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Sample records for aluminum hlw glasses

  1. HIGH ALUMINUM HLW GLASSES FOR HANFORD'S WTP

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-01-04

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-12-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Kim, Dong-Sang; Vienna, John D.

    2001-01-01

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

  11. Development Of Glass Matrices For HLW Radioactive Wastes

    International Nuclear Information System (INIS)

    Jantzen, C.

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    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

  13. DEVELOPMENT OF GLASS MATRICES FOR HLW RADIOACTIVE WASTES

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C.

    2010-03-18

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-04-02

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

  15. SOURCE TERMS FOR HLW GLASS CANISTERS

    International Nuclear Information System (INIS)

    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)

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

    Science.gov (United States)

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

    2017-01-01

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

  17. Cooling and cracking of technical HLW glass products

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-15

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

  3. HLW immobilization in glass

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Rutledge, V.J.; Maio, V.

    2013-01-01

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

  5. Molecular glasses for nuclear waste encapsulation

    International Nuclear Information System (INIS)

    Ropp, R.C.

    1982-01-01

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

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

    International Nuclear Information System (INIS)

    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

  7. ALUMINUM AND CHROMIUM LEACHING WORKSHOP WHITEPAPER

    International Nuclear Information System (INIS)

    McCabe, D; Jeff Pike, J; Bill Wilmarth, B

    2007-01-01

    A workshop was held on January 23-24, 2007 to discuss the status of processes to leach constituents from High Level Waste (HLW) sludges at the Hanford and Savannah River Sites. The objective of the workshop was to examine the needs and requirements for the HLW flowsheet for each site, discuss the status of knowledge of the leaching processes, communicate the research plans, and identify opportunities for synergy to address knowledge gaps. The purpose of leaching of non-radioactive constituents from the sludge waste is to reduce the burden of material that must be vitrified in the HLW melter systems, resulting in reduced HLW glass waste volume, reduced disposal costs, shorter process schedules, and higher facility throughput rates. The leaching process is estimated to reduce the operating life cycle of SRS by seven years and decrease the number of HLW canisters to be disposed in the Repository by 1000 [Gillam et al., 2006]. Comparably at Hanford, the aluminum and chromium leaching processes are estimated to reduce the operating life cycle of the Waste Treatment Plant by 20 years and decrease the number of canisters to the Repository by 15,000-30,000 [Gilbert, 2007]. These leaching processes will save the Department of Energy (DOE) billions of dollars in clean up and disposal costs. The primary constituents targeted for removal by leaching are aluminum and chromium. It is desirable to have some aluminum in glass to improve its durability; however, too much aluminum can increase the sludge viscosity, glass viscosity, and reduce overall process throughput. Chromium leaching is necessary to prevent formation of crystalline compounds in the glass, but is only needed at Hanford because of differences in the sludge waste chemistry at the two sites. Improving glass formulations to increase tolerance of aluminum and chromium is another approach to decrease HLW glass volume. It is likely that an optimum condition can be found by both performing leaching and improving

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Veronica J Rutledge; Vince Maio

    2013-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  18. Chemical compatibility of HLW borosilicate glasses with actinides

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

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

    1982-04-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  5. Fatigue crack initiation in hybrid boron/glass/aluminum fiber metal laminates

    International Nuclear Information System (INIS)

    Chang, P.-Y.; Yeh, P.-C.; Yang, J.-M.

    2008-01-01

    The fatigue crack initiation behavior of a high modulus and hybrid boron/glass/aluminum fiber/metal laminate (FML) was investigated experimentally and analytically. Two types of hybrid boron/glass/aluminum FMLs were fabricated and studied, which consisted of aluminum alloy sheets as the metal layers and a mixture of boron fibers and glass fibers as the composite layers. For the first type, the boron fiber/prepreg and the glass fiber/prepreg were used separately in the composite layers, and for the second type, the boron fibers and the glass fibers were mingled together to form a hybrid boron/glass/prepreg composite layer. These hybrid FMLs were consolidated using an autoclave curing process. The incorporation of the boron fibers improved the Young's modulus of the composite layer in FMLs, which in turn, would improve the fatigue crack initiation life of the Al sheet. The experimental results clearly showed that the fatigue crack initiation lives for both types of hybrid boron/glass/aluminum FMLs were superior to the monolithic aluminum alloy under the same loading condition. An analytical approach was proposed to calculate the fatigue crack initiation lives of hybrid boron/glass/aluminum FMLs based on the classical laminate theory and the small-crack theory. A good correlation was obtained between the predictions and the experimental results

