developed by some countries according to the rapid advance in the development of drilling technology, as an alternative method to the deep geological disposal method, was reviewed. After then an analysis on key factors for the distance between boreholes for the disposal of HLW was carried out. In this paper, the general concept for deep borehole disposal of spent fuels or HLW wastes, as an alternative method to the deep geological disposal method, were reviewed. After then an analysis on key factors for the determining the distance between boreholes for the disposal of HLW was carried out. These results can be used for the development of the HLW deep borehole disposal system

  14. Key Factors to Determine the Borehole Spacing in a Deep Borehole Disposal for HLW

    Lee, Jongyoul; Choi, Heuijoo; Lee, Minsoo; Kim, Geonyoung; Kim, Kyeongsoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Deep fluids also resist vertical movement because they are density stratified and reducing conditions will sharply limit solubility of most dose critical radionuclides at the depth. Finally, high ionic strengths of deep fluids will prevent colloidal transport. Therefore, as an alternative disposal concept, i.e., deep borehole disposal technology is under consideration in number of countries in terms of its outstanding safety and cost effectiveness. In this paper, the general concept for deep borehole disposal of spent fuels or high level radioactive wastes which has been developed by some countries according to the rapid advance in the development of drilling technology, as an alternative method to the deep geological disposal method, was reviewed. After then an analysis on key factors for the distance between boreholes for the disposal of HLW was carried out. In this paper, the general concept for deep borehole disposal of spent fuels or HLW wastes, as an alternative method to the deep geological disposal method, were reviewed. After then an analysis on key factors for the determining the distance between boreholes for the disposal of HLW was carried out. These results can be used for the development of the HLW deep borehole disposal system.

  15. A GoldSim Based Biosphere Assessment Model for a HLW Repository

    Lee, Youn-Myoung; Hwang, Yong-Soo; Kang, Chul-Hyung

    2007-01-01

    To demonstrate the performance of a repository, the dose exposure to a human being due to nuclide releases from a repository should be evaluated and the results compared to the dose limit presented by the regulatory bodies. To evaluate a dose rate to an individual due to a long-term release of nuclides from a HLW repository, biosphere assessment models and their implemented codes such as ACBIO1 and ACBIO2 have been developed with the aid of AMBER during the last few years. BIOMASS methodology has been adopted for a HLW repository currently being considered in Korea, which has a similar concept to the Swedish KBS-3 HLW repository. Recently, not just only for verifying the purpose for biosphere assessment models but also for varying the possible alternatives to assess the consequences in a biosphere due to a HLW repository, another version of the assessment modesl has been newly developed in the frame of development programs for a total system performance assessment modeling tool by utilizing GoldSim. Through a current study, GoldSim approach for a biosphere modeling is introduced. Unlike AMBER by which a compartment scheme can be rather simply constructed with an appropriate transition rate between compartments, GoldSim was designed to facilitate the object-oriented modules by which specific models can be addressed in an additional manner, like solving jig saw puzzles

  16. Design options for HLW repository operation technology. (4) Shotclay technique for seamless construction of EBS

    Kobayashi, Ichizo; Fujisawa, Soh; Nakajima, Makoto; Toida, Masaru; Nakashima, Hitoshi; Asano, Hidekazu

    2011-01-01

    The shotclay method is construction method of the high density bentonite engineered barrier by spraying method. Using this method, the dry density of 1.6 Mg/m 3 , which was considered impossible with the spray method, is achieved. In this study, the applicability of the shotclay method to HLW bentonite-engineered barriers was confirmed experimentally. In the tests, an actual scale vertical-type HLW bentonite-engineered barrier was constructed. This was a bentonite-engineered barrier with a diameter of 2.22 m and a height of 3.13 m. The material used was bentonite with 30% silica sand, and water content was adjusted by mixing chilled bentonite with powdered ice before thawing. Work progress was 11.2 m 3 and the weight was 21.7 Mg. The dry density of the entire buffer was 1.62 Mg/m 3 , and construction time was approximately 8 hours per unit. After the formworks were removed, the core and block of the actual scale HLW bentonite-engineered barrier were sampled to confirm homogeneity. As a result, homogeneity was confirmed, and no gaps were observed between the formwork and the buffer material and between the simulated waste and the buffer material. The applicability to HLW of the shotclay method has been confirmed through this examination. (author)

  17. Disposal of defense spent fuel and HLW from the Idaho Chemical Processing Plant

    Ermold, L.F.; Loo, H.H.; Klingler, R.D.; Herzog, J.D.; Knecht, D.A.

    1992-12-01

    Acid high-level radioactive waste (HLW) resulting from fuel reprocessing at the Idaho Chemical Processing Plant (ICPP) for the US Department of Energy (DOE) has been solidified to a calcine since 1963 and stored in stainless steel bins enclosed by concrete vaults. Several different types of unprocessed irradiated DOE-owned fuels are also in storage ate the ICPP. In April, 1992, DOE announced that spent fuel would no longer be reprocessed to recover enriched uranium and called for a shutdown of the reprocessing facilities at the ICPP. A new Spent Fuel and HLW Technology Development program was subsequently initiated to develop technologies for immobilizing ICPP spent fuels and HLW for disposal, in accordance with the Nuclear Waste Policy Act. The Program elements include Systems Analysis, Graphite Fuel Disposal, Other Spent Fuel Disposal, Sodium-Bearing Liquid Waste Processing, Calcine Immobilization, and Metal Recycle/Waste Minimization. This paper presents an overview of the ICPP radioactive wastes and current spent fuels, with an emphasis on the description of HLW and spent fuels requiring repository disposal

  18. HLW Salt Disposition Alternatives Identification Preconceptual Phase I Summary Report (Including Attachments)

    Piccolo, S.F.

    1999-01-01

    The purpose of this report is to summarize the process used by the Team to systematically develop alternative methods or technologies for final disposition of HLW salt. Additionally, this report summarizes the process utilized to reduce the total list of identified alternatives to an ''initial list'' for further evaluation. This report constitutes completion of the team charter major milestone Phase I Deliverable

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

    McKenzie, W.F.

    1990-01-01

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

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

    McKenzie, W.F.

    1990-01-01

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

  1. Radiation effects on transport and bubble formation in silicate glasses. 1998 annual progress report

    Trifunac, A.D.

    1998-01-01

    'To study the fundamental chemistry of radiation damage in silicate/borosilicate glasses and simulated high-level nuclear waste (HLW) forms. Special emphasis is on delineating molecular processes crucial for understanding the aggregation of defects and formation of oxygen bubbles. The knowledge obtained will provide the needed scientific basis for extrapolating long-term behavior of stored radiative waste glass forms. This report summarizes the first 6 months of a 3-year project. The following issues have been addressed: (i) the production of radiolytic oxygen, (ii) the chemistry of hydrogenous species, and (iii) the effect of glass composition and microstructure on the formation and accumulation of metastable point defects.'

  2. Glass consistency and glass performance

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

    1994-01-01

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

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

    Jain, V.; Pan, Y.M.

    2004-01-01

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

  4. Colloidal glasses

    First page Back Continue Last page Overview Graphics. Colloidal glasses. Glassy state is attained when system fails to reach equilibrium due to crowding of constituent particles. In molecular glasses, glassy state is reached by rapidly lowering the temperature. In colloidal glasses, glassy state is reached by increasing the ...

  5. Retention of Halogens in Waste Glass

    Hrma, Pavel R.

    2010-05-01

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

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

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

    2017-11-01

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

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

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

    1995-03-01

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

  8. High-level waste borosilicate glass a compendium of corrosion characteristics. Volume 1

    Cunnane, J.C.

    1994-03-01

    Current plans call for the United States Department of Energy (DOE) to start up facilities for vitrification of high-level radioactive waste (HLW) stored in tanks at the Savannah River Site, Aiken, South Carolina, in 1995; West Valley Demonstration Project, West Valley, New York, in 1996; and at the Hanford Site, Richland, Washington, after the year 2000. The product from these facilities will be canistered HLW borosilicate glass, which will be stored, transported, and eventually disposed of in a geologic repository. The behavior of this glass waste product, under the range of likely service conditions, is the subject of considerable scientific and public interest. Over the past few decades, a large body of scientific information on borosilicate waste glass has been generated worldwide. The intent of this document is to consolidate information pertaining to our current understanding of waste glass corrosion behavior and radionuclide release. The objective, scope, and organization of the document are discussed in Section 1.1, and an overview of borosilicate glass corrosion is provided in Section 1.2. The history of glass as a waste form and the international experience with waste glass are summarized in Sections 1.3 and 1.4, respectively

  9. High Level Waste (HLW) Processing Experience with Increased Waste Loading

    JANTZEN, CAROL

    2004-01-01

    The Defense Waste Processing Facility (DWPF) Engineering requested characterization of glass samples that were taken after the second melter had been operational for about 5 months. After the new melter had been installed, the waste loading had been increased to about 38 weight percentage after a new quasicrystalline liquidus model had been implemented. The DWPF had also switched from processing with refractory Frit 200 to a more fluid Frit 320. The samples were taken after DWPF observed very rapid buildup of deposits in the upper pour spout bore and on the pour spout insert while processing the high waste loading feedstock. These samples were evaluated using various analytical techniques to determine the cause of the crystallization. The pour stream sample was homogeneous, amorphous, and representative of the feed batch from which it was derived. Chemical analysis of the pour stream sample indicated that a waste loading of 38.5 weight per cent had been achieved. The data analysis indicated that surface crystallization, induced by temperature and oxygen fugacity gradients in the pour spout, caused surface crystallization to occur in the spout and on the insert at the higher waste loadings even though there was no crystallization in the pour stream

  10. Concept of grouping in partitioning of HLW for self-consistent fuel cycle

    Kitamoto, A.; Mulyanto

    1993-01-01

    A concept of grouping for partitioning of HLW has been developed in order to examine the possibility of a self-consistent fuel recycle. The concept of grouping of radionuclides is proposed herein, such as Group MA1 (MA below Cm), Group MA2 (Cm and higher MA), Group A ( 99 Tc and I), Group B (Cs and Sr) and Group R (the partitioned remain of HLW). Group B is difficult to be transmuted by neutron reaction, so a radiation application in an industrial scale should be developed in the future. Group A and Group MA1 can be burned by a thermal reactor, on the other hand Group MA2 should be burned by a fast reactor. P-T treatment can be optimized for the in-core and out-core system, respectively

  11. Compas project stress analysis of HLW containers: behaviour under realistic disposal conditions

    Ove Arup and Partners, London

    1990-01-01

    The Compas project is concerned with the structural performance of metal overpacks which may be used to encapsulate vitrified high-level waste (HLW) forms before disposal in deep geological repositories. In this final stage of the project, analysis of an HLW overpack of realistic design is performed to predict its behaviour when subjected to likely repository loads. This analysis work is undertaken with the benefit of experience gained in previous phases of the project in which the ability to accurately predict overpack behaviour, when subjected to a uniform external pressure, was demonstrated. Burial in clay, granite and salt environments has been considered and two distinct loading arrangements identified, in an attempt to represent the worst conditions that could be imposed by such media. The analysis successfully demonstrates the ability of the containers to withstand extreme, yet credible, repository loads

  12. Study on the properties of Gaomiaozi bentonite as the buffer/backfilling materials for HLW disposal

    Liu Xiaodong; Luo Taian; Zhu Guoping; Chen Qingchun

    2007-12-01

    Systematic studies including mineral composition and structure, physico- chemical properties and thermal properties have been conducted on Gaomiaozi bentonite, Xinghe County, Inner Mongolia Autonomous Region. The compaction characteristics of bentonite and the influence of additive to bentonite have been discussed. The analysis of mineral composition and structure show that the bentonite ores are dominated by montmorillonite. Preliminary studies of the characteristics of ores indicated that No-type bentonite from the deposit has good absorption, excellent swelling and high cation exchangeability. The compressibility of bentonite will be improved by adding the additives such as quartz sand. The studies indicated that the characteristics of Gaomiaozi bentonite can satisfy the requirement of buffer/backfilling materials for HLW repository and the ores can be selected as the preferential candidate to provide buffer/backfill- ing materials for HLW repository in China. (authors)

  13. The interpretation of remote sensing image on the stability of fault zone at HLW repository site

    Liu Linqing; Yu Yunxiang

    1994-01-01

    It is attempted to interpret the buried fault at the preselected HLW repository site in western Gansu province with a remote sensing image. The authors discuss the features of neotectonism of Shule River buried fault zone and its two sides in light of the remote sensing image, geomorphology, stream pattern, type and thickness difference of Quaternary sediments, and structural basin, etc.. The stability of Shule River fault zone is mainly dominated by the neotectonic movement pattern and strength of its two sides. Although there exist normal and differential vertical movements along it, their strengths are small. Therefore, this is a weakly-active passive fault zone. The east Beishan area north to Shule River fault zone is weakliest active and is considered as the target for further pre-selection for HLW repository site

  14. Study on the properties of Gaomiaozi bentonite as the buffer/backfilling materials for HLW disposal

    Xiaodong, Liu [East China Inst. of Technology, Fuzhou (China); [Key Laboratory of Nuclear Resources and Environment of Ministry of Education, Fuzhou (China); Taian, Luo; Guoping, Zhu; Qingchun, Chen [East China Inst. of Technology, Fuzhou (China)

    2007-12-15

    Systematic studies including mineral composition and structure, physico- chemical properties and thermal properties have been conducted on Gaomiaozi bentonite, Xinghe County, Inner Mongolia Autonomous Region. The compaction characteristics of bentonite and the influence of additive to bentonite have been discussed. The analysis of mineral composition and structure show that the bentonite ores are dominated by montmorillonite. Preliminary studies of the characteristics of ores indicated that No-type bentonite from the deposit has good absorption, excellent swelling and high cation exchangeability. The compressibility of bentonite will be improved by adding the additives such as quartz sand. The studies indicated that the characteristics of Gaomiaozi bentonite can satisfy the requirement of buffer/backfilling materials for HLW repository and the ores can be selected as the preferential candidate to provide buffer/backfill- ing materials for HLW repository in China. (authors)

  15. Current status and future plans of R and D on geological disposal of HLW in Japan

    Sasaki, Noriaki

    1994-01-01

    As to the final disposal of HLW, it is considered highly important to provide a clear distinction between implementation of disposal and the research and development as independent processes, and to increase the transparency of the overall disposal program by defining concrete schedules and the roles and responsibilities of the organizations involved. The Power Reactor and Nuclear Fuel Development Corporation (PNC) has being conducted research and development on the geological disposal of HLW, as the leading organization. The responsibility of PNC is to ensure smooth progress of research and development project and to carry out studies of geological environment. The role of the Japanese government is to take overall responsibilities for appropriate and steady implementations of the program, as well as enacting any laws or policies required. On the other hand, electricity supply utilities are responsible to secure necessary funds for disposal, and in accordance with their role as waste producers, they are expected to cooperate even at the stage of research and development. Fundamental features of research and development of PNC carried out at this stage are as follows; (1) Generic research and development, (2) To establish scientific and technical bases of geological isolation of HLW in Japan, (3) About 15 years program from 1989 with documentation of progress reports, (4) Approach from near-field to far-field. PNC summarized the findings obtained by 1991, and submitted a document (H3 Report) in September 1992 as the first progress report. H3 Report is the first and comprehensive technical report on geological disposal of HLW in Japan, and provides information for the public to find out the current status of the research and development. This paper reviews the conclusions of H3 Report, overall procedures and schedule for implementing geological disposal, and future plans of R and D in PNC. (J.P.N.)

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

    Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Mcclane, D. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-01

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

  17. Chemical composition analysis and product consistency tests supporting refinement of the Nepheline model for the high aluminum Hanford Glass composition region

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

    2016-02-17

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

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

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

    1991-01-01

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

  19. Development of a Korean Reference disposal System(A-KRS) for the HLW from Advanced Fuel Cycles

    Choi, Heui Joo; Choi, J. W.; Lee, J. Y.

    2010-04-01

    A database program for analyzing the characteristics of spent fuels was developed, and A-SOURCE program for characterizing the source term of HLW from advanced fuel cycles. A new technique for developing a copper canister by introducing a cold spray technique was developed, which could reduce the amount of copper. Also, to enhance the performance of A-KRS, two kinds of properties, thermal performance and iodine adsorption, were studied successfully. A complex geological disposal system which can accommodate all the HLW (CANDU and HANARO spent fuels, HLW from pyro-processing of PWR spent fuels, decommissioning wastes) was developed, and a conceptual design was carried out. Operational safety assessment system was constructed for the long-term management of A-KRS. Three representative accidental cases were analyzed, and the probabilistic safety assessment was adopted as a methodology for the safety evaluation of A-KRS operation. A national program was proposed to support the HLW national policy on the HLW management. A roadmap for HLW management was proposed based on the optimum timing of disposal

  20. Recycle Glass in Foam Glass Production

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

    The foam glass industry turn recycle glass into heat insulating building materials. The foaming process is relative insensitive to impurities in the recycle glass. It is therefore considered to play an important role in future glass recycling. We show and discuss trends of use of recycled glasses...... in foam glass industry and the supply sources and capacity of recycle glass....

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

    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

  2. The use of mineral-like matrices for hlw solidification and spent fuel immobilization

    Pokhitonov, J.A.; Starchenko, V.A.; Strelnikov, A.V.; Sorokin, V.T.; Shvedov, A.A.

    2000-01-01

    The conception of radioactive waste management is based upon the multi-barrier protection principle stating that the long-lived radionuclides safety isolation is ensured by a system of engineering and natural geological barriers. One of the effective ways of the long-lived radionuclides immobilization is the integration of these materials within a mineral-like matrice. This technique may be used both for isolation of separated groups of nuclides (Cs, Sr, TUE, TRE) and for immobilization of spent fuel which for some reason can't be processed at the radiochemical plant. In this paper two variants of flowsheets HLW management are discussed. The following ways of HLW reprocessing are considered: - The first cycle raffinate solidification (without partitioning); - The individual solidification of two separated radionuclide groups (Sr+Cs+FP fraction and TPE+TRE fraction). The calcination of some characteristics (annual and total amounts, specific activity, radiochemical composition and radiogenic heat) of HLW integrated within a mineral-like matrix are performed for both options. The matrix compositions may be also used for spent fuel immobilization by means of the hot isostatic pressing technique. (authors)

  3. Researches on tectonic uplift and denudation with relation to geological disposal of HLW in Japan

    Fujiwara, Osamu; Sanga, Tomoji; Moriya, Toshifumi

    2005-01-01

    This paper reviews the present state of researches on tectonic uplift and denudation, and shows perspective goals and direction of future researches from the viewpoint of geological disposal of HLW in Japan. Detailed history of tectonics and denudation in geologic time scale, including the rates, temporal and spatial distributions and processes, reconstructed from geologic and geomorphologic evidences will enable us to make the geological predictions. Improvements of the analytic methods for the geological histories, e.g. identification of the tectonic and denudational imprints and age determinations, are indispensable for the accurate prediction. Developments of the tools and methodologies for assessments of the degree and extension of influences by the tectonic uplift, subsidence and denudation on the geological environments such as ground water flows are also fundamental problem in the study field of the geological disposal of HLW. Collaboration of scientific researches using the geological and geomorphological methods and applied technology, such as numerical simulations of ground water flows, is important in improving the safety and accuracy of the geological disposal of HLW. (author)

  4. The senate working party on HLW management in Spain - historical perspective

    Lang-Lenton, J.

    2007-01-01

    As the first case history Jorge Lang Lenton, Corporate Director of ENRESA, recounted the failed attempt to establish an underground disposal facility for HLW. The site selection process, which was planned by ENRESA in the 1980's, was aimed at finding the 'technically best' site. The process was conducted by technical experts without public involvement. When 40 candidate siting areas were identified in the mid-1990's, information leaked out, creating vigorous public opposition in all of these locations. In 1998 the siting process was halted. The Senate proposed to continue R and D on geological disposal and on P and T, to reduce waste production, and to develop an energy policy that relies more on renewable energy sources. They also suggested that public participation be promoted. The 5. General Radioactive Waste Management Plan, which was developed in 1999, took these proposals into consideration. Regarding underground disposal, the government postponed any decision until 2010. At the end of 2004 a decision was made by Parliament to establish a centralized storage facility for HLW. Mr. Lang-Lenton highlighted the main lessons of the failed siting attempt. First, it has to be acknowledged that HLW management is a societal rather than a technical problem. Second, for any radioactive waste management facility a socially feasible rather than a technically optimal site should be selected, i.e., 'the best site is the possible site'. Finally, transparency and openness are needed for building confidence in the decision-making process. (author)

  5. Accelerated Leach Testing of GLASS: ALTGLASS Version 3.0

    Trivelpiece, Cory L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jantzen, Carol M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, Charles L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-12-31

    The Accelerated Leach Testing of GLASS (ALTGLASS) database is a collection of data from short- and long-term product consistency tests (PCT, ASTM C1285 A and B) on high level waste (HLW) as well as low activity waste (LAW) glasses. The database provides both U.S. and international researchers with an archive of experimental data for the purpose of studying, modeling, or validating existing models of nuclear waste glass corrosion. The ALTGLASS database is maintained and updated by researchers at the Savannah River National Laboratory (SRNL). This newest version, ALTGLASS Version 3.0, has been updated with an additional 503 rows of data representing PCT results from corrosion experiments conducted in the United States by the Savannah River National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and the Vitreous State Laboratory (SRNL, PNNL, ANL, VSL, respectively) as well as the National Nuclear Laboratory (NNL) in the United Kingdom.

  6. Accelerated Leach Testing of GLASS: ALTGLASS Version 3.0

    Trivelpiece, Cory L.; Jantzen, Carol M.; Crawford, Charles L.

    2016-01-01

    The Accelerated Leach Testing of GLASS (ALTGLASS) database is a collection of data from short- and long-term product consistency tests (PCT, ASTM C1285 A and B) on high level waste (HLW) as well as low activity waste (LAW) glasses. The database provides both U.S. and international researchers with an archive of experimental data for the purpose of studying, modeling, or validating existing models of nuclear waste glass corrosion. The ALTGLASS database is maintained and updated by researchers at the Savannah River National Laboratory (SRNL). This newest version, ALTGLASS Version 3.0, has been updated with an additional 503 rows of data representing PCT results from corrosion experiments conducted in the United States by the Savannah River National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and the Vitreous State Laboratory (SRNL, PNNL, ANL, VSL, respectively) as well as the National Nuclear Laboratory (NNL) in the United Kingdom.

  7. Direct conversion of surplus fissile materials, spent nuclear fuel, and other materials to high-level-waste glass

    Forsberg, C.W.; Elam, K.R.

    1995-01-01

    With the end of the cold war the United States, Russia, and other countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. The United States Academy of Sciences (NAS) has recommended that these surplus fissile materials (SFMs) be processed so they are no more accessible than plutonium in spent nuclear fuel (SNF). This spent fuel standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. The NAS recommended investigation of three sets of options for disposition of SFMs while meeting the spent fuel standard: (1) incorporate SFMs with highly radioactive materials and dispose of as waste, (2) partly burn the SFMs in reactors with conversion of the SFMs to SNF for disposal, and (3) dispose of the SFMs in deep boreholes. The US Government is investigating these options for SFM disposition. A new method for the disposition of SFMs is described herein: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptinium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal

  8. Effects of solar radiation on glass

    Tucker, Dennis S.; Kinser, Donald L.

    1991-01-01

    The effects of solar radiation of selected glasses are reported. Optical property degradation is studied using UV-Vis spectrophotometry. Strength changes are measured using a concentric ring bend test. Direct fracture toughness measurements using an indentation test are planned.