  6. Melter Throughput Enhancements for High-Iron HLW

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  9. Aluminum and Phthalates in Calcium Gluconate: Contribution From Glass and Plastic Packaging.

    Science.gov (United States)

    Yokel, Robert A; Unrine, Jason M

    2017-01-01

    Aluminum contamination of parenteral nutrition solutions has been documented for 3 decades. It can result in elevated blood, bone, and whole body aluminum levels associated with neurotoxicity, reduced bone mass and mineral content, and perhaps hepatotoxicity. The primary aluminum source among parenteral nutrition components is glass-packaged calcium gluconate, in which aluminum concentration in the past 3 decades has averaged approximately 4000 μg/L, compared with nutrition solutions; 2 packaged in glass (from France and the United States) and 1 in plastic (from Germany); in a recently released plastic-packaged solution (from the United States); and in the 2 glass containers. Phthalate concentration was determined in selected samples of each product and leachate of the plastic containers. The initial aluminum concentration was approximately 5000 μg/L in the 2 glass-packaged products and approximately 20 μg/L in the plastic-packaged product, and increased approximately 30%, 50%, and 100% in 2 years, respectively. The aluminum concentration in a recently released Calcium Gluconate Injection USP was approximately 320 μg/L. Phthalates were not detected in any calcium gluconate solutions or leachates. Plastic packaging greatly reduces the contribution of aluminum to parenteral nutrition solutions from calcium gluconate compared with the glass-packaged product.

  10. Yttrium aluminum garnet coating on glass substrate

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Camila M.A.; Freiria, Gabriela S.; Faria, Emerson H. de; Rocha, Lucas A.; Ciuffi, Katia J.; Nassar, Eduardo J., E-mail: eduardo.nassar@unifran.edu.br

    2016-02-15

    Thin luminescent films have seen great technological advances and are applicable in the production of a variety of materials such as sensors, solar cells, photovoltaic devices, optical magnetic readers, waveguides, lasers, and recorders. Systems that contain yttrium aluminum oxide are important hosts for lanthanide ions and serve as light emission devices. This work deals with the deposition of yttrium aluminum garnet (YAG) film doped with Eu{sup 3+} onto a glass substrate obtained by the sol–gel methodology. Spray pyrolysis furnished the yttrium aluminum oxide powder. Dip-coating at a withdrawal speed of 10 mm min{sup −1} and evaporation led to deposition of different numbers of layers of the YAG:Eu{sup 3+} film onto the glass substrate from a YAG:Eu{sup 3+} powder suspension containing ethanol, water, and tetraethylorthosilicate. Photoluminescence, X-ray diffraction, scanning electron microscopy, and transparency measurements aided film characterization. The emission spectra revealed that the number of layers influenced film properties. - Highlights: • The spray pyrolysis was used to obtain luminescent YAG:Eu{sup 3+}. • The matrix was deposited as transparent films. • The YAG:Eu{sup 3+} was deposited by sol–gel process onto glass substrate.

  11. Yttrium aluminum garnet coating on glass substrate

    International Nuclear Information System (INIS)

    Ferreira, Camila M.A.; Freiria, Gabriela S.; Faria, Emerson H. de; Rocha, Lucas A.; Ciuffi, Katia J.; Nassar, Eduardo J.