  9. The Effects of Oxygen Partial Pressure on Liquidus Temperature of a High-Level Waste Glass with Spinel as the Primary Phase

    Izak, Pavel; Hrma, Pavel R.; Wilson, Benjamin K.; Vienna, John D.

    2000-01-01

    The redox state of iron affects spinal crystallization in vitrified high-level waste (HLW) glass. Simulated HLW glass with spinel as the primary crystalline phase field was heat treated at constant temperatures within the interval from 850 C to 1300 C under varying atmospheres with oxygen partial pressure, Po2, ranging from 1x10-16 kPa (pure CO) to 101 kPa (pure O2). Liquidus temperature (TL) of glass increased with decreasing Po2 up to Po2 > 3 x 10-9 kPa. At Po2 < 3 x 10-9 kPa, Ni-Fe alloy precipitated from the glass, and TL decreased. Samples were analyzed with optical microscope and scanning electron microscope. The mass fraction of spinel in glass was determined using quantitative X-ray diffraction. Spinel composition was investigated with energy disperse spectroscopy. Ferrous-ferric equilibrium at TL was calculated in a HLW glass as a function of temperature and Po2, based on the previous studies by Schreiber. TL/FeO over the interval 0.0063 < gFeO < 0.051 (1x10-2 kPa < Po2 < 3x10-9 kPa) was estimated from calculated ferrous-ferric equilibrium at TL as 1835 C

  10. Formulation of special glass frit and its use for decontamination of Joule melter employed for vitrification of high level and radioactive liquid waste

    Valsala, T.P.; Mishra, P.K.; Thakur, D.A.; Ghongane, D.E.; Jayan, R.V.; Dani, U.; Sonavane, M.S.; Kulkarni, Y.

    2012-01-01

    Advanced vitrification system at TWMP Tarapur was used for successful vitrification of large volume of HLW stored in waste tank farm. After completion of the operational life of the joule melter, dismantling was planned. Prior to the dismantling, the hold up inventory of active glass product from the melter was flushed out using specially formulated inactive glass frit to reduce the air activity buildup in the cell during dismantling operations. The properties of the special glass frit prepared are comparable with that of the regular product glass. More than 94% of holdup activity was flushed out from the joule melter prior to the dismantling of the melter. (author)

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

    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

  12. Fabrication and characterization of MCC approved testing material - ATM-8 glass

    Wald, J.W.

    1985-10-01

    The Materials Characterization Center (MCC) Approved Testing Material ATM-8 is a borosilicate glass that incorporates elements typical of high-level waste (HLW) resulting from the reprocessing of commercial nuclear reactor fuel. Its composition is based upon the simulated HLW glass type 76-68 (Mendel, J.E. et al., 1977, Annual Report of the Characteristics of High-Level Waste Glasses, BNWL-2252, Pacific Northwest Laboratory, Richland, Washington), to which depleted uranium, technetium-99, neptunium-237 and plutonium-239 have been added at moderate to low levels. The glass was requested by the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. It was produced by the MCC at the Pacific Northwest Laboratory (PNL) operated for the Department of Energy (DOE) by Battelle Memorial Institute. ATM-8 glass was produced in April of 1984, and is the second in a series of testing materials for NNWSI. This report discusses its fabrication (starting materials, batch and glass preparation, measurement and testing equipment, other equipment, procedures, identification system and materials availability and storage, and characterization (bulk density) measurements, chemical analysis, microscopic examination, and x-ray diffraction analysis. 4 refs., 2 figs., 10 tabs

  13. Cosmos & Glass

    Beim, Anne

    1996-01-01

    The article unfolds the architectural visions of glass by Bruno Taut. It refers to inspirations by Paul Sheerbart and litterature and the Crystal Chain, also it analyses the tectonic univers that can be found in the glass pavillion for the Werkbund exposition in Cologne....

  14. Glass Glimpsed

    Lock, Charles

    2015-01-01

    Glass in poetry as it reflects the viewer and as its power of reflection are both reduced and enhanced by technology.......Glass in poetry as it reflects the viewer and as its power of reflection are both reduced and enhanced by technology....

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

    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.

  16. Spin glasses

    Bovier, Anton

    2007-01-01

    Spin glass theory is going through a stunning period of progress while finding exciting new applications in areas beyond theoretical physics, in particular in combinatorics and computer science. This collection of state-of-the-art review papers written by leading experts in the field covers the topic from a wide variety of angles. The topics covered are mean field spin glasses, including a pedagogical account of Talagrand's proof of the Parisi solution, short range spin glasses, emphasizing the open problem of the relevance of the mean-field theory for lattice models, and the dynamics of spin glasses, in particular the problem of ageing in mean field models. The book will serve as a concise introduction to the state of the art of spin glass theory, usefull to both graduate students and young researchers, as well as to anyone curious to know what is going on in this exciting area of mathematical physics.

  17. Fabrication and characterization of MCC approved testing material - ATM-12 glass

    Wald, J.W.

    1985-10-01

    The Materials Characterization Center (MCC) Approved Testing Material ATM-12 is a borosilicate glass that incorporates elements typical of high-level waste (HLW) resulting from the reprocessing of commercial nuclear reactor fuels. The composition has been adjusted to match that predicted for HLW type 76-68 glass at an age of 300 y. Radioactive constituents contained in this glass include depleted uranium, 99 Tc, 237 Np, 239 Pu, and 241 Am. The glass was produced by the MCC at the Pacific Northwest Laboratory (PNL). ATM-12 glass ws produced from July to November of 1984 at the request of the Nevada Nuclear Waste Site Investigations (NNWSI) Program and is the third in a series of glasses produced for NNWSI. Most of the glass produced was in the form of cast bars; special castings and crushed material were also produced. Three kilograms of ATM-12 glass were produced from a feedstock melted in a nitrogen-atmosphere glove box at 1150 0 C in a platinum crucible, and formed into stress-annealed rectangular bars and the special casting shapes requested by NNWSI. Bars of ATM-12 were nominally 1.9 x 1.9 x 10 cm, with an average mass of 111 g each. Nineteen bars and 37 special castings were made. ATM-12 glass has been provided to the NNWSI Program, in the form of bars, crushed powder and special castings. As of August 1985 approximately 590 g of ATM-12 is available for distribution. Requests for materials or services related to this glass should be directed to the Materials Characterization Center Program Office, PNL

  18. GLASS BOX

    Curtis, Laura

    2008-01-01

    The goals of this effort were to develop Glass Box capabilities to allow for the capturing of analyst activities and the associated data resources, track and log the results of automated processing...

  19. Effect of Alumina Incorporation on the Surface Mineralization and Degradation of a Bioactive Glass (CaO-MgO-SiO2-Na2O-P2O5-CaF2-Glycerol Paste

    Dilshat Tulyaganov

    2017-11-01

    Full Text Available This study investigates the dissolution behavior as well as the surface biomineralization in simulated body fluid (SBF of a paste composed of glycerol (gly and a bioactive glass in the system CaO-MgO-SiO2-Na2O-P2O5-CaF2 (BG. The synthesis of the bioactive glass in an alumina crucible has been shown to significantly affect its bioactivity due to the incorporation of aluminum (ca. 1.3–1.4 wt % into the glass network. Thus, the kinetics of the hydroxyapatite (HA mineralization on the glass prepared in the alumina crucible was found to be slower than that reported for the same glass composition prepared in a Pt crucible. It is considered that the synthesis conditions lead to the incorporation of small amount of aluminum into the BG network and thus delay the HA mineralization. Interestingly, the BG-gly paste was shown to have significantly higher bioactivity than that of the as-prepared BG. Structural analysis of the paste indicate that glycerol chemically interacts with the glass surface and strongly alter the glass network architecture, thus generating a more depolymerized network, as well as an increased amount of silanol groups at the surface of the glass. In particular, BG-gly paste features early intermediate calcite precipitation during immersion in SBF, followed by hydroxyapatite formation after ca. seven days of SBF exposure; whereas the HA mineralization seems to be suppressed in BG, probably a consequence of the incorporation of aluminum into the glass network. The results obtained within the present study reveal the positive effect of using pastes based on bioactive glasses and organic carriers (here alcohols which may be of interest not only due to their advantageous visco-elastic properties, but also due to the possibility of enhancing the glass bioactivity upon surface interactions with the organic carrier.

  20. Actinide partitioning from HLW in a continuous DIDPA extraction process by means of centrifugal extractors

    Morita, Y.; Kubota, M.; Glatz, J.P.; Koch, L.; Pagliosa, G.; Roemer, K.; Nicholl, A.

    1996-01-01

    An experiment on actinide partitioning from real high level waste (HLW) was performed in a continuous process by extraction with diisodecylphosphoric acid (DIDPA) using a battery of 12 centrifugal extractors installed in a hot cell. The HNO 3 concentration of the HLW was adjusted to 0.5 M by dilution. The extraction section had 8 stages, and H 2 O 2 was added to extract Np effectively. After extraction, Am and Cm were back-extracted with 4 M HNO 3 in 4 stages and Np and Pu were stripped with 0.8 M H 2 C 2 O 4 in 8 stages. The actinides, expect Np, were extracted from HLW with a very high yield. Although only 84% of the Np were recovered in the present experiment, the recovery would be improved to 99.7 % by increasing the temperature to 45 degree C, the number of stages from 8 to 16 and the H 2 O 2 concentration from 1 M to 2 M. Long-lived Tc and the main heat and radiation emitters Cs and Sr were not extracted and were thus separated from the actinides with high decontamination factors. About 98% of Am and Cm were recovered from the loaded solvent in the first stripping step with 4M HNO 3 . About 86% of Np and about 92% of Pu were back-extracted with 0.8 M H 2 C 2 O 4 . These incomplete recoveries would be improved by increasing the number of stages and by optimizing the other process parameters. 18 refs., 5 figs., 3 tabs

  1. Cost effects of Cu powder and bentonite on the disposal costs of an HLW repository in

    Kim, Sung Ki; Lee, Min Soo; Lee, Jong Youl; Choi, Heui Joo; Choi, Jong Won

    2008-01-01

    This paper provides the cost effect results of Cu powder and bentonite on the disposal cost for an HLW repository in Korea. In the cost analysis for both of these cost drivers, the price of Cu powder and the bentonite can affect the canister cost and the bentonite cost of the disposal holes as well as backfilling cost of the tunnels, respectively. Finally, we found that the unit cost of Cu and bentonite was the dominant cost drivers for the surface and underground facilities of an HLW repository. Therefore, an optimization of a canister and the layout of a disposal hole and disposal tunnels are essential to decrease the direct disposal cost of spent fuels. The disposal costs can be largely divided into two parts such as a surface facilities' cost and an underground facilities' cost. According to the KRS' cost analysis, the encapsulation material as well as the buffering and backfilling cost were the significant costs. Especially, a canister's cost was approximately estimated to be more than one fourth of the overall disposal costs. So it can be estimated that the unit cost of Cu powder is an important cost diver. Because the outer shell of the canister was made of Cu powder by a cold spray coating method. In addition, the unit cost of bentonite can also affect the buffering and the backfilling costs of the disposal holes and the disposal tunnels. But, these material costs will be highly expensive and unstable due to the modernization of the developing countries. So the studies for a material cost should be continued to identify the actual cost of an HLW repository

  2. An analytical overview of the consequences of microbial activity in a Swiss HLW repository

    McKinley, I.G.; West, J.M.; Grogan, H.A.

    1985-04-01

    Microorganisms are known to be important factors in many geochemical processes and their presence can be assured throughout the envisaged Swiss type C repository for HLW. It is likely that both introduced and resident microbes will colonise the near-field even at times when ambient temperature and radiation fields are relatively high. A simple quantitative model has been developed which indicates that microbial growth in the near-field is limited by the rate of supply of chemical energy from corrosion of the canister. Microbial processes examined include biodegradation of structural and packaging materials, alteration of groundwater chemistry (Eh, pH, organic complexant concentration) and direct nuclide uptake by microorganisms. The most important effects of such organisms are likely to be enhancement of release and mobility of key nuclides due to their complexation by microbial by-product. Resident micro-organisms in the far-field could potentially act as 9 living colloids' thus enhancing nuclide transport. In the case of flow paths through shear zones (kakirites), however, any microbes capable of penetrating the surrounding weathered rock matrix would be extensively retarded. It is concluded that microbial processes are unlikely to be of significance for HLW but will be more important for low/intermediate waste types. As data requirements are similar for all waste types, results from such studies would also resolve the main uncertainties remaining for the HLW case. Key research areas are identified as characterisation of a) nutrient availability in the near-field, b) the bioenergetics of iron corrosion, c) production of organic by-products, d) nuclide sorption by organisms and e) microbial mobility in the near-and far-field

  3. Application of QA to R ampersand D support of HLW programs

    Ryder, D.E.

    1988-01-01

    Quality has always been of primary importance in the research and development (R ampersand D) environment. An organization's ability to attract funds for new or continued research is largely dependent on the quality of past performance. However, with the possible exceptions of peer reviews for fund allocation and the referee process prior to publication, past quality assurance (QA) activities were primarily informal good practices. This resulted in standards of acceptable practice that varied from organization to organization. The increasing complexity of R ampersand D projects and the increasing need for project results to be upheld outside the scientific community (i.e., lawsuits and licensing hearings) are encouraging R ampersand D organizations and their clients to adopt more formalized methods for the scientific process and to increase control over support organizations (i.e., suppliers and subcontractors). This has become especially true for R ampersand D organizations involved in the high-level (HLW) projects for a number of years. The PNL began to implement QA program requirements within a few HLW repository preliminary studies in 1978. In 1985, PNL developed a comprehensive QA program for R ampersand D activities in support of two of the proposed repository projects. This QA program was developed by the PNL QA department with a significant amount of support assistance and guidance from PNL upper management, the Basalt Waste Isolation Project (BWIP), and the Salt Repository Program Office (SPRO). The QA program has been revised to add a three-level feature and is currently being implemented on projects sponsored by the Office of Geologic Repositories (DOE/OGR), Repository Technology Program (DOE-CH), Nevada Nuclear Waste Storage Investigation (NNWSI) Project, and other HLW projects

  4. Sensitivity of Nuclide Release Behavior to Groundwater Flow in an HLW Repository

    Lee, Youn-Myoung; Hwang, Yong-Soo

    2008-01-01

    Evaluation of the dose exposure rate to human being due to long-term nuclide releases from a high-level waste repository (HLW) is of importance to meet the dose limit presented by the regulatory bodies in order to ensure the performance of a repository. During the last few years, tools by which such a dose rate to an individual can be evaluated have been developed and implemented for a practical calculation to demonstrate the suitability of an HLW repository, with the aid of commercial tools such as AMBER and GoldSim, both of which are capable of probabilistic and deterministic calculations with their convenient user interface. Recently a migration from AMBER based models to GoldSim based ones has been made in accordance with a better feature of GoldSim, which is designed to facilitate the object-oriented modules to address any specialized programs, similar to solving jig saw puzzles and shows more advantage in a detailed complex modeling over AMBER. Recently a compartment modeling approach both for a geosphere and biosphere has been mainly carried out with AMBER in KAERI, which causes a necessity for a newly devised system performance evaluation model in which geosphere and biosphere models could be coupled organically together with less conservatism in the frame of the development of a total system performance assessment modeling tool, which could be successfully done with the aid of GoldSim. Therefore, through the current study, some probabilistic results of the GoldSim approach for a normal situation that could take place in a typical HLW repository are introduced

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

    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.

  6. Development of database and QA systems for post closure performance assessment on a potential HLW repository

    Hwang, Y. S.; Kim, S. G.; Kang, C. H.

    2002-01-01

    In TSPA of long-term post closure radiological safety on permanent disposal of HLW in Korea, appropriate management of input and output data through QA is necessary. The robust QA system is developed using the T2R3 principles applicable for five major steps in R and D's. The proposed system is implemented in the web-based system so that all participants in TSRA are able to access the system. In addition, the internet based input database for TSPA is developed. Currently data from literature surveys, domestic laboratory and field experiments as well as expert elicitation are applied for TSPA

  7. The AGP-Project conceptual design for a Spanish HLW final disposal facility

    Biurrun, E.; Engelmann, H.-J.; Huertas, F.; Ulibarri, A.

    1992-01-01

    Within the framework of the AGP Project a Conceptual Design for a HLW Final Disposal Facility to be eventually built in an underground salt formation in Spain has been developed. The AGP Project has the character of a system analysis. In the current project phase I several alternatives has been considered for different subsystems and/or components of the repository. The system variants, developed to such extent as to allow a comparison of their advantages and disadvantages, will allow the selection of a reference concept, which will be further developed to technical maturity in subsequent project phases. (author)

  8. Collaboration, Automation, and Information Management at Hanford High Level Radioactive Waste (HLW) Tank Farms

    Aurah, Mirwaise Y.; Roberts, Mark A.

    2013-01-01

    Washington River Protection Solutions (WRPS), operator of High Level Radioactive Waste (HLW) Tank Farms at the Hanford Site, is taking an over 20-year leap in technology, replacing systems that were monitored with clipboards and obsolete computer systems, as well as solving major operations and maintenance hurdles in the area of process automation and information management. While WRPS is fully compliant with procedures and regulations, the current systems are not integrated and do not share data efficiently, hampering how information is obtained and managed

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

    Hara, Shigeo; Terai, Ryohei; Yamanaka, Hiroshi

    1983-01-01

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

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

    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

  11. Chemical analysis of simulated high level waste glasses to support stage III sulfate solubility modeling

    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.

  12. Studies on the immobilization of simulated HLW in NaTi2(PO4)3 (NTP) matrix

    Raja Madhavan, R.; Govindan Kutty, K.V.; Gandhi, A.S.

    2015-01-01

    Immobilization of high level nuclear waste (HLW) is a big challenge faced by the nuclear industry today. The HLW has to be contained and isolated from the biosphere for geological timescales. NZP family of compounds is very versatile monophasic hosts for HLW immobilization. Their crystal structure can accommodate nearly all the cations known to be present in HLW due to its open structure with voids of different size. In the present study a systematic investigation on NaTi 2 (PO 4 ) 3 belonging to the NZP family; as a potential host for HLW immobilization was carried out. A simulated HLW expected from Fast Breeder Test Reactor, India (FBTR) (150Gwd/T burnup, 1 year cooling) was used. Simulated NTP waste forms with 5, 10, 15 wt. % waste loading were prepared by employing a wet chemical method and characterized. Single phase simulated NTP waste forms with up to 5 wt.% waste loading could be prepared for samples sintered in air and above 5 wt.% waste loading, monazite phase is observed as a minor secondary phase. It was found that when sintering is done in Ar/10%H 2 , NTP matrix accepts up to 10 wt.% waste loading without formation of any second phase. From the SEM studies, it was observed that samples sintered in air as well as Ar/10%H 2 palladium segregated as a metal phase and uniformly distributed throughout the waste matrix. The elemental mapping revealed retention of some of the fission products like Ru, Mo, Cs that are volatile during sintering above 1173 K and are homogenously distributed in the matrix. (author)

  13. Crystal Settling, Redox, and High Temperature Properties of ORP HLW and LAW Glasses, VSL-09R1510-1

    Kruger, A. A. [The Catholic Univ. of America, Washington D. C., (United States); Gan, H. [The Catholic Univ. of America, Washington D. C., (United States); Viragh, C. [The Catholic Univ. of America, Washington D. C., (United States); Mckeown, D. A. [The Catholic Univ. of America, Washington D. C., (United States); Muller, I. S. [The Catholic Univ. of America, Washington D. C., (United States); Cecil, R. [The Catholic Univ. of America, Washington D. C., (United States); Kot, W. K. [The Catholic Univ. of America, Washington D. C., (United States); Joseph, I. [EnergySolutions, Laurel, MD (United States); Wang, C. [The Catholic Univ. of America, Washington D. C., (United States); Pegg, I. L. [The Catholic Univ. of America, Washington D. C., (United States); Chaudhuri, M. [The Catholic Univ. of America, Washington D. C., (United States); Zhao, W. [The Catholic Univ. of America, Washington D. C., (United States); Feng, Z. [The Catholic Univ. of America, Washington D. C., (United States)

    2015-06-08

    This report describes the results of testing specified by the Test Plans (VSL-08T1520-1 Rev 0 and VSL-08T1510-1 Rev 0). The work was performed in compliance with the quality assurance requirements specified in the Test Plans. Results required by the Test Plans are reported. The test results and this report have been reviewed for correctness, technical adequacy, completeness, and accuracy.

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

    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)

  15. Capacity of burning and transmutation reactor and grouping in partitioning of HLW in self-consistent fuel recycle

    Kitamoto, A.; Mulyanto

    1993-01-01

    The concept of capacity of B/T reactor and grouping for partitioning of HLW has been developed in order to perform self-consistent fuel recycle. The concept of grouping of radionuclides is proposed herein, such as Group MA1 (MA below Cm), Group MA2 (Cm and higher MA), Group A ( 99 Te, 129 I, and 135 Cs), Group B ( 137 Cs and 90 Sr) and Group R (the partitioned remain of HLW). In this study P-T treatment were optimized for the in-core and out-core system, respectively. (author). 7 refs., 10 figs

  16. Enhanced sludge processing of HLW: Hydrothermal oxidation of chromium, technetium, and complexants by nitrate. 1998 annual progress report

    Buelow, S.J.; Robinson, J.M.

    1998-01-01

    'The objective of this project is to develop the scientific basis for hydrothermal separation of chromium from High Level Waste (HLW) sludges. The worked is aimed at attaining a fundamental understanding of chromium speciation, oxidation/reduction and dissolution kinetics, reaction mechanisms, and transport properties under hydrothermal conditions in both simple and complex salt solutions that will ultimately lead to an efficient chromium leaching process. This report summarizes the research over the first 1.5 years of a 3 year project. The authors have examined the dissolution of chromium hydroxide using different oxidants as a function of temperature and alkalinity. The results and possible applications to HLW sludges are discussed'

  17. Time evolution of the Clay Barrier Chemistry in a HLW deep geological disposal in granite

    Font, I.; Miguel, M. J.; Juncosa, R.

    2000-01-01

    The main goal of a high level waste geological disposal is to guarantee the waste isolation from the biosphere, locking them away into very deep geological formations. The best way to assure the isolation is by means of a multiple barrier system. These barriers, in a serial disposition, should assure the confinement function of the disposal system. Two kinds of barriers are considered: natural barriers (geological formations) and engineered barriers (waste form, container and backfilling and sealing materials). Bentonite is selected as backfilling and sealing materials for HLW disposal into granite formations, due to its very low permeability and its ability to fill the remaining spaces. bentonite has also other interesting properties, such as, the radionuclide retention capacity by sorption processes. Once the clay barrier has been placed, the saturation process starts. The granite groundwater fills up the voids of the bentonite and because of the chemical interactions, the groundwater chemical composition varies. Near field processes, such as canister corrosion, waste leaching and radionuclide release, strongly depends on the water chemical composition. Bentonite pore water composition is such a very important feature of the disposal system and its determination and its evolution have great relevance in the HLW deep geological disposal performance assessment. The process used for the determination of the clay barrier pore water chemistry temporal evolution, and its influence on the performance assessment, are presented in this paper. (Author)

  18. Proposals of geological sites for L/ILW and HLW repositories. Geological background. Text volume

    2008-01-01

    On April 2008, the Swiss Federal Council approved the conceptual part of the Sectoral Plan for Deep Geological Repositories. The Plan sets out the details of the site selection procedure for geological repositories for low- and intermediate-level waste (L/ILW) and high-level waste (HLW). It specifies that selection of geological siting regions and sites for repositories in Switzerland will be conducted in three stages, the first one (the subject of this report) being the definition of geological siting regions within which the repository projects will be elaborated in more detail in the later stages of the Sectoral Plan. The geoscientific background is based on the one hand on an evaluation of the geological investigations previously carried out by Nagra on deep geological disposal of HLW and L/ILW in Switzerland (investigation programmes in the crystalline basement and Opalinus Clay in Northern Switzerland, investigations of L/ILW sites in the Alps, research in rock laboratories in crystalline rock and clay); on the other hand, new geoscientific studies have also been carried out in connection with the site selection process. Formulation of the siting proposals is conducted in five steps: A) In a first step, the waste inventory is allocated to the L/ILW and HLW repositories; B) The second step involves defining the barrier and safety concepts for the two repositories. With a view to evaluating the geological siting possibilities, quantitative and qualitative guidelines and requirements on the geology are derived on the basis of these concepts. These relate to the time period to be considered, the space requirements for the repository, the properties of the host rock (depth, thickness, lateral extent, hydraulic conductivity), long-term stability, reliability of geological findings and engineering suitability; C) In the third step, the large-scale geological-tectonic situation is assessed and large-scale areas that remain under consideration are defined. For the L

  19. Threshold Assessment: Definition of Acceptable Sites as Part of Site Selection for the Japanese HLW Program

    McKenna, S.A.; Wakasugi, Keiichiro; Webb, E.K.; Makino, Hitoshi; Ishihara, Yoshinao; Ijiri, Yuji; Sawada, Atsushi; Baba, Tomoko; Ishiguro, Katsuhiko; Umeki, Hiroyuki

    2000-01-01

    For the last ten years, the Japanese High-Level Nuclear Waste (HLW) repository program has focused on assessing the feasibility of a basic repository concept, which resulted in the recently published H12 Report. As Japan enters the implementation phase, a new organization must identify, screen and choose potential repository sites. Thus, a rapid mechanism for determining the likelihood of site suitability is critical. The threshold approach, described here, is a simple mechanism for defining the likelihood that a site is suitable given estimates of several critical parameters. We rely on the results of a companion paper, which described a probabilistic performance assessment simulation of the HLW reference case in the H12 report. The most critical two or three input parameters are plotted against each other and treated as spatial variables. Geostatistics is used to interpret the spatial correlation, which in turn is used to simulate multiple realizations of the parameter value maps. By combining an array of realizations, we can look at the probability that a given site, as represented by estimates of this combination of parameters, would be good host for a repository site

  20. Natural analogues for containment-providing barriers for a HLW repository in salt

    Wolf, J.; Noseck, U.

    2015-06-15

    In 2005, a German research project was started to develop a novel approach to prove safety for a HLW repository in a salt formation, to refine the safety concept, to identify open scientific issues and to define necessary R&D work. This project aimed at identifying the key information for a HLW repository in salt. One important question is how this information may be best fulfilled by natural analogue studies. This question is answered by starting a review of the required key information needs of the safety case (post-closure phase) in order to assess whether or not these requirements can be supported by natural analogues information. In order to structure the review and to address the key elements of the safety concepts, three types of natural analogues are distinguished: (i) natural analogues for the integrity of the geological barrier, (ii) natural analogues for the integrity of the geotechnical barriers and (iii) natural analogues for release scenarios. For the safety case in salt type (i) and (ii) are of highest importance and are treated in this paper. The assessment documented in this paper on the one hand indicates the high potential benefit of natural analogues for a safety case in salt and on the other hand helps to focus the available human and financial resources for the safety case on the most safety-relevant aspects. (authors)

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

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

    2016-07-14

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

  2. Status of the safety concept and safety demonstration for an HLW repository in salt. Summary report

    Bollingerfehr, W.; Buhmann, D.; Filbert, W.; and others

    2013-12-15

    Salt formations have been the preferred option as host rocks for the disposal of high level radioactive waste in Germany for more than 40 years. During this period comprehensive geological investigations have been carried out together with a broad spectrum of concept and safety related R and D work. The behaviour of an HLW repository in salt formations, particularly in salt domes, has been analysed in terms of assessment of the total system performance. This was first carried out for concepts of generic waste repositories in salt and, since 1998, for a repository concept with specific boundary conditions, taking the geology of the Gorleben salt dome as an example. Suitable repository concepts and designs were developed, the technical feasibility has been proven and operational and long-term safety evaluated. Numerical modelling is an important input into the development of a comprehensive safety case for a waste repository. Significant progress in the development of numerical tools and their application for long-term safety assessment has been made in the last two decades. An integrated approach has been used in which the repository concept and relevant scientific and engineering data are combined with the results from iterative safety assessments to increase the clarity and the traceability of the evaluation. A safety concept that takes full credit of the favourable properties of salt formations was developed in the course of the R and D project ISIBEL, which started in 2005. This concept is based on the safe containment of radioactive waste in a specific part of the host rock formation, termed the containment providing rock zone, which comprises the geological barrier, the geotechnical barriers and the compacted backfill. The future evolution of the repository system will be analysed using a catalogue of Features, Events and Processes (FEP), scenario development and numerical analysis, all of which are adapted to suit the safety concept. Key elements of the

  3. Incorporating Cold Cap Behavior in a Joule-heated Waste Glass Melter Model

    Varija Agarwal; Donna Post Guillen

    2013-08-01

    In this paper, an overview of Joule-heated waste glass melters used in the vitrification of high level waste (HLW) is presented, with a focus on the cold cap region. This region, in which feed-to-glass conversion reactions occur, is critical in determining the melting properties of any given glass melter. An existing 1D computer model of the cold cap, implemented in MATLAB, is described in detail. This model is a standalone model that calculates cold cap properties based on boundary conditions at the top and bottom of the cold cap. Efforts to couple this cold cap model with a 3D STAR-CCM+ model of a Joule-heated melter are then described. The coupling is being implemented in ModelCenter, a software integration tool. The ultimate goal of this model is to guide the specification of melter parameters that optimize glass quality and production rate.

  4. Glass compositions

    France, P W

    1985-05-30

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

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

    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)

  6. Final Report Integrated DM1200 Melter Testing Of Bubbler Configurations Using HLW AZ-101 Simulants VSL-04R4800-4, Rev. 0, 10/5/04

    Kruger, A.A.; Matlack, K.S.; Gong, W.; Bardakci, T.; D'Angelo, N.A.; Lutze, W.; Callow, R.A.; Brandys, M.; Kot, W.K.; Pegg, I.L.

    2011-01-01

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 HLW simulants. The tests reported herein are a subset of six tests from a larger series of tests described in the Test Plan for the work; results from the other tests have been reported separately. The solids contents of the melter feeds were based on the WTP baseline value for the solids content of the feeds from pretreatment which changed during these tests from 20% to 15% undissolved solids resulting in tests conducted at two feed solids contents. Based on the results of earlier tests with single outlet 'J' bubblers, initial tests were performed with a total bubbling rate of 651 pm. The first set of tests (Tests 1A-1E) addressed the effects of skewing this total air flow rate back and forth between the two installed bubblers in comparison to a fixed equal division of flow between them. The second set of tests (2A-2D) addressed the effects of bubbler depth. Subsequently, as the location, type and number of bubbling outlets were varied, the optimum bubbling rate for each was determined. A third (3A-3C) and fourth (8A-8C) set of tests evaluated the effects of alternative bubbler designs with two gas outlets per bubbler instead of one by placing four bubblers in positions simulating multiple-outlet bubblers. Data from the simulated multiple outlet bubblers were used to design bubblers with two outlets for an additional set of tests (9A-9C). Test 9 was also used to determine the effect of small sugar additions to the feed on ruthenium volatility. Another set of tests (10A-10D) evaluated the effects on production rate of spiking the feed with chloride and sulfate. Variables held constant to the extent possible included melt temperature, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The feed rate was increased to the point that a constant, essentially complete, cold cap was achieved

  7. FINAL REPORT INTEGRATED DM1200 MELTER TESTING OF BUBBLER CONFIGURATIONS USING HLW AZ-101 SIMULANTS VSL-04R4800-4 REV 0 10/5/04

    KRUGER AA; MATLACK KS; GONG W; BARDAKCI T; D' ANGELO NA; LUTZE W; CALLOW RA; BRANDYS M; KOT WK; PEGG IL

    2011-12-29

    This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of AZ-101 HLW simulants. The tests reported herein are a subset of six tests from a larger series of tests described in the Test Plan for the work; results from the other tests have been reported separately. The solids contents of the melter feeds were based on the WTP baseline value for the solids content of the feeds from pretreatment which changed during these tests from 20% to 15% undissolved solids resulting in tests conducted at two feed solids contents. Based on the results of earlier tests with single outlet 'J' bubblers, initial tests were performed with a total bubbling rate of 651 pm. The first set of tests (Tests 1A-1E) addressed the effects of skewing this total air flow rate back and forth between the two installed bubblers in comparison to a fixed equal division of flow between them. The second set of tests (2A-2D) addressed the effects of bubbler depth. Subsequently, as the location, type and number of bubbling outlets were varied, the optimum bubbling rate for each was determined. A third (3A-3C) and fourth (8A-8C) set of tests evaluated the effects of alternative bubbler designs with two gas outlets per bubbler instead of one by placing four bubblers in positions simulating multiple-outlet bubblers. Data from the simulated multiple outlet bubblers were used to design bubblers with two outlets for an additional set of tests (9A-9C). Test 9 was also used to determine the effect of small sugar additions to the feed on ruthenium volatility. Another set of tests (10A-10D) evaluated the effects on production rate of spiking the feed with chloride and sulfate. Variables held constant to the extent possible included melt temperature, plenum temperature, cold cap coverage, the waste simulant composition, and the target glass composition. The feed rate was increased to the point that a constant, essentially complete, cold cap was

  8. Effects of a Capital Investment and a Discount Rate on the Optimal Operational Duration of an HLW Repository

    Kim, Sung Ki; Lee, Min Soo; Choi, Heui Joo; Choi, Jong Won

    2008-01-01

    This study aims to estimate the effects of a capital investment and a discount rate on the optimal operational duration of an HLW repository. According to the previous researches of the KRS(Korea Reference System) for an HLW repository, the amounts of 7,068,200 C$K and 2,636.2 MEUR are necessary to construct and operate surface and underground facilities. Since these huge costs can be a burden to some national economies, a study for a cost optimization should be performed. So we aim to drive the dominant cost driver for an optimal operational duration. A longer operational duration may be needed to dispose of more spent fuels continuously from a nuclear power plant, or to attain a retrievability of an HLW repository at a depth of 500 m below the ground level in a stable plutonic rock body. In this sense, an extended operational duration for an HLW repository affects the overall disposal costs of a repository. In this paper, only the influence of a capital investment and a discount rate was estimated from the view of optimized economics. Because these effects must be significant factors to minimize the overall disposal costs based on minimizing the sum of operational costs and capital investments

  9. Study on systematic integration technology of design and safety assessment for HLW geological disposal. 2. Research document

    Ishihara, Yoshinao; Fukui, Hiroshi; Sagawa, Hiroshi; Matsunaga, Kenichi; Ito, Takaya; Kohanawa, Osamu; Kuwayama, Yuki

    2003-02-01

    The present study was carried out relating to basic design of the Geological Disposal Technology Integration System' that will be systematized as knowledge base for design analysis and safety assessment of HLW geological disposal system by integrating organically and hierarchically various technical information in three study field. The key conclusions are summarized as follows: (1) As referring to the current performance assessment report, the technical information for R and D program of HLW geological disposal system was systematized hierarchically based on summarized information in a suitable form between the work flow (work item) and processes/characteristic flow (process item). (2) As the result of the systematized technical information, database structure and system functions necessary for development and construction to the computer system were clarified in order to secure the relation between technical information and data set for assessment of HLW geological disposal system. (3) The control procedure for execution of various analysis code used by design and safety assessment in HLW geological disposal study was arranged possibility in construction of 'Geological Disposal Technology Integration System' after investigating the distributed computing technology. (author)

  10. Safety case development in the Japanese programme for geological disposal of HLW: Evolution in the generic stage

    Ueda, Hiroyoshi; Ishiguro, Katsuhiko; Takeuchi, Mitsuo; Fujihara, Hiroshi; Takeda, Seietsu

    2014-01-01

    In the Japanese programme for nuclear power generation, the safe management of the resulting radioactive waste, particularly vitrified high-level waste (HLW) from fuel reprocessing, has been a major concern and a focus of R and D since the late 70's. According to the specifications in a report issued by an advisory committee of the Japan Atomic Energy Commission (JAEC, 1997), the Second Progress Report on R and D for the Geological Disposal of HLW (H12 report) (JNC, 2000) was published after two decades of R and D activities and showed that disposal of HLW in Japan is feasible and can be practically implemented at sites which meet certain geological stability requirements. The H12 report supported government decisions that formed the basis of the 'Act on Final Disposal of Specified Radioactive Waste' (Final Disposal Act), which came into force in 2000. The Act specifies deep geological disposal of HLW at depths greater than 300 metres, together with a stepwise site selection process in three stages. Following the Final Disposal Act, the supporting 'Basic Policy for Final Disposal' and the 'Final Disposal Plan' were authorised in the same year. (authors)

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

    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.

  12. Glass: Rotary Electric Glass Furnace

    Recca, L.

    1999-01-29

    Compared to conventional gas-fired furnaces, the new rotary electric furnace will increase energy efficiency while significantly reducing air emissions, product turnaround time, and labor costs. As this informative new fact sheet explains, the thousand different types of glass optical blanks produced for the photonics industry are used for lasers, telescopes, cameras, lights, and many other products.

  13. Proceedings: EPRI Workshop 2 -- Technical basis for EPA HLW disposal criteria

    Rogers, V.

    1993-03-01

    The Electric Power Research Institute (EPRI) sponsored this workshop to address the scientific and technical issues underlying the regulatory criteria, or standard, for the disposal of spent nuclear fuel, high-level radioactive waste, and transuranic waste, commonly referred to collectively as high-level waste (HLW). These regulatory criteria were originally promulgated by the US Environmental Protection Agency (EPA) in 40 CFR Part 191 in 1985. However, significant portions of the regulation were remanded by the Ninth Circuit Court of Appeals in 1987. This is the second of two workshops. Topics discussed include: gas pathway; individual and groundwater protection; human intrusion; population protection; performance; TRU conversion factors and discussions. Individual projects re processed separately for the databases

  14. KAERI Underground Research Facility (KURF) for the Demonstration of HLW Disposal Technology

    Hahn, P. S.; Cho, W. J.; Kwon, S.

    2006-01-01

    In order to dispose of high-level radioactive waste(HLW) safely in geological formations, it is necessary to assess the feasibility, safety, appropriateness, and stability of the disposal concept at an underground research site, which is constructed in the same geological formation as the host rock. In this paper, the current status of the conceptual design and the construction of a small scale URL, which is named as KURF, were described. To confirm the validity of the conceptual design of the underground facility, a geological survey including a seismic refraction survey, an electronic resistivity survey, a borehole drilling, and in situ and laboratory tests had been carried out. Based on the site characterization results, it was possible to effectively design the KURF. The construction of the KURF was started in May 2005 and the access tunnel was successfully completed in March 2006. Now the construction of the research modules is under way

  15. Role of international collaboration in PNC's R ampersand D programme for HLW disposal

    Masuda, Sumio; Umeki, Hiroyuki; Yamakawa, Minoru

    1996-01-01

    PNC has been active in promoting international cooperation in connection with the Japanese HLW disposal programme, based on both a bilateral and multilateral approach. Both types of cooperation are extremely useful; in particular, bilateral cooperation has the advantage of providing opportunities for in-depth discussions in mutual areas of interest. By way of contrast, multilateral cooperation also provides an international arena for broader discussion and corroboration of output from individual R ampersand D programmes. International collaboration also provides young researchers with an opportunity to learn from experience. Depending on the issues to be tackled, appropriate forms of collaboration have been integrated into PNC's strategy for maximizing output. The lessons learned from collaboration are very valuable and can be used directly in their programme to enhance its credibility. The format of collaboration has also been extensively developed: it has been found that resources can be utilized more effectively by sharing them appropriately

  16. Depth optimization for the Korean HLW repository System within a discontinuous and saturated granitic rock mass

    Kim, Jhin Wung; Bae, Dae Seok; Choi, Jong Won

    2005-12-01

    The present study is to evaluate the material properties of the compacted bentonite, backfill material, canister cast iron insert, and the rock mass for the Korean HLW repository system. These material properties are either measured, or taken from other countries, through the evaluation of the thermal, hydraulic, and mechanical interaction behavior of a repository. After the evaluation of the material properties, the most appropriate and economical depth as well as the layout of a single layer repository is to be recommended. Material properties used for the granitic rock mass, rock joints, PWR spent fuel, disposal canister, compacted bentonite, backfill material, and ground water are the data collected domestically, and foreign data are used for some of the data not available domestically. The repository model includes a saturated granitic rock mass with joints, PWR spent fuel in a disposal canister surrounded by compacted bentonite inside a deposition hole, and backfill material in the rest of the space within a repository cavern

  17. Technical Standards on the Safety Assessment of a HLW Repository in Other Countries

    Lee, Sung Ho; Hwang, Yong Soo

    2009-01-01

    The basic function of HLW disposal system is to prevent excessive radio-nuclides being leaked from the repository in a short time. To do this, many technical standards should be developed and established on the components of disposal system. Safety assessment of a repository is considered as one of technical standards, because it produces quantitative results of the future evolution of a repository based on a reasonably simplified model. In this paper, we investigated other countries' regulations related to safely assessment focused on the assessment period, radiation dose limits and uncertainties of the assessment. Especially, in the investigation process of the USA regulations, the USA regulatory bodies' approach to assessment period and peak dose is worth taking into account in case of a conflict between peak dose from safety assessment and limited value in regulation.

  18. Influence of the reprocessing flow sheet on the HLW solidification technology

    Baetsle, L.H.

    1981-01-01

    The introduction of Pu recycled LWR and CMFBR fuel will require the addition of a second dissolution step to quantitation recover Pu. If process modifications can be brought to the head-end procedures it is advisable to remove Ru, Te, Mo, Pd by high performance centrifugation and to volatilize soluble RuNO(NO 3 ) 2 by sparing with ozone. This changes improve the liquid extraction efficiency and simplify the off gas treatment during calcination and vitrification of HAWC. The conversion of HAW to HAWC by evaporation is accompagnied by some-volatilization of Ru and Cs. Organic reductants reduce the Ru volatilization. The introduction of salt free reagents during feed adjustment steps will decrease Na content in the HLW. The main impact of the use of salt free reagent will have its bearing on the LAW and ILIW treatment and conditioning. (DG)

  19. HLW disposal by fission reactors; calculation of trans-mutation rate and recycle

    Mulyanto

    1997-01-01

    Transmutation of MA (Minor actinide) and LLFPS (long-lived fission products) into stable nuclide or short-lived isotopes by fission reactors seem to become an alternative technology for HLW disposal. in this study, transmutation rate and recycle calculation were developed in order to evaluate transmutation characteristics of MA and LLFPs in the fission reactors. inventory of MA and LLFPs in the transmutation reactors were determined by solving of criticality equation with 1-D cylindrical geometry of multigroup diffusion equations at the beginning of cycle (BOC). transmutation rate and burn-up was determined by solving of depletion equation. inventory of MA and LLFPs was calculated for 40 years recycle. From this study, it was concluded that characteristics of MA and LLFPs in the transmutation reactors can be evaluated by recycle calculation. by calculation of transmutation rate, performance of fission reactor for transmutation of MA or LLFPs can be discussed

  20. An Ilustrative Nuclide Release Behavior from an HLW Repository due to an Earthquake Event

    Lee, Youn-Myoung; Hwang, Yong-Soo; Choi, Jong-Won

    2008-01-01

    Program for the evaluation of a high-level waste repository which is conceptually modeled. During the last few years, programs developed with the aid of AMBER and GoldSim by which nuclide transports in the near- and far-field of a repository as well as transport through the biosphere under various normal and disruptive release scenarios could be modeled and evaluated, have been continuously demonstrated. To show its usability, as similarly done for the natural groundwater flow scheme, influence of a possible disruptive event on a nuclide release behavior from an HLW repository system caused naturally due to an earthquake has been investigated and illustrated with the newly developed GoldSim program

  1. Assessment of dose conversion factors in a generic biosphere of a Korea HLW repository

    Hwang, Y. S.; Park, J. B.; Kang, C. H.

    2002-01-01

    Radioactive species released from a waste repository migrate through engineered and natural barriers and eventually reach the biosphere. Once entered the biosphere, contaminants transport various exposure pathways and finally reach a human. In this study the full RES matrix explaining the key compartments in the biosphere and their interactions is introduced considering the characteristics of the Korean biosphere. Then the three exposure groups are identified based on the compartments of interest. The full exposure pathways and corresponding mathematical expression for mass transfer coefficients and etc are developed and applied to assess the dose conversion factors of nuclides for a specific exposure group. Dose conversion factors assessed in this study will be used for total system performance assessment of a potential Korean HLW repository

  2. Current status of preparing buffer/backfill block in HLW disposal abroad

    Yan Ming; Wang Xuewen; Zhang Huyuan

    2014-01-01

    There is an urgent need for China to commence the full-scale compaction test, resolving the preparation problem for buffer/backfill blocks when underground research laboratory project is planned for High Level Radioactive Waste (HLW) disposal. The foreign countries have some research about the preparation of buffer/backfill blocks in engineered barrier systems. The foreign research shows that installation of clay blocks with sector shape at waste pollution area is a feasible engineering method. Compacted clay blocks need to be cured in a cabinet with controlled temperature and humidity to avoid desiccation and surface powdering. A freeze mixing method, mixing powdered-ice and cooled bentonite, can be operated more easily and obtain more uniform hydration than the traditional mixing of water and bentonite. It is helpful to review and adsorb the foreign research results for the design of full-scale test of bentonite compaction. (authors)

  3. Report - Melter Testing of New High Bismuth HLW Formulations VSL-13R2770-1

    Kruger, Albert A.; Pegg, I. L.; Kot, W. K.; Gan, H.; Matlack, K. S.

    2013-11-13

    The primary objective of the work described was to test two glasses formulated for a high bismuth waste stream on the DM100 melter system. Testing was designed to determine processing characteristics and production rates, assess the tendency for foaming, and confirm glass properties. The glass compositions tested were previously developed to maintain high waste loadings and processing rates while suppressing the foaming observed in previous tests

  4. Drop Calculations of HLW Canister and Pu Can-in-Canister

    Sreten Mastilovic

    2001-01-01

    The objective of this calculation is to determine the structural response of the standard high-level waste (HLW) canister and the canister containing the cans of immobilized plutonium (Pu) (''can-in-canister'' [CIC] throughout this document) subjected to drop DBEs (design basis events) during the handling operation. The evaluated DBE in the former case is 7-m (23-ft) vertical (flat-bottom) drop. In the latter case, two 2-ft (0.61-m) corner (oblique) drops are evaluated in addition to the 7-m vertical drop. These Pu CIC calculations are performed at three different temperatures: room temperature (RT) (20 C), T = 200 F = 93.3 C , and T = 400 F = 204 C ; in addition to these the calculation characterized by the highest maximum stress intensity is performed at T = 750 F = 399 C as well. The scope of the HLW canister calculation is limited to reporting the calculation results in terms of: stress intensity and effective plastic strain in the canister, directional residual strains at the canister outer surface, and change of canister dimensions. The scope of Pu CIC calculation is limited to reporting the calculation results in terms of stress intensity, and effective plastic strain in the canister. The information provided by the sketches from Reference 26 (Attachments 5.3,5.5,5.8, and 5.9) is that of the potential CIC design considered in this calculation, and all obtained results are valid for this design only. This calculation is associated with the Plutonium Immobilization Project and is performed by the Waste Package Design Section in accordance with Reference 24. It should be noted that the 9-m vertical drop DBE, included in Reference 24, is not included in the objective of this calculation since it did not become a waste acceptance requirement. AP-3.124, ''Calculations'', is used to perform the calculation and develop the document

  5. Final Report Melter Tests With AZ-101 HLW Simulant Using A Duramelter 100 Vitrification System VSL-01R10N0-1, Rev. 1, 2/25/02

    Kruger, A.A.; Matlack, K.S.; Kot, W.K.; Pegg, I.L.