    2016-01-01

    Thin luminescent films have seen great technological advances and are applicable in the production of a variety of materials such as sensors, solar cells, photovoltaic devices, optical magnetic readers, waveguides, lasers, and recorders. Systems that contain yttrium aluminum oxide are important hosts for lanthanide ions and serve as light emission devices. This work deals with the deposition of yttrium aluminum garnet (YAG) film doped with Eu 3+ onto a glass substrate obtained by the sol–gel methodology. Spray pyrolysis furnished the yttrium aluminum oxide powder. Dip-coating at a withdrawal speed of 10 mm min −1 and evaporation led to deposition of different numbers of layers of the YAG:Eu 3+ film onto the glass substrate from a YAG:Eu 3+ powder suspension containing ethanol, water, and tetraethylorthosilicate. Photoluminescence, X-ray diffraction, scanning electron microscopy, and transparency measurements aided film characterization. The emission spectra revealed that the number of layers influenced film properties. - Highlights: • The spray pyrolysis was used to obtain luminescent YAG:Eu 3+ . • The matrix was deposited as transparent films. • The YAG:Eu 3+ was deposited by sol–gel process onto glass substrate.

  12. Experimental studies of glued Aluminum-glass joints

    Science.gov (United States)

    Ligaj, B.; Wirwicki, M.; Karolewska, K.; Jasińska, A.

    2018-04-01

    Glued steel-glass or aluminum-glass joints are to be found, among other things, in vehicles (cars, buses, trains, trams) as windscreen assembly pieces for the supporting structure. For the purposes of the experiments, samples were made in which the top beam was made of the AW-2017A aluminum alloy and the bottom beam was made of thermally reinforced soda-lime glass whereas the glued joints were made of one-component polyurethane glue Körapur 175. The tests were performed under four-point bending conditions at monotonic incremental bending moment values on the Instron 5965 durability machine. The experimental study of the durability of glued joints under four-point bending conditions with the monotonic incremental bending moment allows to determine the values of stresses, whose value is related to initiation of damage of the tested joint.

  13. Static and Dynamic Behavior of High Modulus Hybrid Boron/Glass/Aluminum Fiber Metal Laminates

    Science.gov (United States)

    Yeh, Po-Ching

    2011-12-01

    This dissertation presents the investigation of a newly developed hybrid fiber metal laminates (FMLs) which contains commingled boron fibers, glass fibers, and 2024-T3 aluminum sheets. Two types of hybrid boron/glass/aluminum FMLs are developed. The first, type I hybrid FMLs, contained a layer of boron fiber prepreg in between two layers of S2-glass fiber prepreg, sandwiched by two aluminum alloy 2024-T3 sheets. The second, type II hybrid FMLs, contained three layer of commingled hybrid boron/glass fiber prepreg layers, sandwiched by two aluminum alloy 2024-T3 sheets. The mechanical behavior and deformation characteristics including blunt notch strength, bearing strength and fatigue behavior of these two types of hybrid boron/glass/aluminum FMLs were investigated. Compared to traditional S2-glass fiber reinforced aluminum laminates (GLARE), the newly developed hybrid boron/glass/aluminum fiber metal laminates possess high modulus, high yielding stress, and good blunt notch properties. From the bearing test result, the hybrid boron/glass/aluminum fiber metal laminates showed outstanding bearing strength. The high fiber volume fraction of boron fibers in type II laminates lead to a higher bearing strength compared to both type I laminates and traditional GLARE. Both types of hybrid FMLs have improved fatigue crack initiation lives and excellent fatigue crack propagation resistance compared to traditional GLARE. The incorporation of the boron fibers improved the Young's modulus of the composite layer in FMLs, which in turn, improved the fatigue crack initiation life and crack propagation rates of the aluminum sheets. Moreover, a finite element model was established to predict and verify the properties of hybrid boron/glass/aluminum FMLs. The simulated results showed good agreement with the experimental results.

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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. COMSOL Multiphysics Model for HLW Canister Filling

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-18

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

    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.