    2011-01-01

    This report provides data, analyses, and conclusions from a series of tests that were conducted at the Vitreous State Laboratory of The Catholic of America (VSL) to determine the processing rates that are achievable with AZ-101 HLW simulants and corresponding melter feeds on a DuraMelter 100 (DM100) vitrification system. One of the most critical pieces of information in determining the required size of the RPP-WTP HLW melter is the specific glass production rate in terms of the mass of glass that can be produced per unit area of melt surface per unit time. The specific glass production rate together with the waste loading (essentially, the ratio of waste-in to glass-out, which is determined from glass formulation activities) determines the melt area that is needed to achieve a given waste processing rate with due allowance for system availability. Tests conducted during Part B1 (VSL-00R2590-2) on the DM1000 vitrification system installed at the Vitreous State Laboratory of The Catholic University of America showed that, without the use of bubblers, glass production rates with AZ-101 and C-106/AY-102 simulants were significantly lower than the Project design basis rate of 0.4 MT/m 2 /d. Conversely, three-fold increases over the design basis rate were demonstrated with the use of bubblers. Furthermore, an un-bubbled control test using a replica of the melter feed used in cold commissioning tests at West Valley reproduced the rates that were observed with that feed on the WVDP production melter. More recent tests conducted on the DM1200 system, which more closely represents the present RPP-WTP design, are in general agreement with these earlier results. Screening tests conducted on the DM10 system have provided good indications of the larger-scale processing rates with bubblers (for both HL W and LAW feeds) but significantly overestimated the DM1000 un-bubbled rate observed for C-106/AY-102 melter feeds. This behavior is believed to be a consequence of the role of

  6. FINAL REPORT MELTER TESTS WITH AZ-101 HLW SIMULANT USING A DURAMELTER 100 VITRIFICATION SYSTEM VSL-01R10N0-1 REV 1 2/25/02

    KRUGER AA; MATLACK KS; KOT WK; PEGG IL

    2011-12-29

    This report provides data, analyses, and conclusions from a series of tests that were conducted at the Vitreous State Laboratory of The Catholic of America (VSL) to determine the processing rates that are achievable with AZ-101 HLW simulants and corresponding melter feeds on a DuraMelter 100 (DM100) vitrification system. One of the most critical pieces of information in determining the required size of the RPP-WTP HLW melter is the specific glass production rate in terms of the mass of glass that can be produced per unit area of melt surface per unit time. The specific glass production rate together with the waste loading (essentially, the ratio of waste-in to glass-out, which is determined from glass formulation activities) determines the melt area that is needed to achieve a given waste processing rate with due allowance for system availability. Tests conducted during Part B1 (VSL-00R2590-2) on the DM1000 vitrification system installed at the Vitreous State Laboratory of The Catholic University of America showed that, without the use of bubblers, glass production rates with AZ-101 and C-106/AY-102 simulants were significantly lower than the Project design basis rate of 0.4 MT/m{sup 2}/d. Conversely, three-fold increases over the design basis rate were demonstrated with the use of bubblers. Furthermore, an un-bubbled control test using a replica of the melter feed used in cold commissioning tests at West Valley reproduced the rates that were observed with that feed on the WVDP production melter. More recent tests conducted on the DM1200 system, which more closely represents the present RPP-WTP design, are in general agreement with these earlier results. Screening tests conducted on the DM10 system have provided good indications of the larger-scale processing rates with bubblers (for both HL W and LAW feeds) but significantly overestimated the DM1000 un-bubbled rate observed for C-106/AY-102 melter feeds. This behavior is believed to be a consequence of the role of

  7. Nitrate glass

    Kirilenko, I.A.; Vinogradov, E.E.

    1977-01-01

    Experimental evidence on behaviour of nitrate glasses is reviewed in terms of relationships between the presence of water in vitrescent nitrate systems and the properties of the systems. The glasses considered belong to systems of Mg(NO 3 ) 2 - Nd(NO 3 ) 3 ; Hg(NO 3 ) 2 -Nd(NO 3 ) 3 ; NaNO 3 -Mg(NO 3 ) 2 -Nd(NO 3 ) 3 ; M-Zn(NO 3 ) 3 , where M is a mixture of 20% mass NaNO 3 and 80% mass Mg(NO 3 ) 2 , and Zn is a rare earth ion. Nitrate glass is shown to be a product of dehydration. Vitrification may be regarded as a resusl of formation of molecular complexes in the chain due to hydrogen bonds of two types, i.e. water-water, or water-nicrate group. Chain formation, along with low melting points of the nitrates, hinder crystallization of nitrate melts. Provided there is enough water, this results in vitrification

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

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

    1986-09-01

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

  9. Simulation of HTM processes in buffer-rock barriers based on the French HLW disposal concept

    Li, Xiaoshuo; Roehlig, Klaus-Juergen; Zhang, Chunliang

    2012-01-01

    Document available in extended abstract form only. The main objectives of this paper are to gain experience with modelling and analysis of HTM processes in clay rock and bentonite buffer surrounding heat-generating radioactive waste. The French concept for HLW disposal in drifts with backfilled bentonite buffer considered in numerical calculations which are carried out by using the computer code CODE-BRIGHT developed by the Technical University of Catalonia in Barcelona. The French repository designed by ANDRA is located in the middle of the Callovo-Oxfordian argillaceous formation (COX) of 250 m thickness at a depth of 500 to 630 m below the surface. The French concept has been simplified at this simulation work. A drift is considered to be excavated at a depth of 500 m below the surface. It has a diameter of 2.2 m and a length of 20 m. A large volume of the rock mass around the drift is taken into account by an axisymmetric model of 100 m radius and 100 m length. In fact, this model represents a cylindrical rock-buffer-system with the central axis of the containers, as shown in Figure 1. Some points are selected in the buffer and the rock along the radial line (dash yellow line) in the middle of the drift for recording HTM parameters with time. The display and analysis of the results at this paper are chiefly along this line. The simulation work has been divided to two time steps. At the first step, the drift excavation and ventilation is simulated by reducing the stress normal to the drift wall down to zero and circulating gas along the drift wall with relative humidity of 85 %. Following the drift excavation and ventilation, the HLW containers and the bentonite are emplaced in the drift as the second step of the simulation. This is simulated by simultaneously applying the initial conditions of the buffer and the decayed heat emitting from the waste containers as thermal boundary conditions. Two materials (Clay rock and bentonite buffer) are taken into account

  10. Discussing compliance. Summary report from discussions with Robert Bernero and Chris Whipple regarding compliance with the Swedish HLW Regulations from meetings in Stockholm May 3 and 4, 1999

    Jensen, Mikael

    1999-06-01

    Summary report from discussions with Robert Bernero and Chris Whipple regarding compliance with the Swedish HLW Regulations from meetings in Stockholm. The report also contains bibliographical information and preliminary observations made by Robert Bernero and Chris Whipple.

  11. Discussing compliance. Summary report from discussions with Robert Bernero and Chris Whipple regarding compliance with the Swedish HLW Regulations from meetings in Stockholm May 3 and 4, 1999

    Jensen, Mikael

    1999-06-01

    Summary report from discussions with Robert Bernero and Chris Whipple regarding compliance with the Swedish HLW Regulations from meetings in Stockholm. The report also contains bibliographical information and preliminary observations made by Robert Bernero and Chris Whipple

  12. A Strategy for Maintenance of the Long-Term Performance Assessment of Immobilized Low-Activity Waste Glass

    Ryan, Joseph V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Freedman, Vicky L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-28

    Approximately 50 million gallons of high-level radioactive mixed waste has accumulated in 177 buried single- and double-shell tanks at the Hanford Site in southeastern Washington State as a result of the past production of nuclear materials, primarily for defense uses. The United States Department of Energy (DOE) is proceeding with plans to permanently dispose of this waste. Plans call for separating the tank waste into high-level waste (HLW) and low-activity waste (LAW) fractions, which will be vitrified at the Hanford Waste Treatment and Immobilization Plant (WTP). Principal radionuclides of concern in LAW are 99Tc, 129I, and U, while non-radioactive contaminants of concern are Cr and nitrate/nitrite. HLW glass will be sent off-site to an undetermined federal site for deep geological disposal while the much larger volume of immobilized low-activity waste will be placed in the on-site, near-surface Integrated Disposal Facility (IDF).

  13. HLW Feed Delivery AZ101 Batch Transfer to the Private Contractor Transfer and Mixing Process Improvements [Initial Release at Rev 2

    DUNCAN, G.P.

    2000-02-28

    The primary purpose of this business case is to provide Operations and Maintenance with a detailed transfer process review for the first High Level Waste (HLW) feed delivery to the Privatization Contractor (PC), AZ-101 batch transfer to PC. The Team was chartered to identify improvements that could be implemented in the field. A significant penalty can be invoked for not providing the quality, quantity, or timely delivery of HLW feed to the PC.

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

    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

  15. Safety studies of HLW-disposal in the Mors salt dome - Support to the salt option of the Pagis project

    Lindstroem Jensen, K.E.

    1987-01-01

    The study, which is a support to the Pagis project, covers three tasks concerning the evaluation of the Danish salt dome Mors (variant disposal site): evaluation of the human intrusion scenario where a cavern is excavated near the HLW-repository by solution mining technique. The waste is supposed to be leached during the operation period until the abandoned cavern is closed by convergence and the contaminated brine is pressed up into the overburden. Evaluation of the brine intrusion scenario, where the HLW-repository is inadvertently located close to a major brine pocket which subsequently releases its brine content through defects in the repository to the discharge stream for the catchment area. Collection and description of hydrological data of surface and deep layers (down to circa 700 metres) in the repository region. The data will be used by GSF to calculate the radionuclide migration in the geosphere

  16. Regional Geologic Evaluations for Disposal of HLW and SNF: The Pierre Shale of the Northern Great Plains

    Perry, Frank Vinton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kelley, Richard E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-09-14

    The DOE Spent Fuel and Waste Technology (SWFT) R&D Campaign is supporting research on crystalline rock, shale (argillite) and salt as potential host rocks for disposal of HLW and SNF in a mined geologic repository. The distribution of these three potential repository host rocks is limited to specific regions of the US and to different geologic and hydrologic environments (Perry et al., 2014), many of which may be technically suitable as a site for mined geologic disposal. This report documents a regional geologic evaluation of the Pierre Shale, as an example of evaluating a potentially suitable shale for siting a geologic HLW repository. This report follows a similar report competed in 2016 on a regional evaluation of crystalline rock that focused on the Superior Province of the north-central US (Perry et al., 2016).

  17. Cavern disposal concepts for HLW/SF: assuring operational practicality and safety with maximum programme flexibility

    McKinley, Ian G.; Apted, Mick; Umeki, Hiroyuki; Kawamura, Hideki

    2008-01-01

    Most conventional engineered barrier system (EBS) designs for HLW/SF repositories are based on concepts developed in the 1970s and 1980s that assured feasibility with high margins of safety, in order to convince national decision makers to proceed with geological disposal despite technological uncertainties. In the interval since the advent of such 'feasibility designs', significant progress has been made in reducing technological uncertainties, which has lead to a growing awareness of other, equally important uncertainties in operational implementation and challenges regarding social acceptance in many new, emerging national repository programs. As indicated by the NUMO repository concept catalogue study (NUMO, 2004), there are advantages in reassessing how previous designs can be modified and optimised in the light of improved system understanding, allowing a robust EBS to be flexibly implemented to meet nation-specific and site-specific conditions. Full-scale emplacement demonstrations, particularly those carried out underground, have highlighted many of the practical issues to be addressed; e.g., handling of compacted bentonite in humid conditions, use of concrete for support infrastructure, remote handling of heavy radioactive packages in confined conditions, quality inspection, monitoring / ease of retrieval of emplaced packages and institutional control. The CAvern REtrievable (CARE) concept reduces or avoids such issues by emplacement of HLW or SF within multi-purpose transportation / storage / disposal casks in large ventilated caverns at a depth of several hundred metres. The facility allows the caverns to serve as inspectable stores for an extended period of time (up to a few hundred years) until a decision is made to close them. At this point the caverns are backfilled and sealed as a final repository, effectively with the same safety case components as conventional 'feasibility designs'. In terms of operational practicality an d safety, the CARE

  18. Determination of the corrosion mechanisms of high level waste containing glass

    Conradt, R.; Roggendorf, H.

    1985-01-01

    The purpose of the reported work was to determine the corrosion behaviour of the inactive HLW glass SM 58 LW 11 in Q-solution at temperatures up to 200 0 C and elevated pressures up to 13 MPa. In particular, a parametric study on the effects of time, temperature, pressure, crystallization, metallic impurities a.o. was performed. Further tests helped to identify the rate determining steps in the entire process and the most likely long-term corrosion law. (orig./RB)

  19. Alkali-free bioactive glasses for bone regeneration

    Kapoor, Saurabh

    2014-01-01

    Bioactive glasses and glass-ceramics are a class of third generation biomaterials which elicit a special response on their surface when in contact with biological fluids, leading to strong bonding to living tissues. The purpose of the present study was to develop diopside based alkali-free bioactive glasses in order to achieve good sintering behaviour, high bioactivity, and a dissolution/ degradation rates compatible with the target applications in bone regeneration and tiss...

  20. Strontium borate glass: potential biomaterial for bone regeneration

    Pan, H. B.; Zhao, X. L.; Zhang, X.; Zhang, K. B.; Li, L. C.; Li, Z. Y.; Lam, W. M.; Lu, W. W.; Wang, D. P.; Huang, W. H.; Lin, K. L.; Chang, J.

    2009-01-01

    Boron plays important roles in many life processes including embryogenesis, bone growth and maintenance, immune function and psychomotor skills. Thus, the delivery of boron by the degradation of borate glass is of special interest in biomedical applications. However, the cytotoxicity of borate glass which arises with the rapid release of boron has to be carefully considered. In this study, it was found that the incorporation of strontium into borate glass can not only moderate the rapid relea...

  1. Design and validation of the THMC China-Mock-Up test on buffer material for HLW disposal

    Yuemiao Liu

    2014-04-01

    Full Text Available According to the preliminary concept of the high-level radioactive waste (HLW repository in China, a large-scale mock-up facility, named China-Mock-Up was constructed in the laboratory of Beijing Research Institute of Uranium Geology (BRIUG. A heater, which simulates a container of radioactive waste, is placed inside the compacted Gaomiaozi (GMZ-Na-bentonite blocks and pellets. Water inflow through the barrier from its outer surface is used to simulate the intake of groundwater. The numbers of water injection pipes, injection pressure and the insulation layer were determined based on the numerical modeling simulations. The current experimental data of the facility are herein analyzed. The experiment is intended to evaluate the thermo-hydro-mechano-chemical (THMC processes occurring in the compacted bentonite-buffer during the early stage of HLW disposal and to provide a reliable database for numerical modeling and further investigation of engineered barrier system (EBS, and the design of HLW repository.

  2. Development of geological disposal system; localization of element cost data and cost evaluation on the HLW repository

    Lee, Byung Sik; Kim, Kil Jung; Yang, Young Jin; Kim, Sung Chun [KOPEC, Taejeon (Korea)

    2002-03-01

    To estimate Total Life Cycle Cost (TSLCC) for Korea HLW Repository through localization of element cost data, we review and re-organize each basic element cost data for reference repository system, localize various element cost and finally estimate TSLCC considering economic parameters. As results of the study, TSLCC is estimated as 17,167,689 million won, which includes costs for site preparation, surface facilities, underground facilities and management/integration. Since HLW repository Project is an early stage of pre-conceptual design at present, the information of design and project information are not enough to perform cost estimate and cost localization for the Project. However, project cost structure is re-organized based on the local condition and Total System Life Cycle Cost is estimated using the previous cost data gathered from construction experience of the local nuclear power plant. Project results can be used as basic reference data to assume total construction cost for the local HLW repository and should be revised to more reliable cost data with incorporating detail project design information into the cost estimate in a future. 20 refs. (Author)

  3. Study on a Preliminary Survey and Analysis of HLW Management Technology Suitable for Nuclear Industrial Environment in Korea

    Kim, Eun Ka; Suh, In Suk; Ro, Seong Gy; Yoo, Kun Joong; Yoo, Jae Hyung; Cho, Sung Soo

    2010-12-01

    The purpose of this study is to suggest development direction of related technologies to analyze patented technology filed as a leading technology and to identify the technology trend for developing HLW management technology suitable for atomic industrial environment in Korea. For patent analysis of HLW management technology, international patent data were collected. And international application number, patent share of applicant and nationality, annual number of applications, application trends of assignees and detail technology, and frequency of patent citations / citations-to were analyzed by statistical analysis. Technical level and competitiveness through quantitative analysis by indicators of patent analysis were confirmed. And technology developments of blank technology, similarity analysis, the point of the main patent and a range of patent rights were analyzed through in-depth analysis. Trends of the patented technology of our country and world patent technology in such results have been identified, and statistical data on patents were secured. Especially in HLW management technology, patent application in Korea compared ti United States, Japan and European Union was began much later for the '90s, and are showing the annual increase on trend of patent application. Patent trend in Korea corresponds to development generation, while declining in foreign patent. The result of this study will be usefully applied to setting a development direction and blank technology of patent technology to pursue future in Korea

  4. CRYSTALLIZATION IN MULTICOMPONENT GLASSES

    KRUGER AA; HRMA PR

    2009-10-08

    In glass processing situations involving glass crystallization, various crystalline forms nucleate, grow, and dissolve, typically in a nonuniform temperature field of molten glass subjected to convection. Nuclear waste glasses are remarkable examples of multicomponent vitrified mixtures involving partial crystallization. In the glass melter, crystals form and dissolve during batch-to-glass conversion, melter processing, and product cooling. Crystals often agglomerate and sink, and they may settle at the melter bottom. Within the body of cooling glass, multiple phases crystallize in a non-uniform time-dependent temperature field. Self-organizing periodic distribution (the Liesegnang effect) is common. Various crystallization phenomena that occur in glass making are reviewed.

  5. Crystallization In Multicomponent Glasses

    Kruger, A.A.; Hrma, P.R.

    2009-01-01

    In glass processing situations involving glass crystallization, various crystalline forms nucleate, grow, and dissolve, typically in a nonuniform temperature field of molten glass subjected to convection. Nuclear waste glasses are remarkable examples of multicomponent vitrified mixtures involving partial crystallization. In the glass melter, crystals form and dissolve during batch-to-glass conversion, melter processing, and product cooling. Crystals often agglomerate and sink, and they may settle at the melter bottom. Within the body of cooling glass, multiple phases crystallize in a non-uniform time-dependent temperature field. Self-organizing periodic distribution (the Liesegnang effect) is common. Various crystallization phenomena that occur in glass making are reviewed.

  6. Time evolution of dissolved oxygen and redox conditions in a HLW repository

    Wersin, P.; Spahiu, K.; Bruno, J.

    1994-02-01

    The evolution of oxygen in a HLW repository has been studied using presently available geochemical background information. The important processes affecting oxygen migration in the near-field include diffusion and oxidation of pyrite and dissolved Fe(II). The evaluation of time scales of oxygen decrease is carried out with 1. an analytical approach involving the coupling of diffusion and chemical reaction, 2. a numerical geochemical approach involving the application of a newly developed diffusion-extended version of the STEADYQL code. Both approaches yield consistent rates of oxygen decrease and indicate that oxidation of pyrite impurities in the clay is the dominant process. The results obtained fRom geochemical modelling are interpreted in terms of evolution of redox conditions. Moreover, a sensitivity analysis of the major geochemical and physical parameters is performed. These results indicate that the uncertainties associated with reactive pyrite surface area impose the overall uncertainties of prediction of time scales. Thus, the obtained time of decrease to 1% of initial O 2 concentrations range between 7 and 290 years. The elapsed time at which the transition to anoxic conditions occurs is estimated to be within the same time range. Additional experimental information on redox sensitive impurities in the envisioned buffer and backfill material would further constrain the evaluated time scales. 41 refs

  7. Use of Gap-fills in the Buffer and Backfill of an HLW Repository

    Lee, Jae Owan; Lee, Min Soo; Choi, Heui Joo [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The buffer and backfill are significant barrier components of the repository. They play the roles of preventing the inflow of groundwater from the surrounding rock, retarding the release of radionuclides from the waste, supporting disposal container against external impacts, and discharging decay heat from the waste. When the buffer and backfill are installed for the HLW repository, there may be gaps between the container and buffer and between the backfill and the wall of disposal tunnels, respectively. These gaps occur because spaces are allowed for ease of the installation of the buffer and backfill in excavated deposition boreholes and disposal tunnels. If the gaps are left without any sealing as they are, however, the buffer and backfill can't accomplish their functions as the barrier components. This paper reviews the gap-fill concepts of the developed foreign countries, and then suggests a gap-fill concept which is applicable for the KRS. The gap-fill is suggested to employ bentonite- based materials with a type of pellet, granule, and pellet-granule mixture. The roller compression method and extrusion-cutting method are applicable for the fabrication of the bentonite pellets which can have the high density and the required amount for use to the buffer and backfill. For the installation of the gap-fill, the pouring and then pressing method and the shotcrete- blowing method are preferable for the gap of the deposition borehole and the gap of the disposal tunnel, respectively.

  8. Performance Assessment Uncertainty Analysis for Japan's HLW Program Feasibility Study (H12)

    BABA, T.; ISHIGURO, K.; ISHIHARA, Y.; SAWADA, A.; UMEKI, H.; WAKASUGI, K.; WEBB, ERIK K.

    1999-01-01

    Most HLW programs in the world recognize that any estimate of long-term radiological performance must be couched in terms of the uncertainties derived from natural variation, changes through time and lack of knowledge about the essential processes. The Japan Nuclear Cycle Development Institute followed a relatively standard procedure to address two major categories of uncertainty. First, a FEatures, Events and Processes (FEPs) listing, screening and grouping activity was pursued in order to define the range of uncertainty in system processes as well as possible variations in engineering design. A reference and many alternative cases representing various groups of FEPs were defined and individual numerical simulations performed for each to quantify the range of conceptual uncertainty. Second, parameter distributions were developed for the reference case to represent the uncertainty in the strength of these processes, the sequencing of activities and geometric variations. Both point estimates using high and low values for individual parameters as well as a probabilistic analysis were performed to estimate parameter uncertainty. A brief description of the conceptual model uncertainty analysis is presented. This paper focuses on presenting the details of the probabilistic parameter uncertainty assessment

  9. Strategy for safety case development: impact of a volunteering approach to siting a japanese HLW repository

    Kitayama, K.; Ishiguro, K.; Takeuchi, M.; Tsuchi, H.; Kato, T.; Sakabe, Y.; Wakasugi, K.

    2008-01-01

    NUMO strategy for safety case development is constrained by a staged siting approach, which has been initiated by a call for volunteer municipalities to host the HLW repository. For each site, the safety case is an important factor to be considered at the selection steps which narrow down towards the preferred repository location. This is particularly challenging, however, as every site requires a tailored repository concept, with associated performance assessment and an individual site evaluation programme all of which evolve with gradually increasing understanding of the host environment. In order to maintain flexibility without losing focus, NUMO has developed a formalized tailoring procedure, termed the NUMO Structured Approach (NSA). The NSA guides the interaction of the key site characterisation, repository design and performance assessment groups and is facilitated by tools to help the decision making associated with the tailoring process (e.g. a requirements management system) and with comparison of siting and design options (e.g. multi-attribute analysis). Pragmatically, the post-closure safety case will initially emphasize near-field processes and a robust engineering barrier system, considering the limited geological information at early stages. This will be complemented by a more realistic assessment of total system performance, as needed to compare options. In addition, efforts to rigorously assess operational phase safety and the practicality of assuring quality of the constructed engineered barriers are components of the total safety case which are receiving particular attention now, as they may better discriminate between sites while information is still limited. (authors)

  10. Cleanup of a HLW nuclear fuel-reprocessing center using 3-D database modeling technology

    Sauer, R.C.

    1992-01-01

    A significant challenge in decommissioning any large nuclear facility is how to solidify the large volume of residual high-level radioactive waste (HLW) without structurally interfering with the existing equipment and piping used at the original facility or would require rework due to interferences which were not identified during the design process. This problem is further compounded when the nuclear facility to be decommissioned is a 35 year old nuclear fuel reprocessing center designed to recover usable uranium and plutonium. Facilities of this vintage usually tend to lack full documentation of design changes made over the years and as a result, crude traps or pockets of high-level contamination may not be fully realized. Any miscalculation in the construction or modification sequences could compound the overall dismantling and decontamination of the facility. This paper reports that development of a 3-dimensional (3-D) computer database tool was considered critical in defining the most complex portions of this one-of-a-kind vitrification facility

  11. Nuclide release calculation in the near-field of a reference HLW repository

    Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung

    2004-01-01

    The HLW-relevant R and D program for disposal of high-level radioactive waste has been carried out at Korea Atomic Energy Research Institute (KAERI) since early 1997 in order to develop a conceptual Korea Reference Repository System for direct disposal of nuclear spent fuel by the end of 2007. A preliminary reference geologic repository concept considering such established criteria and requirements as waste and generic site characteristics in Korea was roughly envisaged in 2003 focusing on the near-field components of the repository system. According to above basic repository concept, which is similar to that of Swedish KBS-3 repository, the spent fuel is first encapsulated in corrosion resistant canisters, even though the material has not yet been determined, and then emplaced into the deposition holes surrounded by high density bentonite clay in tunnels constructed at a depth of about 500 m in a stable plutonic rock body. Not only to demonstrate how much a reference repository is safe in the generic point of view with several possible scenarios and cases associated with a preliminary repository concept by conducting calculations for nuclide release and transport in the near-field components of the repository, even though enough information has not been available that much yet, but also to show a methodology by which a generic safety assessment could be performed for further development of Korea reference repository concept, nuclide release calculation study strongly seems to be necessary

  12. Biosphere Modeling for the Dose Assessment of a HLW Repository: Development of ACBIO

    Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung

    2006-01-15

    For the purpose of evaluating a dose rate to an individual due to a long-term release of nuclides from a HLW repository, a biosphere assessment model and an implemented code, ACBIO, based on the BIOMASS methodology have been developed by utilizing AMBER, a general compartment modeling tool. To demonstrate its practicability and usability as well as to observe the sensitivity of the compartment scheme, the concentration, the activity in the compartments as well as the annual flux between the compartments at their peak values, were calculated and investigated. For each case when changing the structure of the compartments and GBIs as well as varying selected input Kd values, all of which seem very important among the others, the dose rate per nuclide release rate is calculated separately and analyzed. From the maximum dose rates, the flux to dose conversion factors for each nuclide were derived, which are used for converting the nuclide release rate appearing from the geosphere through various GBIs to dose rates (Sv/y) for an individual in a critical group. It has also been observed that the compartment scheme, the identification of a possible exposure group and the GBIs could all be highly sensitive to the final consequences in a biosphere modeling.

  13. Storage of HLW in engineered structures: air-cooled and water-cooled concepts

    Ahner, S.; Dekais, J.J.; Puttke, B.; Staner, P.

    1981-01-01

    A comparative study on an air-cooled and a water-cooled intermediate storage of vitrified, highly radioactive waste (HLW) in overground installations has been performed by Nukem and Belgonucleaire respectively. In the air-cooled storage concept the decay heat from the storage area will be removed using natural convection. In the water-cooled storage concept the decay heat is carried off by a primary and secondary forced-cooling system with redundant and diverse devices. The safety study carried out by Nukem used a fault tree method. It shows that the reliability of the designed water-cooled system is very high and comparable to the inherent, safe, air-cooled system. The impact for both concepts on the environment is determined by the release route, but even during accident conditions the release is far below permissible limits. The economic analysis carried out by Belgonucleaire shows that the construction costs for both systems do not differ very much, but the operation and maintenance costs for the water-cooled facility are higher than for the air cooled facility. The result of the safety and economic analysis and the discussions with the members of the working group have shown some possible significant modifications for both systems, which are included in this report. The whole study has been carried out using certain national criteria which, in certain Member States at least, would lead to a higher standard of safety than can be justified on any social, political or economic grounds

  14. Development of a computer tool to support scenario analysis for safety assessment of HLW geological disposal

    Makino, Hitoshi; Kawamura, Makoto; Wakasugi, Keiichiro; Okubo, Hiroo; Takase, Hiroyasu

    2007-02-01

    In 'H12 Project to Establishing Technical Basis for HLW Disposal in Japan' a systematic approach that was based on an international consensus was adopted to develop scenarios to be considered in performance assessment. Adequacy of the approach was, in general term, appreciated through the domestic and international peer review. However it was also suggested that there were issues related to improving transparency and traceability of the procedure. To achieve this, improvement of scenario analysis method has been studied. In this study, based on an improvement method for treatment of FEP interaction a computer tool to support scenario analysis by specialists of performance assessment has been developed. Anticipated effects of this tool are to improve efficiency of complex and time consuming scenario analysis work and to reduce possibility of human errors in this work. This tool also enables to describe interactions among a vast number of FEPs and the related information as interaction matrix, and analysis those interactions from a variety of perspectives. (author)

  15. Biosphere Modeling for the Dose Assessment of a HLW Repository: Development of ACBIO

    Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung

    2006-01-01

    For the purpose of evaluating a dose rate to an individual due to a long-term release of nuclides from a HLW repository, a biosphere assessment model and an implemented code, ACBIO, based on the BIOMASS methodology have been developed by utilizing AMBER, a general compartment modeling tool. To demonstrate its practicability and usability as well as to observe the sensitivity of the compartment scheme, the concentration, the activity in the compartments as well as the annual flux between the compartments at their peak values, were calculated and investigated. For each case when changing the structure of the compartments and GBIs as well as varying selected input Kd values, all of which seem very important among the others, the dose rate per nuclide release rate is calculated separately and analyzed. From the maximum dose rates, the flux to dose conversion factors for each nuclide were derived, which are used for converting the nuclide release rate appearing from the geosphere through various GBIs to dose rates (Sv/y) for an individual in a critical group. It has also been observed that the compartment scheme, the identification of a possible exposure group and the GBIs could all be highly sensitive to the final consequences in a biosphere modeling

  16. Stress analysis of HLW containers. Preliminary ring test exercise Compas project

    1989-01-01

    This document describes the series of experiments and associated calculations performed as the Compas preliminary ring test exercise. A number of mild steel rings, representative of sections through HLW containers, some notched and pre-cracked, were tested in compression right up to and beyond their ultimate load. The Compas project partners independently modelled the behaviour of these rings using their finite element codes. Four different ring types were tested, and each test was repeated three times. For three of the ring types, the three test repetitions gave identical results. The fourth ring, which was not modelled by the partners, had a 4 mm thick layer of weld metal deposited on its surface. The three tests on this ring did not give identical results and suggested that the effect of welding methods should be addressed at a later stage of the project. Fracture was not found to be a significant cause of ring failure. The results of the ring tests were compared with the partners predictions, and additionally some time was spent assessing where the use of the codes could be improved. This exercise showed that the partners codes have the ability to produce results within acceptable limits. Most codes were unable to model stable crack growth. There were indications that some codes would not be able to cope with a significantly more complex three-dimensional analysis

  17. Seismic design evaluation guidelines for buried piping for the DOE HLW Facilities

    Lin, Chi-Wen; Antaki, G.; Bandyopadhyay, K.; Bush, S.H.; Costantino, C.; Kennedy, R.

    1995-01-01

    This paper presents the seismic design and evaluation guidelines for underground piping for the Department of Energy (DOE) High-Level-Waste (HLW) Facilities. The underground piping includes both single and double containment steel pipes and concrete pipes with steel lining, with particular emphasis on the double containment piping. The design and evaluation guidelines presented in this paper follow the generally accepted beam-on-elastic-foundation analysis principle and the inertial response calculation method, respectively, for piping directly in contact with the soil or contained in a jacket. A standard analysis procedure is described along with the discussion of factors deemed to be significant for the design of the underground piping. The following key considerations are addressed: the design feature and safety requirements for the inner (core) pipe and the outer pipe; the effect of soil strain and wave passage; assimilation of the necessary seismic and soil data; inertial response calculation for the inner pipe; determination of support anchor movement loads; combination of design loads; and code comparison. Specifications and justifications of the key parameters used, stress components to be calculated and the allowable stress and strain limits for code evaluation are presented

  18. Environmental risk assessment: its contribution to criteria development for HLW disposal

    Smith, G.M.; Little, R.H.; Watkins, B.M.

    1999-01-01

    Principles for radioactive waste management have been provided by the International Atomic Energy Agency in Safety Series No.111-F, which was published in 1995. This has been a major step forward in the process of achieving acceptance for proposals for disposal of radioactive waste, for example, for High Level Waste disposal in deep repositories. However, these principles have still to be interpreted and developed into practical radiation protection criteria. Without prejudicing final judgements on the acceptability of waste proposals, an important aspect is that practical demonstration of compliance (or the opposite) with these criteria must be possible. One of the IAEA principles requires that radioactive waste shall be managed in such a way as to provide an acceptable level of protection of the environment. There has been and continues to be considerable debate as to how to demonstrate compliance with such a principle. This paper briefly reviews the current status and considers how experience in other areas of environmental protection could contribute to criteria development for HLW disposal

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

    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

  20. Effect of irradiation on the evolution of alteration layer formed during nuclear glass leaching

    Mougnaud, Sarah

    2016-01-01

    High-level radioactive waste (HLW) remaining after spent nuclear fuel reprocessing is immobilized within a glass matrix, eventually destined for geological disposal. Water intrusion into the repository is expected after several thousand years. The alteration of a non-radioactive surrogate for nuclear glass has been extensively studied and it has been determined that successive leaching mechanisms lead to the formation of a 'passivating' alteration layer and to the establishment of a residual rate regime in the long term. However, glass packages are submitted to the radioactivity of confined radioelements. This work focuses on the influence of irradiation on the alteration layer formed during the residual rate regime, in a structural and mechanistic point of view. Three focal areas have been selected. Non-radioactive simple glasses have been leached and externally irradiated in order to determine modifications induced by electronic effects (irradiations with electrons and alpha particles). The same type of glass samples have been previously irradiated with heavy ions and their leaching behavior have been studied in order to assess the impact of ballistic dose cumulated by the glass before water intrusion. Leaching behavior of a complex radioactive glass, doped with an alpha-emitter, has been studied to consider a more realistic situation. (author) [fr

  1. Recycling of Glass

    Christensen, Thomas Højlund; Damgaard, Anders

    2011-01-01

    Glass is used for many purposes, but in the waste system glass is predominantly found in terms of beverage and food containers with a relatively short lifetime before ending up in the waste. Furthermore there is a large amount of flat glass used in building materials which also ends up in the waste...... system; this glass though has a long lifetime before ending up in the waste. Altogether these product types add up to 82% of the production of the European glass industry (IPCC, 2001). Recycling of glass in terms of cleaning and refilling of bottles as well as the use of broken glass in the production...... of new glass containers is well established in the glass industry. This chapter describes briefly howglass is produced and howwaste glass is recycled in the industry. Quality requirements and use of recycled products are discussed, as are the resource and environmental issues of glass recycling....

  2. A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

    Sevougian, S. David [Advanced Nuclear Energy Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); MacKinnon, Robert J. [Advanced Nuclear Energy Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Leigh, Christi D. [Defense Waste Management Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Hansen, Frank D. [Geoscience Research and Applications Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States)

    2013-07-01

    The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential

  3. A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

    Sevougian, S. David; MacKinnon, Robert J.; Leigh, Christi D.; Hansen, Frank D.

    2013-01-01

    The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential

  4. Preliminary report on a glass burial experiment in granite

    Clark, D.E.; Zhu, B.F.; Robinson, R.S.; Wicks, G.G.

    1983-01-01

    Preliminary results of a two-year burial experiment in granite are discussed. Three compositions of simulated alkali borosilicate waste glasses were placed in boreholes approximately 350 meters deep. The glass sample configurations include mini-cans (stainless steel rings into which glass has been cast) and pineapple slices (thin sections from cylindrical blocks). Assemblies of these glass samples were prepared by stacking them together with granite, compacted bentonite and metal rings to provide several types of interfaces that are expected to occur in the repository. The assemblies were maintained at either ambient mine temperature (8 to 10 0 C) or 90 0 C. The glasses were analyzed before burial and after one month storage at 90 0 C. The most extensive surface degradation occurred on the glasses interfaced with bentonite. In general, very little attack was observed on glass surfaces in contact with the other materials. The limited field and laboratory data are compared

  5. Alkali-free bioactive glasses for bone regeneration =

    Kapoor, Saurabh

    Bioactive glasses and glass-ceramics are a class of third generation biomaterials which elicit a special response on their surface when in contact with biological fluids, leading to strong bonding to living tissues. The purpose of the present study was to develop diopside based alkali-free bioactive glasses in order to achieve good sintering behaviour, high bioactivity, and a dissolution/ degradation rates compatible with the target applications in bone regeneration and tissue engineering. Another aim was to understand the structure-property relationships in the investigated bioactive glasses. In this quest, various glass compositions within the Diopside (CaMgSi2O6) - Fluorapatite (Ca5(PO4)3F) - Tricalcium phosphate (3CaO•P2O5) system have been investigated. All the glasses were prepared by melt-quenching technique and characterized by a wide array of complementary characterization techniques. The glass-ceramics were produced by sintering of glass powders compacts followed by a suitable heat treatment to promote the nucleation and crystallization phenomena. Furthermore, selected parent glass compositions were doped with several functional ions and an attempt to understand their effects on the glass structure, sintering ability and on the in vitro bio-degradation and biomineralization behaviours of the glasses was made. The effects of the same variables on the devitrification (nucleation and crystallization) behaviour of glasses to form bioactive glass-ceramics were also investigated. Some of the glasses exhibited high bio-mineralization rates, expressed by the formation of a surface hydroxyapatite layer within 1-12 h of immersion in a simulated body fluid (SBF) solution. All the glasses showed relatively lower degradation rates in comparison to that of 45S5 Bioglass. Some of the glasses showed very good in vitro behaviour and the glasses co-doped with zinc and strontium showed an in vitro dose dependent behaviour. The as-designed bioactive glasses and glass

  6. Spin glasses

    Mookerjee, Abhijit

    1976-01-01

    ''Spin glasses'', are entire class of magnetic alloys of moderate dilution, in which the magnetic atoms are far enough apart to be unlike the pure metal, but close enough so that the indirect exchange energy between them (mediated by the s-d interaction between local moments and conduction electrons) dominates all other energies. Characteristic critical phenomena displayed such as freezing of spin orientation at 'Tsub(c)' and spreading of magnetic ordering, are pointed out. Anomalous behaviour, associated with these critical phenomena, as reflected in : (i) Moessbauer spectroscopy giving hyperfine splitting at Tsub(c), (ii) maxima in susceptibility and remanent magnetism, (iii) thermopower maxima and change in slope, (iv) Characteristic cusp in susceptibility and its removal by very small magnetic fields, and (v) conductivity-resistivity measurements, are discussed. Theoretical developments aimed at explaining these phenomena, in particular, the ideas from percolation and localisation theories, and the approach based on the gellations of polymers, are discussed. Finally, a new approach based on renormalisation group in disordered systems is also briefly mentioned. (K.B.)

  7. Methodology of fuel cycles long-term safety assessment of SNF/HLW geological disposal

    Pritrsky, J.

    2008-02-01

    Methodology for the long-term safety assessment of nuclear fuel cycles is given in the presented doctoral thesis. The aim of work was to develop a geological repository model for disposal of spent nuclear fuel (SNF) and high level waste (HLW) using an appropriate software code able to calculate the influence of partitioning and transmutation in advanced fuel cycles. The first step in this process was specifying of indicators which can be used to quantify the radiological impact of each fuel cycle. Indicators such as annual effective dose and radiotoxicity of inventory have been quantitatively analysed to determine the potential risk and radiological consequences associated with production of SNF/HLW. Advanced fuel types bring a number of advantages in comparison to uranium oxide fuel UO 2 used worldwide nowadays in terms of safety improvement due to minor actinides transmutation and non-proliferation aspects as well. Within the scope of work, three different fuel cycles are compared from the point of view of long-term safety of deep geological repository. The first considered fuel cycle is the currently used open fuel cycle (UOX) which uses only U-FA (Uranium Fuel Assembly). The second assessed cycle is a closed fuel cycle (MOX) with MOX-FA (Mixed OXides Fuel Assembly) and the third considered one is a partially closed fuel cycle (IMF) with IMC-FA (Inert Matrix Combined Fuel Assembly). Description and input data of advanced fuel cycles have been gained by participation in the EC project RED-IMPACT. Results were calculated using code AMBER, which is a flexible software tool that allows building dynamic compartmental models to represent the migration and fate of contaminants in a system, for example in the surface and sub-surface environment. Contaminants in solid, liquid and gaseous phases can be considered. AMBER gives the user the flexibility to define any number of compartments; any number of contaminants and associated decays; deterministic, probabilistic and

  8. Methodology of fuel cycles long-term safety assessment of SNF/HLW geological disposal

    Pritrsky, J.

    2008-01-01

    Methodology for the long-term safety assessment of nuclear fuel cycles is given in the presented doctoral thesis. The aim of work was to develop a geological repository model for disposal of spent nuclear fuel (SNF) and high level waste (HLW) using an appropriate software code able to calculate the influence of partitioning and transmutation in advanced fuel cycles. The first step in this process was specifying of indicators which can be used to quantify the radiological impact of each fuel cycle. Indicators such as annual effective dose and radiotoxicity of inventory have been quantitatively analysed to determine the potential risk and radiological consequences associated with production of SNF/HLW. Advanced fuel types bring a number of advantages in comparison to uranium oxide fuel UO 2 used worldwide nowadays in terms of safety improvement due to minor actinides transmutation and non-proliferation aspects as well. Within the scope of work, three different fuel cycles are compared from the point of view of long-term safety of deep geological repository. The first considered fuel cycle is the currently used open fuel cycle (UOX) which uses only U-FA (Uranium Fuel Assembly). The second assessed cycle is a closed fuel cycle (MOX) with MOX-FA (Mixed OXides Fuel Assembly) and the third considered one is a partially closed fuel cycle (IMF) with IMC-FA (Inert Matrix Combined Fuel Assembly). Description and input data of advanced fuel cycles have been gained by participation in the EC project RED-IMPACT. Results were calculated using code AMBER, which is a flexible software tool that allows building dynamic compartmental models to represent the migration and fate of contaminants in a system, for example in the surface and sub-surface environment. Contaminants in solid, liquid and gaseous phases can be considered. AMBER gives the user the flexibility to define any number of compartments; any number of contaminants and associated decays; deterministic, probabilistic and

  9. The solubilities of significant organic compounds in HLW tank supernate solutions -- FY 1995 progress report

    Barney, G.S.

    1996-01-01

    At the Hanford Site organic compounds were measured in tank supernate simulant solutions during FY 1995. This solubility information will be used to determine if these organic salts could exist in solid phases (saltcake or sludges) in the waste where they might react violently with the nitrate or nitrite salts present in the tanks. Solubilities of sodium glycolate, succinate, and caproate salts; iron and aluminum and butylphosphate salts; and aluminum oxalate were measured in simulated waste supernate solutions at 25 degree C, 30 degree C, 40 degree C, and 50 degree C. The organic compounds were selected because they are expected to exist in relatively high concentrations in the tanks. The solubilities of sodium glycolate, succinate, caproate, and butylphosphate in HLW tank supernate solutions were high over the temperature and sodium hydroxide concentration ranges expected in the tanks. High solubilities will prevent solid sodium salts of these organic acids from precipitating from tank supernate solutions. The total organic carbon concentrations (YOC) of actual tank supernates are generally much lower than the TOC ranges for simulated supernate solutions saturated (at the solubility limit) with the organic salts. This is so even if all the dissolved carbon in a given tank and supernate is due to only one of these eight soluble compounds (an unlikely situation). Metal ion complexes of and butylphosphate and oxalate in supernate solutions were not stable in the presence of the hydroxide concentrations expected in most tanks. Iron and aluminum dibutylphosphate compounds reacted with hydroxide to form soluble sodium dibutylphosphate and precipitated iron and aluminum hydroxides. Aluminum oxalate complexes were also not stable in the basic simulated supernate solutions. Solubilities of all the organic salts decrease with increasing sodium hydroxide concentration because of the common ion effect of Na+. Increasing temperatures raised the solubilities of the organic

  10. Hydro-mechanical behaviour of crushed COx argillite used as backfilling material in HLW repository

    Tang Chaosheng; Shi Bin; Cui Yujun; Anh-Minh Tang

    2010-01-01

    At present, the crushed Callovo-Oxfordian (COx) argillite powder is proposed as an alternative backfilling material in France, which will be constructed in the engineering barrier of high-level radioactive waste (HLW) repository. In this investigation, the compression behavior of two crushed COx argillite powders (coarser one and finer one) was studied by running l-D compression tests with several loading-unloading cycles. After the final dry density 2.0 g/cm 3 was reached, the specimen was flooding with distilled water and the evolution of axial stress was studied during saturation process. The effects of initial axial stress level and grain size distribution (GSD) on hydro-mechanical behaviour of compacted specimen were analyzed. The results show that the compression curves are significantly influenced by the GSD of the soils. To obtain the same degree of compaction, the axial stress applied to finer soil is much higher than that of coarser soil. In addition, the compression index of the finer soil is bigger than that of coarser soil. The swelling index at initial water content increases with the dry density and seems to be independent of the GSD. During saturation, the initial lower axial stress causes obvious swelling behavior for both the coarser and finer powder samples and the corresponding axial stress increase gradually. At initial higher axial stress condition, monotone collapse behavior is observed for the coarser powder samples. Whereas the axial stress decrease firstly, then increase and finally decrease again for the finer powder samples. After saturation, the equilibrium axial stresses of finer powder samples are higher than that of coarser powder samples. (authors)

  11. Dissolution of ORNL HLW sludge and partitioning of the actinides using the TRUEX process

    Spencer, B.B.; Egan, B.Z.; Beahm, E.C.; Chase, C.W.; Dillow, T.A.