  1. Formation and stability of aluminum-based metallic glasses in Al-Fe-Gd alloys

    International Nuclear Information System (INIS)

    He, Y.; Poon, S.J.; Shiflet, G.J.

    1988-01-01

    Metallic glasses, a class of amorphous alloys made by rapid solidification, have been studied quite extensively for almost thirty years. It has been recognized for a long time that metallic glasses are usually very strong and ductile, and exhibit high corrosion resistance relative to crystalline alloys with the same compositions. Recently, metallic glasses containing as much as 90 atomic percent aluminum have been discovered independently by two groups. This discovery has both scientific and technological implications. The formability of these new glasses have been found to be unusual. Studies of mechanical properties in these new metallic glasses show that many of them have tensile strengths over 800MPa, greatly exceeding the strongest commercial aluminum alloys. The high strengths of aluminum-rich metallic glasses can be of significant importance in obtaining high strength low density materials. Therefore, from both scientific and technological standpoints, it is important to understand the formation and thermal stability of these metallic glasses. Al-Fe-Gd alloys were chosen for a more detailed study since they exhibit high tensile strengths

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  5. Industrial scale-plant for HLW partitioning in Russia

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-17

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

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

    1998-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-12

    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.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sonavane, M S; Mishra, P.K., E-mail: maheshss@barc.gov.in [Nuclear Recycle Board, Bhabha Atomic Research Centre, Mumbai (India); Mandal, S; Barik, S; Roy Chowdhury, A; Sen, R [Central Glass and Ceramic Institute, Kolkata (India)

    2012-10-15

    High level nuclear liquid waste (HLW) is immobilized using borosilicate glass matrix. Presently joule heated ceramic melter is being employed for vitrification of HLW in India. Preformed nodules of base glass are fed to melter along with liquid waste in predetermined ratio. In order to reduce the cost incurred for production of glass nodules of base glass, an alternative option of using glass grains was evaluated for its preparation and its suitability for the melter operation. (author)

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

    International Nuclear Information System (INIS)

    Sonavane, M.S.; Mishra, P.K.; Mandal, S.; Barik, S.; Roy Chowdhury, A.; Sen, R.

    2012-01-01

    High level nuclear liquid waste (HLW) is immobilized using borosilicate glass matrix. Presently joule heated ceramic melter is being employed for vitrification of HLW in India. Preformed nodules of base glass are fed to melter along with liquid waste in predetermined ratio. In order to reduce the cost incurred for production of glass nodules of base glass, an alternative option of using glass grains was evaluated for its preparation and its suitability for the melter operation. (author)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-05

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-09-23

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

  3. Historical fuel reprocessing and HLW management in Idaho

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

  5. Spectral-luminescence properties of trivalent titanium in aluminum-sodium phosphate glass

    International Nuclear Information System (INIS)

    Sukhanov, S.B.; Batyaev, I.M.

    1992-01-01

    Since development of the first crystal laser, Al 2 O 3 crystals remain the most widely used in quantum electronics. In the present work, the aluminum-sodium phosphate glass, Al 2 O 3 -Na 2 O 3 -P 2 O 5 , was studied with different proportions of components. A luminescence medium is obtained based on phosphate glass doped by Ti 3+ ions with intense emission in the 700-900-nm spectral range. This glass is a promising lasing medium for tunable solid-state lasers. 12 refs., 2 figs

  6. Source term measurements on vitrified HLW

    International Nuclear Information System (INIS)

    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)

  7. 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

    Energy Technology Data Exchange (ETDEWEB)

    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

  8. 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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

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

    1996-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    Amoroso, J.

    2011-01-01

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

  18. An examination of impact damage in glass-phenolic and aluminum honeycomb core composite panels

    Science.gov (United States)

    Nettles, A. T.; Lance, D. G.; Hodge, A. J.