    1997-01-01

    Experiments were conducted to evaluate the transuranium extraction (TRUEX) process for partitioning actinides from actual dissolved high-level radioactive waste (HLW) sludge. Samples of sludge from melton Valley Storage Tank W-25 were rinsed with mild caustic (0.2 M NaOH) to reduce the concentrations of nitrates and fission products associated with the interstitial liquid. In one campaign the rinsed sludge was leached in nitric acid, and about 50% of the dry mass of the sludge was dissolved. The resulting solution contained total metal concentrations of ∼ 1.8 M with a nitric acid concentration of 2.9 M. In the other campaign the sludge was neutralized with nitric acid to destroy the carbonates, then leached with 2.6 M NaOH for ∼ 6 h before rinsing with the mild caustic. The sludge was then leached in nitric acid, and about 80% of the sludge dissolved. The resulting solution contained total metal concentrations of ∼ 0.6 M with a nitric acid concentration of 1.7 M. Chemical analyses of both phases were used to evaluate the process. Evaluation was based on two metrics: the fraction of TRU elements removed from the dissolved sludge and comparison of the results with predictions made with the Generic TRUEX Model (GTM). The fractions of Eu, Pu, Cm, Th and U species removed from aqueous solution in only one extraction stage were > 95% and were close to the values predicted by the GTM. Mercury was also found to be strongly extracted, with a one-stage removal of > 92%. In one test, vanadium appeared to be moderately extracted

  12. Time-frames and the demonstration of safety for HLW disposal

    Watkins, B.; Kessler, J.

    1999-01-01

    An important principle which is often embodied in the criteria for the safe disposal of long-lived radioactive wastes is that a similar level of radiation protection should be provided to future generations as that provided for those alive today. This has resulted in the development of performance assessment methodologies to evaluate the potential long term impacts of HLW disposal on humans, usually in terms of individual dose or risk. However, the actual periods of time over which it is expected that there will be full control over high level waste disposals are extremely short in comparison with the times over which radionuclides in the wastes could potentially move from the deep repository and emerge into the surface environment. This leads to problems in setting quantitative dose or risk based standard appropriate for the short and long term, and in setting the time-frames for which calculations should be carried out. This is especially difficult in view of the uncertainty in predicting changes in human behaviour and changes in the biosphere and geosphere over the time-scales involved. Different assessment time-frames and approaches proposed by IAEA, Nordic countries, Britain and US guidance documents are briefly reviewed. Whilst accepting the basic radiation protection objective of protecting future generations, no international consensus bas been agreed on what time-frames should be used in performance assessments. It is recommended that different time-frames should be associated with different quantitative or qualitative performance measures. As a result, a range of indicators of safety may be appropriate in demonstrating compliance with regulatory performance criteria and the consequent overall assessment context. It is argued that what is required is a simple, robust yet defensible approach to time-frames and performance indicators which can be accepted by the public, regulators and the nuclear industry

  13. DESIGN OF A CONCRETE SLAB FOR STORAGE OF SNF AND HLW CASKS

    J. Bisset

    2005-01-01

    This calculation documents the design of the Spent Nuclear Fuel (SNF) and High-Level Waste (HLW) Cask storage slab for the Aging Area. The design is based on the weights of casks that may be stored on the slab, the weights of vehicles that may be used to move the casks, and the layout shown on the sketch for a 1000 Metric Ton of Heavy Metal (MTHM) storage pad on Attachment 2, Sht.1 of the calculation 170-C0C-C000-00100-000-00A (BSC 2004a). The analytical model used herein is based on the storage area for 8 vertical casks. To simplify the model, the storage area of the horizontal concrete modules and their related shield walls is not included. The heavy weights of the vertical storage casks and the tensile forces due to pullout at the anchorages will produce design moments and shear forces that will envelope those that would occur in the storage area of the horizontal modules. The design loadings will also include snow and live loads. In addition, the design will also reflect pertinent geotechnical data. This calculation will document the preliminary thickness and general reinforcing steel requirements for the slab. This calculation also documents the initial design of the cask anchorage. Other slab details are not developed in this calculation. They will be developed during the final design process. The calculation also does not include the evaluation of the effects of cask drop loads. These will be evaluated in this or another calculation when the exact cask geometry is known

  14. Remote Fiber Laser Cutting System for Dismantling Glass Melter - 13071

    Mitsui, Takashi; Miura, Noriaki [IHI Corporation, 1 Shin-Nakahara-cho, Isogo-ku, Yokohama, Kanagawa (Japan); Oowaki, Katsura; Kawaguchi, Isao [IHI Inspection and Instrumentation Co., Ltd, 1 Shin-Nakahara-cho, Isogo-ku, Yokohama, Kanagawa (Japan); Miura, Yasuhiko; Ino, Tooru [Japan Nuclear Fuel Limited, 4-108, Aza Okitsuke, Oaza Obuchi, Rokkasho-Mura, Kamikita-gun, Aomori (Japan)

    2013-07-01

    Since 2008, the equipment for dismantling the used glass melter has been developed in High-level Liquid Waste (HLW) Vitrification Facility in the Japanese Rokkasho Reprocessing Plant (RRP). Due to the high radioactivity of the glass melter, the equipment requires a fully-remote operation in the vitrification cell. The remote fiber laser cutting system was adopted as one of the major pieces of equipment. An output power of fiber laser is typically higher than other types of laser and so can provide high-cutting performance. The fiber laser can cut thick stainless steel and Inconel, which are parts of the glass melter such as casings, electrodes and nozzles. As a result, it can make the whole of the dismantling work efficiently done for a shorter period. Various conditions of the cutting test have been evaluated in the process of developing the remote fiber cutting system. In addition, the expected remote operations of the power manipulator with the laser torch have been fully verified and optimized using 3D simulations. (authors)

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

    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

  16. MIIT: International in-situ testing of nuclear-waste glasses: Performance of SRS simulated waste glass after five years of burial at the Waste Isolation Pilot Plant (WIPP)

    Wicks, G.G.; Lodding, A.R.; Macedo, P.B.; Clark, D.E.

    1991-01-01

    In July of 1986, the first in-situ test involving burial of simulated high-level waste (HLW) forms conducted in the United States was started. This program, called the Materials Interface Interactions Test or MIIT, comprises the largest, most cooperative field-testing venture in the international waste management community. In July of 1991, the experimental portion of the 5-year MIIT study was completed on schedule. During this time interval, many in-situ measurements were performed, thousands of brine analyses conducted, and hundreds of waste glass and package components exhumed and evaluated after 6 mo., 1 yr., 2 yr. and 5 yr. burial periods. Although analyses are still in progress, the performance of SRS waste glass based on all data currently available has been seen to be excellent thus far. Initial analyses and assessment of Savannah River (SR) waste glass after burial in WIPP at 90 degrees C for 5 years are presented in this document

  17. Setting up a safe deep repository for long-lived HLW and ILW in Russia: Current state of the works

    Polyakov, Yu.D.; Porsov, A.Yu.; Beigul, V.P.; Palenov, M.V.

    2014-01-01

    The concept of RW disposal in Russia in accordance with the Federal Law 'On Radioactive Waste Management and Amendments to Specific Legal Acts of the Russian Federation' No. 190-FL dated 11 July 2011, is oriented at the ultimate disposal of waste, without an intent for their subsequent retrieval. The law 190-FL has it as follows: - A radioactive waste repository is a radioactive waste storage facility intended for disposal of the radioactive wastes without an intent for their subsequent retrieval. - Disposal of solid long-lived high-level waste and solid long-lived intermediate-level waste is carried out in deep repositories for radioactive waste. - Import into the Russian Federation of radioactive waste for the purpose of its storage, processing and disposal, except for spent sealed sources of ionising radiation originating from the Russian Federation, is prohibited. For safe final disposal of long-lived HLW and ILW, it is planned to construct a deep repository for radioactive waste (DRRW) in a low-pervious monolith rock massif in the Krasnoyarsk region in the production territory of the Mining and Chemical Combine (FSUE 'Gorno-khimicheskiy kombinat'). According to the IAEA recommendations and in line with the international experience in feasibility studies for setting up of HLW and SNF underground disposal facilities, the first mandatory step is the construction of an underground research laboratory. An underground laboratory serves the following purposes: - itemised research into the characteristics of enclosing rock mass, with verification of massive material suitability for safe disposal of long-lived HLW and ILW; - research into and verification of the isolating properties of an engineering barrier system; - development of engineering solutions and transportation and process flow schemes for construction and running of a future RW ultimate isolation facility. (authors)

  18. Suggestions on selection of clay site as a key alternative of underground repository for HLW geological disposal in China

    Zheng Hualing; Fu Bingjun; Fan Xianhua; Chen Shi; Sun Donghui

    2006-01-01

    Site selection for the underground repository is a vital problem with respect to the HLW geological disposal. Over the past decades, we have been focusing our attention on granite as a priority in China. However, there are some problems have to be discussed on this matter. In this paper, both experiences gained and lessons learned in the international community regarding the site selection are described. And then, after analyzing a lot of some key factors affecting the site selection, some comments and suggestions on selection of clay site as a key alternative before final decision making in China are presented. (authors)

  19. Natural analogue of redox front formation in near-field environment at post-closure phase of HLW geological disposal

    Yoshida, Hidekazu; Yamamoto, Koushi; Amano, Yuki

    2005-01-01

    Redox fronts are created in the near field of rocks, in a range of oxidation environments, by microbial activity in rock groundwater. Such fronts, and the associated oxide formation, are usually unavoidable around high level radioactive waste (HLW) repositories, whatever their design. The long term behaviour of these oxides after repositories have been closed is however little known. Here we introduce an analogue of redox front formation, such as 'iron oxide' deposits, known as takashikozo forming cylindrical nodules, and the long term behaviour of secondarily formed iron oxyhydroxide in subsequent geological environments. (author)

  20. Status of Progress Made Toward Safety Analysis and Technical Site Evaluations for DOE Managed HLW and SNF.

    Sevougian, S. David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Stein, Emily [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gross, Michael B [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hammond, Glenn Edward [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Frederick, Jennifer M [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mariner, Paul [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-11-01

    The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) is conducting research and development (R&D) on generic deep geologic disposal systems (i.e., repositories). This report describes specific activities in FY 2016 associated with the development of a Defense Waste Repository (DWR)a for the permanent disposal of a portion of the HLW and SNF derived from national defense and research and development (R&D) activities of the DOE.

  1. lead glass brick

    When you look through the glass at a picture behind, the picture appears raised up because light is slowed down in the dense glass. It is this density (4.06 gcm-3) that makes lead glass attractive to physicists. The refractive index of the glass is 1.708 at 400nm (violet light), meaning that light travels in the glass at about 58% its normal speed. At CERN, the OPAL detector uses some 12000 blocks of glass like this to measure particle energies.

  2. Survey of life-cycle costs of glass-paper HEPA filters

    Moore, P.; Bergman, W.; Gilbert, H.

    1992-08-01

    We have conducted a survey of the major users of glass-paper HEPA filters in the DOE complex to ascertain the life cycle costs of these filters. Purchase price of the filters is only a minor portion of the costs; the major expenditures are incurred during the removal and disposal of contaminated filters. Through personal interviews, site visits and completion of questionnaires, we have determined the costs associated with the use of HEPA filters in the DOE complex. The total approximate life-cycle cost for a standard (2 in. x 2 in. x 1 in.) glass-paper HEPA filter is $3,000 for one considered low-level waste (LLW), $11,780 for transuranic (TRU) and $15,000 for high-level waste (HLW). The weighted-average cost for a standard HEPA filter in the complex is $4,753

  3. Characterization of the microstructure of zirconolite-based glass-ceramics

    Loiseau, P.; Caurant, D.; Touet, I.; Destre, Y.; Fillet, C.

    2000-01-01

    December 1991 legislation in France has spurred research on enhanced separation and conditioning or transmutation of long-lived radionuclides from high level radioactive wastes (HLW). In this field, we have studied zirconolite-based glass-ceramics in which the crystalline phase (zirconolite: CaZrTi 2 O 7 ) aimed to preferentially incorporate minor actinides is embedded in a glassy calcium aluminosilicate matrix. At the laboratory scale, the crystallization of the parent glass is carried out thanks to a two-step thermal treatment: a nucleation stage followed by a growth stage. This paper presents the evolution of the crystallization, followed by scanning electron microscopy (SEM) and X-ray diffraction (XRD), with the temperature of the crystal growth thermal treatment, in the range 950 deg. C - 1350 deg. C. (authors)

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

    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

  5. Measurement of ac electrical conductivity of molten glass by impedance measurement using co-axial cylinder electrode

    Shah, J.G.; Yalmali, V.S.; Tawde, Manisha; Mishra, R.

    2006-01-01

    The need of nuclear power as an energy source requires the solution of many problems. One of the most important is fixation of high level radioactive waste (HLW) in suitable borosilicate glass formulation. The major issue with this process is maximum waste loading in the final vitrified product without compromising on long term product characteristics. The electrical resistivity measurement at high temperature could not be measured with good precision using standard parallel plate electrode configuration due to error in cell constant measurement. Hence a high accuracy, calibration free technique consisting of co-axial electrodes was employed

  6. Glass transition of anhydrous starch by fast scanning calorimetry.

    Monnier, Xavier; Maigret, Jean-Eudes; Lourdin, Denis; Saiter, Allisson

    2017-10-01

    By means of fast scanning calorimetry, the glass transition of anhydrous amorphous starch has been measured. With a scanning rate of 2000Ks -1 , thermal degradation of starch prior to the glass transition has been inhibited. To certify the glass transition measurement, structural relaxation of the glassy state has been investigated through physical aging as well as the concept of limiting fictive temperature. In both cases, characteristic enthalpy recovery peaks related to the structural relaxation of the glass have been observed. Thermal lag corrections based on the comparison of glass transition temperatures measured by means of differential and fast scanning calorimetry have been proposed. The complementary investigations give an anhydrous amorphous starch glass transition temperature of 312±7°C. This estimation correlates with previous extrapolation performed on hydrated starches. Copyright © 2017 Elsevier Ltd. All rights reserved.

  7. Glass and nuclear wastes

    Sombret, C.

    1982-10-01

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

  8. Microstructuring of glasses

    Hülsenberg, Dagmar; Bismarck, Alexander

    2008-01-01

    As microstructured glass becomes increasingly important for microsystems technology, the main application fields include micro-fluidic systems, micro-analysis systems, sensors, micro-actuators and implants. And, because glass has quite distinct properties from silicon, PMMA and metals, applications exist where only glass devices meet the requirements. The main advantages of glass derive from its amorphous nature, the precondition for its - theoretically - direction-independent geometric structurability. Microstructuring of Glasses deals with the amorphous state, various glass compositions and their properties, the interactions between glasses and the electromagnetic waves used to modify it. Also treated in detail are methods for influencing the geometrical microstructure of glasses by mechanical, chemical, thermal, optical, and electrical treatment, and the methods and equipment required to produce actual microdevices.

  9. Conceptual design for vitrification of HLW at West Valley using a rotary calciner/metallic melter

    Giraud, J.P.; Conord, J.P.; Saverot, P.M.

    1984-01-01

    The CEA has had an extensive research program in the field of vitrification technology for over 24 years, and several testing facilities were used throughout all phases of development and engineering: The Vulcain facility comprises a vitrification hot cell and four auxiliary hot cells. Vulcain allows the production of 2-kg samples of active glass. The off-gas treatment system allows testing the DF of each equipment. The auxiliary cells are equipped with leach-rate tests, diffusion tests, and irradiation tests on the glass samples. The Atlas facility is a reproduction of AVM calcination and vitrification furnaces at 1/2 scale enclosed in a glove box. This facility is used for testing ruthenium volatility and containment in the vitrification process. The full-scale AVM inactive pilot facility is used for testing calcination and vitrification of new compositions of high-level waste and for developing new types of vitrification furnaces. The inactive test loop is for testing air cooling of glass containers. The full-scale AVH inactive pilot facility is used for testing AVH technology and has been in operation since late 1981

  10. Modelling of radionuclide migration and heat transport from an High-Level-Radioactive-Waste-repository (HLW) in Boom clay

    Put, M.; Henrion, P.

    1992-01-01

    For the modelling of the migration of radionuclides in the Boom clay formation, the analytical code MICOF has been updated with a 3-dimensional analytical solution for discrete sources. the MICOF program is used for the calculation of the release of α and β emitters from the HIGH LEVEL RADIOACTIVE WASTES (HLW). A coherent conceptual model is developed which describes all the major physico-chemical phenomena influencing the migration of radionuclides in the Boom clay. The concept of the diffusion accessible porosity is introduced and included in the MICOF code. Different types of migration experiments are described with their advantages and disadvantages. The thermal impact of the HLW disposal in the stratified Boom clay formation has been evaluated by a finite element simulation of the coupled heat and mass transport equation. The results of the simulations show that under certain conditions thermal convection cells may form, but the convective heat transfer in the clay formation is negligible. 6 refs., 19 figs., 2 tabs., 5 appendices

  11. A methodology of uncertainty/sensitivity analysis for PA of HLW repository learned from 1996 WIPP performance assessment

    Lee, Y. M.; Kim, S. K.; Hwang, Y. S.; Kang, C. H.

    2002-01-01

    The WIPP (Waste Isolation Pilot Plant) is a mined repository constructed by the US DOE for the permanent disposal of transuranic (TRU) wastes generated by activities related to defence of the US since 1970. Its historical disposal operation began in March 1999 following receipt of a final permit from the State of NM after a positive certification decision for the WIPP was issued by the EPA in 1998, as the first licensed facility in the US for the deep geologic disposal of radioactive wastes. The CCA (Compliance Certification Application) for the WIPP that the DOE submitted to the EPA in 1966 was supported by an extensive Performance Assessment (PA) carried out by Sandia National Laboratories (SNL), with so-called 1996 PA. Even though such PA methodologies could be greatly different from the way we consider for HLW disposal in Korea largely due to quite different geologic formations in which repository are likely to be located, a review on lots of works done through the WIPP PA studies could be the most important lessons that we can learn from in view of current situation in Korea where an initial phase of conceptual studies on HLW disposal has been just started. The objective of this work is an overview of the methodology used in the recent WIPP PA to support the US DOE WIPP CCA ans a proposal for Korean case

  12. Measurement of optical glasses

    Nicolau-Rebigan, S.

    1978-11-01

    The possibilities of measurement of the optical glasses parameters needed in building optical devices especially in lasers devices are presented. In the first chapter the general features of the main optical glasses as well as the modalities of obtaining them are given. Chapter two defines the optical glass parameters, and the third chapter describes the measuring methods of the optical glass parameters. Finally, the conclusions which point out the utilization of this paper are presented. (author)

  13. Thermo-mechanical analysis for multi-level HLW repository concept

    Kwon, Sang Ki; Choi, Jong Won

    2004-01-01

    This work aims to investigate the influence of design parameters for the underground high-level nuclear waste repository with multi-level concept. B. Necessity o In order to construct an HLW repository in deep underground, it is required to select a site, which is far from major discontinuities. To dispose the whole spent fuels generated from the Korean nuclear power plants in a repository, the underground area of about 4km 2 is required. This would be a constraints for selecting an adequate repository site. It is recommended to dispose the two different spent fuels, PWR and CANDU, in different areas at the operation efficiency point of view. It is necessary to investigate the influence of parameters, which can affect the stability of multi-level repository. It is also needed to consider the influence of heat generated from the HLW and the high in situ stress in deep location. Therefore, thermo-mechanical coupling analysis should be carried out and the results should be compared with the results from single-level repository concept. Three-dimensional analysis is required to model the disposal tunnel and deposition hole. It is recommended to use the Korean geological condition and actually measured rock properties in Korea in order to achieve reliable modeling results. A FISH routine developed for effective modeling of Thermal-Mechanical coupling was implemented in the modeling using FLAC3D, which is a commercial three-dimensional FDM code. The thermal and mechanical properties of rock and rock mass achieved from Yusung drilling site, were used for the computer modeling. Different parameters such as level distance, waste type disposed on different levels, and time interval between the operation on different levels, were considered in the three-dimensional analysis. From the analysis, it was possible to derive adequate multi-level repository concept. Results and recommendations for application From the thermal-mechanical analysis for the multi-level repository

  14. Mechanically reinforced glass beams

    Nielsen, Jens Henrik; Olesen, John Forbes

    2007-01-01

    laminated float glass beam is constructed and tested in four-point bending. The beam consist of 4 layers of glass laminated together with a slack steel band glued onto the bottom face of the beam. The glass parts of the tested beams are \\SI{1700}{mm} long and \\SI{100}{mm} high, and the total width of one...

  15. Interfacial degradation of organic composite material by irradiation in reactor

    Nishijima, Shigehiro; Nishiura, Tetsuya; Okada, Toichi [Osaka Univ., Ibaraki (Japan). Inst. of Scientific and Industrial Research

    1996-04-01

    Glass fiber reinforced plastics (GFRP) with many kinds of matrix resins were made of E glass treated with silane as the reinforced material. Degradation of shearing strength of GFRP irradiated at low temperature was determined. It was clear from the results of comparing the degradation process with the fractured surface that the degradation was very affected by the radiation resistance of the bonded part between resin and coupling agents. It means that we had to be careful in the choice of interfacial treatments and epoxy matrices corresponded to it. (S.Y.)

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

    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)

  17. Immobilisation of radio cesium loaded ammonium molybdo phosphate in glass matrices

    Yalmali, Vrunda S.; Singh, I.J.; Sathi Sasidharan, N.; Deshingkar, D.S.

    2004-11-01

    Long half life and easy availability from high level wastes make 137 Cesium most economical radiation source. High level liquid waste processing for 137 Cesium removal has become easier due to development of Cesium specific granulated ammonium molybdophosphate (AMP) composite. In such applications, resulting spent composite AMP itself represents high active solid waste and immobilization of these materials in cement may not be acceptable. Studies on immobilization of 137 Cs loaded AMP were taken up in order to achieve twin goals of increasing safety and minimizing processing costs of the final matrix. Studies indicated that phosphate modified sodium borosilicate SPNM glasses prepared under usual oxidizing conditions are not suitable for immobilization of 137 Cs loaded on AMP .Phosphate glasses containing Na 2 O, P 2 O 5 , B 2 O 3 , Fe 2 O 3 , Al 2 O 3 and SiO 2 as major constituents are capable of incorporating 6 to 8 % AMP. The Normalized Leach rates of these glasses for sodium, cesium, boron and silica are 10 -4 to 10 -6 gm/cm 2 /day which are comparable to or better than those reported for NBS glasses incorporating HLW. Homogeneity of the final matrix was confirmed by x-ray diffraction analysis. Further studies on characterization of these glasses would establish their acceptability. (author)

  18. Fluoride glass fiber optics

    Aggarwal, Ishwar D

    1991-01-01

    Fluoride Glass Fiber Optics reviews the fundamental aspects of fluoride glasses. This book is divided into nine chapters. Chapter 1 discusses the wide range of fluoride glasses with an emphasis on fluorozirconate-based compositions. The structure of simple fluoride systems, such as BaF2 binary glass is elaborated in Chapter 2. The third chapter covers the intrinsic transparency of fluoride glasses from the UV to the IR, with particular emphasis on the multiphonon edge and electronic edge. The next three chapters are devoted to ultra-low loss optical fibers, reviewing methods for purifying and

  19. Multiple Glass Ceilings

    Russo, Giovanni; Hassink, Wolter

    2011-01-01

    Both vertical (between job levels) and horizontal (within job levels) mobility can be sources of wage growth. We find that the glass ceiling operates at both margins. The unexplained part of the wage gap grows across job levels (glass ceiling at the vertical margin) and across the deciles of the intra-job-level wage distribution (glass ceiling at the horizontal margin). This implies that women face many glass ceilings, one for each job level above the second, and that the glass ceiling is a p...

  20. Homogeneity of Inorganic Glasses

    Jensen, Martin; Zhang, L.; Keding, Ralf

    2011-01-01

    Homogeneity of glasses is a key factor determining their physical and chemical properties and overall quality. However, quantification of the homogeneity of a variety of glasses is still a challenge for glass scientists and technologists. Here, we show a simple approach by which the homogeneity...... of different glass products can be quantified and ranked. This approach is based on determination of both the optical intensity and dimension of the striations in glasses. These two characteristic values areobtained using the image processing method established recently. The logarithmic ratio between...

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

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

    2010-01-01

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

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

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

    2010-09-23

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

  3. Leaching of glass

    Hench, L.L.

    1977-01-01

    Understanding surface compositional profiles of glasses over a range of 0-2000 A with a variety of analytical instruments shows that five general types of glass surfaces exist. The surface character of a glass article depends upon bulk composition and environmental history during which surface dealkalization, film formation, and network dissolution can occur. Environmental-surface interactions generally result in complex compositional profiles of all the constituents in a glass. Durable glasses almost always develop a stable surface film which has a higher concentration of network formers than the bulk composition. Compositional effects that are used to improve glass durability usually improve the stability of the surface films. Durability tests or service conditions that lead to film destruction are especially severe for the most silicate glasses. 43 references

  4. An Assessment of Using Vibrational Compaction of Calcined HLW and LLW in DWPF Canisters

    Yi, Yun-Bo; Amme, Robert C.; Shayer, Zeev

    2008-01-01

    both of them) of applying the vibrational forces? 2) What is best mode of operation: first fill the canister with calcined waste and then vibrate it and refill it again, or apply vibrational forces during the filling process. By optimum or best we mean less creation of stress/strain forces during the volume reduction vibration process. Lessons learnt: This preliminary study shows that; 1) The maximum stress concentration always occurs in the canister wall, however its location varies and depends on the loading condition, and vibration process. 2) The proposed vibrational process would not cause any damages to the granulated calcined waste. 3) The first natural frequency of the longitudinal vibration of the canister is around 400 Hz, which is far away from the applied vibrational frequencies and from possibility of resonance phenomena that may cause damage to the canister 4) The relationship between the maximum internal stress and the frequency of the applied load is not parabolic. 5) The mechanical properties of the granulated calcined nuclear waste have small impact on the internal stress of the canister. Finally, the calculated data suggested that applying vibrational forces will keep the entire canister whole without any indication of development defects, and will have significant economical benefits of handling HLW and LLW in calcined forms, from waste manipulation, storage and transportation

  5. Benefits Of Vibration Analysis For Development Of Equipment In HLW Tanks - 12341

    Stefanko, D.; Herbert, J.

    2012-01-01

    Vibration analyses of equipment intended for use in the Savannah River Site (SRS) radioactive liquid waste storage tanks are performed during pre-deployment testing and has been demonstrated to be effective in reducing the life-cycle costs of the equipment. Benefits of using vibration analysis to identify rotating machinery problems prior to deployment in radioactive service will be presented in this paper. Problems encountered at SRS and actions to correct or lessen the severity of the problem are discussed. In short, multi-million dollar cost saving have been realized at SRS as a direct result of vibration analysis on existing equipment. Vibration analysis of equipment prior to installation can potentially reduce inservice failures, and increases reliability. High-level radioactive waste is currently stored in underground carbon steel waste tanks at the United States Department of Energy (DOE) Savannah River Site and at the Hanford Site, WA. Various types of rotating machinery (pumps and separations equipment) are used to manage and retrieve the tank contents. Installation, maintenance, and repair of these pumps and other equipment are expensive. In fact, costs to remove and replace a single pump can be as high as a half million dollars due to requirements for radioactive containment. Problems that lead to in-service maintenance and/or equipment replacement can quickly exceed the initial investment, increase radiological exposure, generate additional waste, and risk contamination of personnel and the work environment. Several different types of equipment are considered in this paper, but pumps provide an initial example for the use of vibration analysis. Long-shaft (45 foot long) and short-shaft (5-10 feet long) equipment arrangements are used for 25-350 horsepower slurry mixing and transfer pumps in the SRS HLW tanks. Each pump has a unique design, operating characteristics and associated costs, sometimes exceeding a million dollars. Vibration data are routinely

  6. Risk and uncertainty assessment for a potential HLW repository in Korea: TSPA 2006

    Hwang, Y.S.; Kang, C.H.

    2004-01-01

    KAERI has worked on the concept development on permanent disposal of HLW and its total system performance assessment since 1997. More than 36 000 MT of spent nuclear fuel from PWR and CANDU reactors is planned to be disposed of in crystalline bed-rocks. The total system performance assessment (TSPA) tools are under development. The KAERI FEP encyclopedia is actively developed to include all potential FEP suitable for Korean geo- and socio conditions. The FEPs are prioritized and then categorized to the intermediate level FEP groups. These groups become elements of the rock engineering system (RES) matrix. Then the sub-scenarios such as a container failure, groundwater migration, solute transport, etc are developed by connecting interactions between diagonal elements of the RES matrix. The full scenarios are developed from the combination of sub-scenarios. For each specific scenario, the assessment contexts and associated assessment method flow charts are developed. All information on these studies is recorded into the web based programme, FEAS (FEP to Assessment through Scenarios.) KAERI applies three basic programmes for the post closure radionuclide transport calculations; MASCOT-K, AMBER, and the new MDPSA under development. The MASCOT-K originally developed by Serco for a LLW repository has been extended extensively by KAERI to simulate release reactions such as congruent and gap releases in spent nuclear fuel. The new MDPSA code is dedicated for the probabilistic assessment of radio-nuclides in multi-dimensions of a fractured porous medium. To acquire input data for TSPA domestic experiment programmes as well as literature survey are performed. The data are stored in the Performance Assessment Input Data system (PAID.) To assure the transparency, traceability, retrievability, reproducibility, and review (T2R3) the web based KAERI QA system is developed. All tasks in TSPA are recorded under the concept of a 'Project' in this web system. Currently, FEAS, PAID

  7. BENEFITS OF VIBRATION ANALYSIS FOR DEVELOPMENT OF EQUIPMENT IN HLW TANKS - 12341

    Stefanko, D.; Herbert, J.

    2012-01-10

    Vibration analyses of equipment intended for use in the Savannah River Site (SRS) radioactive liquid waste storage tanks are performed during pre-deployment testing and has been demonstrated to be effective in reducing the life-cycle costs of the equipment. Benefits of using vibration analysis to identify rotating machinery problems prior to deployment in radioactive service will be presented in this paper. Problems encountered at SRS and actions to correct or lessen the severity of the problem are discussed. In short, multi-million dollar cost saving have been realized at SRS as a direct result of vibration analysis on existing equipment. Vibration analysis of equipment prior to installation can potentially reduce inservice failures, and increases reliability. High-level radioactive waste is currently stored in underground carbon steel waste tanks at the United States Department of Energy (DOE) Savannah River Site and at the Hanford Site, WA. Various types of rotating machinery (pumps and separations equipment) are used to manage and retrieve the tank contents. Installation, maintenance, and repair of these pumps and other equipment are expensive. In fact, costs to remove and replace a single pump can be as high as a half million dollars due to requirements for radioactive containment. Problems that lead to in-service maintenance and/or equipment replacement can quickly exceed the initial investment, increase radiological exposure, generate additional waste, and risk contamination of personnel and the work environment. Several different types of equipment are considered in this paper, but pumps provide an initial example for the use of vibration analysis. Long-shaft (45 foot long) and short-shaft (5-10 feet long) equipment arrangements are used for 25-350 horsepower slurry mixing and transfer pumps in the SRS HLW tanks. Each pump has a unique design, operating characteristics and associated costs, sometimes exceeding a million dollars. Vibration data are routinely

  8. Vitrification of HLW produced by uranium/molybdenum fuel reprocessing in cogema's cold crucible melter

    Quang, R. Do; Petitjean, V.; Hollebeque, F.; Pinet, O.; Flament, T.; Prodhomme, A.; Dalcorso, J. P.

    2003-01-01

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R and D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12% in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed

  9. Vitrification of HLW Produced by Uranium/Molybdenum Fuel Reprocessing in COGEMA's Cold Crucible Melter

    Do Quang, R.; Petitjean, V.; Hollebecque, F.; Pinet, O.; Flament, T.; Prod'homme, A.

    2003-01-01

    The performance of the vitrification process currently used in the La Hague commercial reprocessing plants has been continuously improved during more than ten years of operation. In parallel COGEMA (industrial Operator), the French Atomic Energy Commission (CEA) and SGN (respectively COGEMA's R and D provider and Engineering) have developed the cold crucible melter vitrification technology to obtain greater operating flexibility, increased plant availability and further reduction of secondary waste generated during operations. The cold crucible is a compact water-cooled melter in which the radioactive waste and the glass additives are melted by direct high frequency induction. The cooling of the melter produces a solidified glass layer that protects the melter's inner wall from corrosion. Because the heat is transferred directly to the melt, high operating temperatures can be achieved with no impact on the melter itself. COGEMA plans to implement the cold crucible technology to vitrify high level liquid waste from reprocessed spent U-Mo-Sn-Al fuel (used in gas cooled reactor). The cold crucible was selected for the vitrification of this particularly hard-to-process waste stream because it could not be reasonably processed in the standard hot induction melters currently used at the La Hague vitrification facilities : the waste has a high molybdenum content which makes it very corrosive and also requires a special high temperature glass formulation to obtain sufficiently high waste loading factors (12 % in molybdenum). A special glass formulation has been developed by the CEA and has been qualified through lab and pilot testing to meet standard waste acceptance criteria for final disposal of the U-Mo waste. The process and the associated technologies have been also being qualified on a full-scale prototype at the CEA pilot facility in Marcoule. Engineering study has been integrated in parallel in order to take into account that the Cold Crucible should be installed

  10. New compositions of fluoroindate glasses with higher chemical resistance

    B. J. Costa

    1998-06-01

    Full Text Available In this paper, glasses in the systems In-Ba-Mg and In-Ba-Zn-Sr-Mg were water leachead at 80ºC showing surface degradation after 72 hours of leaching. The extent of such degradation is determined by the solubility and the concentration of the elemental fluorides that constitute the glasses. The formation of a layer of crystallized phases on the surface of the samples was observed. Small weight losses were registered and the absence of water on the glass matrix after the attack suggested that the use of MgF2 in the systems studied can lead to better results against moisture corrosion when compared to other fluoride glasses such as the fluorozirconates.

  11. Execution techniques and approach for high level radioactive waste disposal in Japan: Demonstration of geological disposal techniques and implementation approach of HLW project

    Kawanishi, M.; Komada, H.; Kitayama, K.; Akasaka, H.; Tsuchi, H.

    2001-01-01

    In Japan, the high-level radioactive waste (HLW) disposal project is expected to start fully after establishment of the implementing organization, which is planned around the year 2000 and to dispose the wastes in the 2030s to at latest in the middle of 2040s. Considering each step in the implementation of the HLW disposal project in Japan, this paper discusses the execution procedure for HLW disposal project, such as the selection of candidate/planned disposal sites, the construction and operation of the disposal facility, the closure and decommissioning of facilities, and the institutional control and monitoring after the closure of disposal facility, from a technical viewpoint for the rational execution of the project. Furthermore, we investigate and propose some ideas for the concept of the design of geological disposal facility, the validation and demonstration of the reliability on the disposal techniques and performance assessment methods at a candidate/planned site. Based on these investigation results, we made clear a milestone for the execution of the HLW disposal project in Japan. (author)

  12. Poled-glass devices: Influence of surfaces and interfaces

    Fage-Pedersen, Jacob; Jacobsen, Rune Shim; Kristensen, Martin

    2007-01-01

    Devices in periodically poled glass must have a large periodic variation of the built-in field. We show that the periodic variation can be severely degraded by charge dynamics taking place at the external (glass–air) interface or at internal (glass–glass) interfaces if the interfaces have...... the device, one can reveal the existence of imperfect interfaces by use of electric field induced second-harmonic generation....

  13. Structural analysis and thermal behavior of diopside-fluorapatite-wollastonite-based glasses and glass-ceramics.

    Kansal, Ishu; Tulyaganov, Dilshat U; Goel, Ashutosh; Pascual, Maria J; Ferreira, José M F

    2010-11-01

    Glass-ceramics in the diopside (CaMgSi2O6)-fluorapatite (Ca5(PO4)3F)-wollastonite (CaSiO3) system are potential candidates for restorative dental and bone implant materials. The present study describes the influence of varying SiO2/CaO and CaF2/P2O5 molar ratio on the structure and thermal behavior of glass compositions in the CaO-MgO-SiO2-P2O5-Na2O-CaF2 system. The structural features and properties of the glasses were investigated by nuclear magnetic resonance (NMR), infrared spectroscopy, density measurements and dilatometry. Sintering and crystallization behavior of the glass powders were studied by hot-stage microscopy and differential thermal analysis, respectively. The microstructure and crystalline phase assemblage in the sintered glass powder compacts were studied under non-isothermal heating conditions at 825 °C. X-ray diffraction studies combined with the Rietveld-reference intensity ratio (R.I.R) method were employed to quantify the amount of amorphous and crystalline phases in the glass-ceramics, while scanning electron microscopy was used to shed some light on the microstructure of resultant glass-ceramics. An increase in CaO/SiO2 ratio degraded the sinterability of the glass powder compacts, resulting in the formation of akermanite as the major crystalline phase. On the other hand, an increase in P2O5/CaF2 ratio improved the sintering behavior of the glass-ceramics, while varying the amount of crystalline phases, i.e. diopside, fluorapatite and wollastonite. Copyright © 2010. Published by Elsevier Ltd.

  14. Mineral surface processes responsible for the decreased retardation (or enhanced mobilization) of 137Cs from HLW tank discharges. 1998 annual progress report

    Bertsch, P.M.; Zachara, J.M.

    1998-01-01

    'Cesium (137) is a major component of high level weapons waste. At Hanford, single shell tanks (SST''s) with high level wastes (HLW) have leaked supernate containing over 10 6 Ci of 137 Cs and other co-contaminants into the vadose zone. In select locations, 137 Cs has migrated further than expected from retardation experiments and performance assessment calculations. Deep 137 Cs migration has been observed beneath the SX tank farm at Hanford with REDOX wastes as the carrier causing regulatory and stakeholder concern. The causes for expedited migration are unclear. This research is investigating how the sorption chemistry of Cs on Hanford vadose zone sediments changes after contact with solutions characteristic of HLW. The central scientific hypothesis is that the high Na concentration of HLW will suppress surface-exchange reactions of Cs, except those to highly-selective frayed edge sites (FES) of the micaceous fraction. The authors further speculate that the concentrations, ion selectivity, and structural aspects of the FES will change after contact with HLW and that these changes will be manifest in the macroscopic sorption behavior of Cs. The authors believe that migration predictions of Cs can be improved substantially if such changes are understood and quantified. The research has three objectives: (1.) identify how the multi-component surface exchange behavior of Cs on Hanford sediments changes after contact with HLW simulants that span a range of relevant chemical (Na, OH, Al, K) and temperature conditions (23-80 C); (2) reconcile changes in sorption chemistry with microscopic and molecular changes in site distribution, chemistry, mineralogy, and surface structure of the micaceous fraction; (3) integrate mass-action-solution exchange measurements with changes in the structure/site distribution of the micaceous fraction to yield a multicomponent exchange model relevant to high ionic strength and hydroxide for prediction of environmental Cs sorption.'

  15. Thermo-hydro-mechanical processes in the nearfield around a HLW repository in argillaceous formations. Vol. I. Laboratory investigations

    Zhang, Chun-Liang; Czaikowski, Oliver; Rothfuchs, Tilmann; Wieczorek, Klaus

    2013-06-15

    All over the world, clay formations are being investigated as host medium for geologic disposal of radioactive waste because of their favourable properties, such as very low hydraulic conductivity against fluid transport, good sorption capacity for retardation of radionuclides, and high potential of self-sealing of fractures. The construction of a repository, the disposal of heat-emitting high-level radioactive waste (HLW), the backfilling and sealing of the remaining voids, however, will inevitably induce mechanical (M), hydraulic (H), thermal (T) and chemical (C) disturbances to the host formation and the engineered barrier system (EBS) over very long periods of time during the operation and post-closure phases of the repository. The responses and resulting property changes of the clay host rock and engineered barriers are to be well understood, characterized, and predicted for assessing the long-term performance and safety of the repository.

  16. Study on a transportation and emplacement system of pre-assembled EBS module for HLW geological disposal

    Awano, Toshihiko; Kanno, Takeshi; Katsumata, Syunsuke; Kosuge, Kazuhiro

    2009-01-01

    HLW disposal is one of the largest issue to utilize Nuclear power safely. In the past study, the concept, which buffer materials and Overpacked waste were transported into underground respectively, have shown. The concept of pre-assembled engineered barrier has advantage to simplify the logistics and emplacement procedure, however there are difficulties to support heavy weight of pre-assembled package by equipment under the condition of little clearance between tunnel and package. In this study, Combination of air bearing and two degree-of-freedom wheels were suggested for transportation, and air jack was suggested for unloading and emplacement system. Also, whole system for transportation and emplacement procedure was designed, and Scale model test was examined to evaluate the feasibility of these concept and functions. (author)

  17. Alternative biosphere modeling for safety assessment of HLW disposal taking account of geosphere-biosphere interface of marine environment

    Kato, Tomoko; Ishiguro, Katsuhiko; Naito, Morimasa; Ikeda, Takao; Little, Richard

    2001-03-01

    In the safety assessment of a high-level radioactive waste (HLW) disposal system, it is required to estimate radiological impacts on future human beings arising from potential radionuclide releases from a deep repository into the surface environment. In order to estimated the impacts, a biosphere model is developed by reasonably assuming radionuclide migration processes in the surface environment and relevant human lifestyles. It is important to modify the present biosphere models or to develop alternative biosphere models applying the biosphere models according to quality and quantify of the information acquired through the siting process for constructing the repository. In this study, alternative biosphere models were developed taking geosphere-biosphere interface of marine environment into account. Moreover, the flux to dose conversion factors calculated by these alternative biosphere models was compared with those by the present basic biosphere models. (author)

  18. A comparison of three methods for determining the amount of nitric acid needed to treat HLW sludge at SRS

    Siegwald, S.F.; Ferrara, D.M.

    1994-01-01

    A comparison was made of three methods for determining the amount of nitric acid which will be needed to treat a sample of high-level waste (HLW) sludge from the Savannah River Site (SRS) Tank Farm. The treatment must ensure the resulting melter feed will have the necessary rheological and oxidation-reduction properties, reduce mercury and manganese in the sludge, and be performed in a fashion which does not produce a flammable gas mixture. The three methods examined where an empirical method based on pH measurements, a computational method based on known reactions of the species in the sludge and a titration based on neutralization of carbonate in the solution

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

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

    2016-08-15

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

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

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

    2015-01-01

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

  1. Sodium Is Not Essential for High Bioactivity of Glasses

    Chen, Xiaojing; Chen, Xiaohui; Brauer, Delia S.; Wilson, Rory M.; Law, Robert V.; Hill, Robert G.; Karpukhina, Natalia

    2017-01-01

    This study aims to demonstrate that excellent bioactivity of glass can be achieved without the presence of an alkali metal component in glass composition. In vitro bioactivity of two sodium-free glasses based on the quaternary system SiO2-P2O5-CaO-CaF2 with 0 and 4.5 mol% CaF2 content was investigated and compared with the sodium containing glasses with equivalent amount of CaF2. The formation of apatite after immersion in Tris buffer was followed by Fourier transform infrared spectroscopy (FTIR), X-ray diffraction (XRD), 31P and 19F solid state MAS-NMR. The dissolution study was completed by ion release measurements in Tris buffer. The results show that sodium free bioactive glasses formed apatite at 3 hours of immersion in Tris buffer, which is as fast as the corresponding sodium containing composition. This signifies that sodium is not an essential component in bioactive glasses and it is possible to make equally degradable bioactive glasses with or without sodium. The results presented here also emphasize the central role of the glass compositions design which is based on understanding of structural role of components and/or predicting the network connectivity of glasses. PMID:29271977

  2. Diopside-Fluorapatite-Wollastonite Based Bioactive Glasses and Glass-ceramics =

    Kansal, Ishu

    Bioactive glasses and glass-ceramics are a class of biomaterials which elicit special response on their surface when in contact with biological fluids, leading to strong bonding to living tissue. This particular trait along with good sintering ability and high mechanical strength make them ideal materials for scaffold fabrication. The work presented in this thesis is directed towards understanding the composition-structure-property relationships in potentially bioactive glasses designed in CaO-MgO-P2O5-SiO2-F system, in some cases with added Na2O. The main emphasis has been on unearthing the influence of glass composition on molecular structure, sintering ability and bioactivity of phosphosilicate glasses. The parent glass compositions have been designed in the primary crystallization field of the pseudo-ternary system of diopside (CaO•MgO•2SiO2) - fluorapatite (9CaO•3P2O5•CaF2) - wollastonite (CaO•SiO2), followed by studying the impact of compositional variations on the structure-property relationships and sintering ability of these glasses. All the glasses investigated in this work have been synthesized via melt-quenching route and have been characterized for their molecular structure, sintering ability, chemical degradation and bioactivity using wide array of experimental tools and techniques. It has been shown that in all investigated glass compositions the silicate network was mainly dominated by Q2 units while phosphate in all the glasses was found to be coordinated in orthophosphate environment. The glass compositions designed in alkali-free region of diopside - fluorapatite system demonstrated excellent sintering ability and good bioactivity in order to qualify them as potential materials for scaffold fabrication while alkali-rich bioactive glasses not only hinder the densification during sintering but also induce cytotoxicity in vitro, thus, are not ideal candidates for in vitro tissue engineering. One of our bioglass compositions with low sodium

  3. Development and characterization of basalt-glass ceramics for the immobilization of transuranic wastes

    Lokken, R.O.; Chick, L.A.; Thomas, L.E.

    1982-09-01

    Basalt-based waste forms were developed for the immobilization of transuranic (TRU) contaminated wastes. The specific waste studied is a 3:1 blend of process sludge and incinerator ash. Various amounts of TRU blended waste were melted with Pomona basalt powder. The vitreous products were subjected to a variety of heat treatment conditions to form glass ceramics. The total crystallinity of the glass ceramic, ranging from 20 to 45 wt %, was moderately dependent on composition and heat treatment conditions. Three parent glasses and four glass ceramics with varied composition and heat treatment were produced for detailed phase characterization and leaching. Both parent glasses and glass ceramics were mainly composed of a continuous, glassy matrix phase. This glass matrix entered into solution during leaching in both types of materials. The Fe-Ti rich dispersed glass phase was not significantly degraded by leaching. The glass ceramics, however, exhibited four to ten times less elemental releases during leaching than the parent glasses. The glass ceramic matrix probably contains higher Fe and Na and lower Ca and Mg relative to the parent glass matrix. The crystallization of augite in the glass ceramics is believed to contribute to the improved leach rates. Leach rates of the basalt glass ceramic are compared to those of other TRU nuclear waste forms containing 239 Pu

  4. Fractography of glass

    Tressler, Richard

    1994-01-01

    As the first major reference on glass fractography, contributors to this volume offer a comprehensive account of the fracture of glass as well as various fracture surface topography Contributors discuss optical fibers, glass containers, and flatglass fractography In addition, papers explore fracture origins; the growth of the original flaws of defects; and macroscopic fracture patterns from which fracture patterns evolve This volume is complete with photographs and schematics

  5. Diamond turning of glass

    Blackley, W.S.; Scattergood, R.O.

    1988-12-01

    A new research initiative will be undertaken to investigate the critical cutting depth concepts for single point diamond turning of brittle, amorphous materials. Inorganic glasses and a brittle, thermoset polymer (organic glass) are the principal candidate materials. Interrupted cutting tests similar to those done in earlier research are Ge and Si crystals will be made to obtain critical depth values as a function of machining parameters. The results will provide systematic data with which to assess machining performance on glasses and amorphous materials

  6. Glass to contain wastes

    Moncouyoux, M.; Jacquet-Francillon, M.

    1994-01-01

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

  7. Glass microspheres for brachytherapy

    Prado, Miguel O.; Prastalo, Simon; Blaumann, Herman; Longhino, Juan M.; Repetto Llamazares, A.H.V.

    2007-01-01

    We developed the capacity to produce glass microspheres containing in their structure one or more radioactive isotopes useful for brachytherapy. We studied the various facts related with their production: (Rare earth) alumino silicate glass making, glass characterization, microspheres production, nuclear activation through (n,γ) nuclear reactions, mechanical characterization before and after irradiation. Corrosion tests in simulated human plasma and mechanical properties characterization were done before and after irradiation. (author) [es

  8. Waste glass corrosion modeling: Comparison with experimental results

    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

  9. Silicate glasses. Chapter 1

    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

  10. Glass and vitrification

    Barton, J.L.; Vacher, R.; Moncouyoux, J.P.; Vernaz, E.

    1997-01-01

    Most glasses used as materials are oxides glasses that are produced by a quick quench of a liquid. Glasses are characterized by the absence of periodicity in the atomic arrangements, they do not have symmetries and do not present order over a long distance. This series of 4 short articles present: 1) the properties of glass and its industrial story, 2) the glass structure, 3) a forty years long story of glass as dies used to confine wastes and 4) the methodology used to study the behaviour of glass over very long periods of time. This methodology is based on 5 steps: 1) define and specify the material to study (the prediction of long term alteration of a material is nonsense unless you know well its initial properties), 2) identify all the alteration processes that are likely to happen, determine their kinetics and the influence of environmental parameters, 3) develop mathematical models in order to simulate long-term behaviour of glasses, 4) determine the release rates of the radionuclides confined in the glass, and 5) validate data and models, it is not possible to expect a complete validation of a model that will be extrapolated over tens of thousands of years, nevertheless some ways of validation can lead to a satisfactory level of confidence taking into account reasonable uncertainties. (A.C.)

  11. Experimental programme to demonstrate the viability of the supercontainer concept for HLW

    Van Humbeeck, Hughes; De Bock, Chris; Bastiaens, Wim; Van Cotthem, Alain

    2008-01-01

    The EIG EURIDICE (a joint venture between the Belgian Organisation for Radioactive Waste Management - ONDRAF/NIRAS - and the Belgian Nuclear Research Centre - SCKoCEN) is responsible for performing large-scale tests, technical demonstrations and experiments to assess the feasibility of a final disposal of vitrified radioactive waste in deep clay layers. This is part of the Belgian Research and Development programme managed by ONDRAF/NIRAS. The current Belgian reference design for vitrified HLW and spent fuel assemblies is the so-called Supercontainer design. The vitrified waste canisters or spent fuel assemblies are enclosed in a carbon steel overpack which has to prevent contact between water from the host formation and the waste during the thermal phase. In order to maintain favourable chemical conditions to avoid corrosion during this period (several hundred or even thousand of years), the overpack is surrounded by a high alkaline concrete buffer of about 70 cm thick. The buffer also provides permanent radiological shielding for the workers, simplifying handling and other operations. All the components of the Supercontainer are constructed in above ground installations, thus creating favourable QA/QC conditions. After the emplacement of the Supercontainers in the disposal galleries, the remaining space will be backfilled. Tests to demonstrate the viability and the construction feasibility of the supercontainer design have been initiated. The viability programme includes Tests to verify the feasibility to construct and emplace the components of the supercontainers, and tests to verify the feasibility to backfill the disposal galleries once the supercontainers are placed. Supercontainer construction: Tests in column to verify the construction feasibility (risk of cracking) of the buffer with two different types of concrete (a self-compacting concrete - SCC - and a rheoplastic concrete RPC) were performed in collaboration with the Belgian concrete factory Socea. A

  12. Operational safety and radiation protection considerations in designing an HLW repository in Germany

    Filbert, W.; Kreienmeyer, M.; Poehler, M.; Niehues, N.

    2008-01-01

    In Germany the reference concept for disposal of heat generating radioactive waste considers emplacing canisters with vitrified waste in deep vertical boreholes drilled from the drifts of a repository mine in salt at a depth of 870 m. Spent fuel is to be disposed of in self-shielding POLLUX casks in horizontal drifts. An optimized disposal concept anticipates emplacing unshielded canisters with vitrified HLW and canisters containing the fuel rods of 3 PWR or 9 BWR fuel assemblies in boreholes with a diameter of 60 cm and a depth of up to 300 m.. In all cases the void space between POLLUX cask and drifts and canisters and borehole wall will be backfilled with crushed salt. (1) Operational Safety: Based on a detailed description of all underground disposal operation steps, the possible impacts on the disposal operations were analysed and the need for further studies determined. The disposal operation steps comprise e.g. rail bound transport from the shaft to the emplacement drift and emplacement process itself. As possible impacts the following occurrences were considered: ventilation failure, power supply failure, rock mechanics impact including cross-section convergence, irregular floor uplift and rock fall, brine and natural gas intrusion, derailing of transport carts and finally internal fire. (2) Radiation Protection: According to the German Atomic Energy Act (AtG), the design, construction and operation of a nuclear site like a final repository has to be licensed by the responsible authority. The Radiological Protection Ordinance and further guidelines i.e. concerning the emission and immission of released radioactive nuclides or the risk analysis of possible failure, build the basis for the licensing procedures. To ensure adequate protection against undue radiation exposure the repository is divided into different radiological protection areas. Generally, the handling of shielded waste packages above und under ground (including all the pathway of transport and

  13. Characterization of glass and glass ceramic nuclear waste forms

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

    1979-01-01

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

  14. Chemical Composition Analysis of INEEL Phase 3 Glasses: Task Technical and QA Plan

    Peeler, D.

    2000-01-01

    For about four decades radioactive wastes have been collected and calcined from nuclear fuels reprocessing at the Idaho Chemical Processing Plant (ICPP). Over this time span, secondary radioactive waste from decontamination, laboratory activities and fuels storage activities have also been collected and stored as liquid. These liquid high-activity wastes (HAW) are collectively called Sodium Bearing Wastes (SBW). Currently about 5.7 million liters of these wastes are temporarily stored in stainless steel tanks at the Idaho National Engineering and Environmental Laboratory (INEEL). Vitrification is being considered as a treatment option for SBW. The resulting glass can be sent to either the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico, as remote handled transuranic waste (RH-TRU) or to the federal geologic repository for final disposal. In addition to the SBW, roughly 4,000 m3 of calcined high-level wastes (HLW) are currently being stored at INEEL in stainless steel bin sets. These calcined HLW may also be vitrified, either with or without a dissolution and separation process, and sent to the federal geologic repository for final disposal

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

    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

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

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

    2017-12-07

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

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

    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.

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

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

    2004-01-01

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

  19. Relaxations in spin glasses: Similarities and differences from ordinary glasses

    Ngai, K.L.; Rajagopal, A.K.; Huang, C.Y.

    1984-01-01

    Relaxation phenomena have become a major concern in the physics of spin glasses. There are certain resemblances of these relaxation properties to those of ordinary glasses. In this work, we compare the relaxation properties of spin glasses near the freezing temperature with those of glasses near the glass transition temperature. There are similarities between the two types of glasses. Moreover, the relaxation properties of many glasses and spin glasses are in conformity with two coupled ''universality'' relations predicted by a recent model of relaxations in condensed matter

  20. Evaluation of Structural Cellular Glass

    Adams, M. A.; Zwissler, J. G.

    1984-01-01

    Preliminary design information presented. First report discusses state of structural-cellular-glass programs as of June 1979. Second report gives further details of program to develop improved cellular glasses and to characterize properties of glasses and commercially available materials.

  1. Electric glass capturing markets

    Wikman, K.; Wikstroem, T.

    1996-11-01

    Electric glass has found its place on the construction market. In public buildings, electrically heatable windows are becoming the leading option for large glass walls. Studies on detached houses, both new and renovated, show that floor heating combined with electrically heatable windowpanes is the best choice with respect to resident`s comfort. (orig.)

  2. Radioresistance of inorganic glasses

    Vorob'ev, A.A.; Zavadovskaya, E.K.; Fedorov, B.V.; Starodubtsev, V.A.

    1977-01-01

    Regularities are considered in the variation of properties of glass due to irradiations. On the basis of previous theoretical statements and experimental investigations, it is inferred that the irradiation resistance of glasses of the same type, synthesis conditions, content of impurities and amount of imperfections, is a function of the ''element-oxygen'' bond energy. The irradiation resistance depends on the number and the nature of glass structure imperfections. The averaged level of bonding forces is indicative of the glass formation temperature; the imperfections in glasses are formed in structure elements whose amount predominates as compared to the others. Electric charges which accumulate on the crack surface tend to increase its size, thus lessening even further the electric strength of the dielectric. The greater the irradiation time, the greater the number of irradiation imperfections causing a drop in the electric strength of glass. When choosing a glass for service in a radiation field, it is necessary to select those of a highest temperature of glass formation and with a least amount of imperfections

  3. Nucleation in ZBLAN glasses

    de Leede, G.L.A.; Waal, de H.

    1989-01-01

    Nucleation rates were detd. in a ZrF4-BaF2-NaF-LaF3-AlF3 glass (ZBLAN) using an optical method. The results were compared with a similar glass having a slightly different compn. The difference in the nucleation rate is explained by classical nucleation theory using calcd. free-energy differences

  4. Photo catalytic degradation of m-cresol

    Chavarria C, N.; Jimenez B, J.; Garcia S, I.; Valenzuela, M.A.

    2002-01-01

    The degradation of m-cresol was studied, a persistent organic compound that is consider a pollutant of residual water. There for a photo catalysis system was used, which consists in a glass reactor where is placed an aqueous solution of m-cresol and a semiconductor is added, in this case, titanium oxide. The solutions were irradiated with ultraviolet light and the surplus m-cresol was measured by UV vis spectrometry. The results indicate that the m-cresol is degraded until a 40% after 5 hours of irradiation in such conditions. (Author)

  5. Mechanical relaxation in glasses

    Hiki, Y.

    2004-01-01

    The basic properties of glasses and the characteristics of mechanical relaxation in glasses were briefly reviewed, and then our studies concerned were presented. Experimental methods adopted were viscosity, internal friction, ultrasonic attenuation, and Brillouin scattering measurements. The specimens used were several kinds of inorganic, organic, and metallic glasses. The measurements were mainly carried out from the room temperature up to the glass transition temperature, and the relaxation time was determined as a function of temperature. The 'double relaxation' composed of two Arrhenius-type relaxations was observed in many materials. In both relaxations, the 'compensation effect' showing a correlation of the pre-exponential factor and the activation energy was observed. These results were explained by considering the 'complex relaxation' due to cooperative motions of atoms or group of atoms. Values of activation energy near the glass transition determined by the various experimental methods were compared with each other

  6. Polymorphism in glasses

    Landa, L.M.; Nikolaeva, I.N.

    1979-01-01

    To defect phase interfaces and spasmodic properties change, the inhomogeneity and the second radiation effects in quartz glass, metamict phase and intermediate states have been investigated. When irradiating with fast neutrons the transformation of quartz glass - metamict phase occurs completely. The transformation is completed at 2x10 20 part./cm 2 dose. Thermal treatment not only increases the number of inhomogeneities but also results in increasing quartz glass density. Annealing transforms the metamict phase into common quartz glass at 1400 K. The fact, that thermal treatment results in the complete transformation of metamict phase into quartz glass, and the inverse transformation occurs only partially, is quite regular, as the metamict phase has a lesser entropy and is a more ordered state. It is shown that different amorphous phases of a chemical composition have different structures and properties, that there are interfaces between them, and the transformation from one state to another in microvolumes is realized spasmodically and requires expenditure of energy

  7. Thermal Conductivity of Foam Glass

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

    Due to the increased focus on energy savings and waste recycling foam glass materials have gained increased attention. The production process of foam glass is a potential low-cost recycle option for challenging waste, e.g. CRT glass and industrial waste (fly ash and slags). Foam glass is used...... as thermal insulating material in building and chemical industry. The large volume of gas (porosity 90 – 95%) is the main reason of the low thermal conductivity of the foam glass. If gases with lower thermal conductivity compared to air are entrapped in the glass melt, the derived foam glass will contain...... only closed pores and its overall thermal conductivity will be much lower than that of the foam glass with open pores. In this work we have prepared foam glass using different types of recycled glasses and different kinds of foaming agents. This enabled the formation of foam glasses having gas cells...

  8. Grouping of HLW in partitioning for B/T (burning and/or transmutation) treatment with neutron reactors based on three criteria

    Kitamoto, Mulyanto; Kitamoto, Asashi

    1995-01-01

    A grouping concept of HLW in partitioning for B/T (burning and/or transmutation) treatment by fission reactor was developed in order to improve the disposal in waste management from the safety aspect. The selecting and grouping concept was proposed herein, such as Group MA1 (Np, Am, and unrecovered U and Pu), Group MA2 (Cm, and trace quantity of Cf, etc.), Group A (Tc and I), Group B (Cs and Sr) and Group R (the partitioned remains of HLW), judging from the three criteria for B/T treatment, based on (1) the concept of the potential risk estimated by the hazard index for long-term tendency based on ALI (2) the concept of the relative dose factor related to the adsorbed migration rate transferred through ground water, and (3) the concept of the decay acceleration factor, the burning and/or transmutation characteristics for recycle B/T treatment. (author)

  9. Application of Kissinger analysis to glass transition and study of ...

    Home; Journals; Bulletin of Materials Science; Volume 35; Issue 4. Application of Kissinger analysis to glass transition and study of thermal degradation kinetics of phenolic–acrylic IPNs ... Author Affiliations. S Goswami1 K Kiran1. Department of Polymer Engineering, Birla Institute of Technology, Ranchi 835 215, India ...

  10. Effect of different glasses in glass bonded zeolite

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

    1995-01-01

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

  11. Role of glass structure in defining the chemical dissolution behavior, bioactivity and antioxidant properties of zinc and strontium co-doped alkali-free phosphosilicate glasses.

    Kapoor, Saurabh; Goel, Ashutosh; Tilocca, Antonio; Dhuna, Vikram; Bhatia, Gaurav; Dhuna, Kshitija; Ferreira, José M F

    2014-07-01

    We investigated the structure-property relationships in a series of alkali-free phosphosilicate glass compositions co-doped with Zn(2+) and Sr(2+). The emphasis was laid on understanding the structural role of Sr(2+) and Zn(2+) co-doping on the chemical dissolution behavior of glasses and its impact on their in vitro bioactivity. The structure of glasses was studied using molecular dynamics simulations in combination with solid state nuclear magnetic resonance spectroscopy. The relevant structural properties are then linked to the observed degradation behavior, in vitro bioactivity, osteoblast proliferation and oxidative stress levels. The apatite-forming ability of glasses has been investigated by X-ray diffraction, infrared spectroscopy and scanning electron microscopy-energy-dispersive spectroscopy after immersion of glass powders/bulk in simulated body fluid (SBF) for time durations varying between 1h and 14 days, while their chemical degradation has been studied in Tris-HCl in accordance with ISO 10993-14. All the glasses exhibit hydroxyapatite formation on their surface within 1-3h of their immersion in SBF. The cellular responses were observed in vitro on bulk glass samples using human osteosarcoma MG63 cell line. The dose-dependent cytoprotective effect of glasses with respect to the concentration of zinc and strontium released from the glasses is also discussed. Copyright © 2014 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

  12. Oxynitride glasses: a review

    Garcia, A.R.; Clausell, C.; Barba, A.

    2016-07-01

    Oxynitride glasses are special types of silicates or silicoaluminates which have been the object of many studies over the last forty years. They can be prepared by means of various complex methods, leading to variable levels of nitrogen incorporation, though in all cases giving limited transparency in the visible range. More recently, a new family of oxynitride glasses incorporating fluorine has been investigated. This paper outlines the effect of composition, in particular nitrogen and fluorine content, on properties such as glass transition temperature, hardness, Young's modulus, compactness and molar volume. (Author)

  13. Orbital glass in HTSC

    Kusmartsev, F.V.

    1992-10-01

    The physical reasons why the orbital glass may exist in granular high-temperature superconductors and the existing experimental data appeared recently are discussed. The orbital glass is characterized by the coexistence of the orbital paramagnetic state with the superconducting state and occurs at small magnetic fields H c0 c1 . The transition in orbital glass arises at the critical field H c0 which is inversely proportional to the surface cross-area S of an average grain. In connection with theoretical predictions the possible experiments are proposed. (author). 10 refs

  14. Bioactive glasses and glass-ceramics

    de Aza, P. N.

    2007-04-01

    Full Text Available Since the late 1960´s, a great interest in the use of bioceramic materials for biomedical applications has been developed. In a previous paper, the authors reviewed crystalline bioceramic materials “sensus stricto”, it is to say, those ceramic materials, constituted for non-metallic inorganic compounds, crystallines and consolidates by thermal treatment of powders at high temperature. In the present review, the authors deal with those called bioactive glasses and glassceramics. Although all of them are also obtained by thermal treatment at high temperature, the first are amorphous and the second are obtained by devitrification of a glass, although the vitreous phase normally prevails on the crystalline phases. After an introduction to the concept of bioactive materials, a short historical review of the bioactive glasses development is made. Its preparation, reactivity in physiological media, mechanism of bonding to living tissues and mechanical strength of the bone-implant interface is also reported. Next, the concept of glass-ceramic and the way of its preparation are exposed. The composition, physicochemical properties and biological behaviour of the principal types of bioactive glasses and glass-ceramic materials: Bioglass®, Ceravital®, Cerabone®, Ilmaplant® and Bioverit® are also reviewed. Finally, a short review on the bioactive-glass coatings and bioactive-composites and most common uses of bioactive-glasses and glass-ceramics are carried out too.

    Desde finales de los años sesenta, se ha despertado un gran interés por el uso de los materiales biocerámicos para aplicaciones biomédicas. En un trabajo previo, los autores hicieron una revisión de los denominados materiales biocerámicos cristalinos en sentido estricto, es decir, de aquellos materiales, constituidos por compuestos inorgánicos no metálicos, cristalinos y consolidados mediante tratamientos térmicos a altas temperaturas. En el presente trabajo, los autores

  15. Progress of the research and development on the geological disposal technology of HLW with aid of the industry/university collaboration system and fixed term researcher system

    Yamada, Fumitaka; Sonobe, Hitoshi; Igarashi, Hiroshi

    2008-02-01

    In Japan Atomic Energy Agency (JAEA), various systems associated with the collaboration with industries and universities on the Nuclear Fuel Cycle and the Postdoctoral Fellow system, etc. are enacted. These systems have been operated considering the needs of JAEA's program, industry and academia, resultantly contributed, for example, to basic research and the project development. The activities under these collaboration systems contain personal exchanges, the publication of the accomplishments and utilization of those, in research and development concerning geological disposal technology of high-level radioactive waste (HLW). These activities have progressed in Power Reactor and Nuclear Fuel Development Corporation (PNC) and Japan Nuclear Cycle Development Institute (JNC), which are the successive predecessors of JAEA, through JAEA. The accomplishments from these systems have been not only published as papers in journals and individual technical reports but also integrated into the project reports, accordingly contributed to the advancement of the national program on the geological disposal of HLW. In this report, the progress of the research and development under these systems was investigated from the beginning of the operation of the systems. The contribution to the research and development on geological disposal technology of HLW was also studied. On the basis of these studies, the future utilization of the systems of the collaboration was also discussed from the view point of the management of research and development program. A CD-ROM is attached as an appendix. (J.P.N.)

  16. Structure, biodegradation behavior and cytotoxicity of alkali-containing alkaline-earth phosphosilicate glasses.

    Kansal, Ishu; Reddy, AlluAmarnath; Muñoz, Francisco; Choi, Seong-Jun; Kim, Hae-Won; Tulyaganov, Dilshat U; Ferreira, José M F

    2014-11-01

    We report on the effect of sodium on the structure, chemical degradation and bioactivity of glasses in the CaO-MgO-SiO2-P2O5-CaF2 system. The (29)Si and (31)P magic angle spinning-nuclear magnetic resonance spectroscopy of melt-quenched glasses with varying Na2O/MgO ratios exhibit a silicate glass network with the dominance of Q(2)(Si) units and phosphorus mainly forming orthophosphate species. Sodium incorporation in the glasses did not induce a significant structural change in the silicate network, while it did influence the phosphate environment due to its lower ionic field strength in comparison with that of magnesium. The apatite forming ability of glasses has been investigated by immersion of glass powders in simulated body fluid (SBF) for time durations varying between 1h and 7 days while their chemical degradation has been studied in Tris-HCl in accordance with ISO-10993-14. Increasing Na(+)/Mg(2+) ratio caused a decrease in the chemical durability of glasses and in the apatite forming ability especially during initial steps of interaction between glass and SBF solution. The cellular responses were observed in vitro on bulk glass samples using mouse-derived pre-osteoblastic MC3T3-E1 cell line. The preliminary study suggested that the increasing alkali-concentration in glasses led to cytotoxicity in the cell culture medium. Copyright © 2014 Elsevier B.V. All rights reserved.

  17. Fun with Singing Wine Glasses

    Boone, Christine; Galloway, Melodie; Ruiz, Michael J.

    2018-01-01

    A fun activity is presented using singing wine glasses for introductory physics students. Students tune a white wine glass and a red wine glass to as many semitones as possible by filling the glasses with the appropriate amounts of water. A smart phone app is used to measure the frequencies of equal-temperament tones. Then plots of frequency…

  18. Waste glass weathering

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

    1994-01-01

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

  19. Super ionic conductive glass

    Susman, S.; Volin, K.J.

    Described is an ionically conducting glass for use as a solid electrolyte in a power or secondary cell containing an alkali metal-containing anode and a cathode separated by an alkali metal ion conducting glass having an ionic transference number of unity and the general formula: A/sub 1 + x/D/sub 2-x/3/Si/sub x/P/sub 3 - x/O/sub 12 - 2x/3/, wherein A is a network modifier for the glass and is an alkali metal of the anode, D is an intermediate for the glass and is selected from the class consisting of Zr, Ti, Ge, Al, Sb, Be, and Zn and X is in the range of from 2.25 to 3.0. Of the alkali metals, Na and Li are preferred and of the intermediate, Zr, Ti and Ge are preferred.

  20. Phosphate glasses, containing nitrogen

    Lisitsyna, E.A.; Khalilev, V.D.; Koryavin, A.A.; Goncharova, L.N.

    1987-01-01

    Possibilities of nitrogen-containing glass synthesis by the introduction into the charge of ammonium salts, as well as aluminium nitride, are studied. Zinc alumoyttrium phosphate glass (mol. %) Zn(PO 3 ) 2 - 4O, Al(PO 3 ) 3 - 3O, Y(PO 3 ) 3 -3O is suggested as a matrix. It is shown that the effect of amide and imide groups on the properties of the glass is less noticeable than the effect of nitride groups. Direct introduction of nitride constituent was realized using AlN, but aluminium introduction was taken into account so that the oxide was subtracted. The attempt to introduce more than 2.5 mass % of nitrogen into initial matrix by aluminium nitride has failed due to repeated restoration of glass with amorphous phosphorus isolation