    1990-01-01

    An examination of low velocity impact damage to glass-phenolic and aluminum core honeycomb sandwich panels with carbon-epoxy facesheets is presented. An instrumented drop weight impact test apparatus was utilized to inflict damage at energy ranges between 0.7 and 4.2 joules. Specimens were checked for extent of damage by cross sectional examination. The effect of core damage was assessed by subjecting impact-damaged beams to four-point bend tests. Skin-only specimens (facings not bonded to honeycomb) were also tested for comparison purposes. Results show that core buckling is the first damage mode, followed by delaminations in the facings, matrix cracking, and finally fiber breakage. The aluminum honeycomb panels exhibited a larger core damage zone and more facing delaminations than the glass-phenolic core, but could withstand more shear stress when damaged than the glass-phenolic core specimens.

  19. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    2004-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. Mechanical properties of nuclear waste glasses

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    International Nuclear Information System (INIS)

    Weber, W.J.

    1997-01-01

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

  7. HLW Tank Space Management, Final Report

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

  10. Silicate glasses

    International Nuclear Information System (INIS)

    Lutze, W.

    1988-01-01

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

  11. Crash worthy capacity of a hybridized epoxy-glass fiber aluminum columnar tube using repeated axial resistive force

    International Nuclear Information System (INIS)

    Paruka, Perowansa; Siswanto, Waluyo Adi; Maleque, Md Abdul; Shah, Mohd Kamal Mohd

    2015-01-01

    A combination of aluminum columnar member with composite laminate to form a hybrid structure can be used as collapsible energy absorbers especially in automotive vehicular structures to protect occupants and cargo. A key advantage of aluminum member in composite is that it provides ductile and stable plastic collapse mechanisms with progressive deformation in a stable manner by increasing energy absorption during collision. This paper presents an experimental investigation on the influence of the number of hybrid epoxy glass layers in overwrap composite columnar tubes. Three columnar tube specimens were used and fabricated by hand lay-up method. Aluminum square hollow shape was combined with externally wrapped by using an isophthalic epoxy resin reinforced with glass fiber skin with an orientation angle of 0 .deg. /90 .deg. The aluminum columnar tube was used as reference material. Crushed hybrid-composite columnar tubes were prepared using one, two, and three layers to determine the crash worthy capacity. Quasi-static crush test was conducted using INSTRON machine with an axial loading. Results showed that crush force and the number of layers were related to the enhancement of energy absorption before the collapse of columnar tubes. The energy absorption properties of the crushed hybrid-composite columnar tubes improved significantly with the addition of layers in the overwrap. Microscopic analysis on the modes of epoxy-glass fiber laminate failure was conducted by using scanning electron microscopy.

  12. Crash worthy capacity of a hybridized epoxy-glass fiber aluminum columnar tube using repeated axial resistive force

    Energy Technology Data Exchange (ETDEWEB)

    Paruka, Perowansa [Jalan Politeknik, Kota Kinabalu (Malaysia); Siswanto, Waluyo Adi [Universiti Tun Hussein Onn Malaysia, Parit Raja (Malaysia); Maleque, Md Abdul [Universiti Islam Antarabangsa Malaysia, Kuala Lumpur (Malaysia); Shah, Mohd Kamal Mohd [Universiti Malaysia Sabah, Kota Kinabalu (Malaysia)

    2015-05-15

    A combination of aluminum columnar member with composite laminate to form a hybrid structure can be used as collapsible energy absorbers especially in automotive vehicular structures to protect occupants and cargo. A key advantage of aluminum member in composite is that it provides ductile and stable plastic collapse mechanisms with progressive deformation in a stable manner by increasing energy absorption during collision. This paper presents an experimental investigation on the influence of the number of hybrid epoxy glass layers in overwrap composite columnar tubes. Three columnar tube specimens were used and fabricated by hand lay-up method. Aluminum square hollow shape was combined with externally wrapped by using an isophthalic epoxy resin reinforced with glass fiber skin with an orientation angle of 0 .deg. /90 .deg. The aluminum columnar tube was used as reference material. Crushed hybrid-composite columnar tubes were prepared using one, two, and three layers to determine the crash worthy capacity. Quasi-static crush test was conducted using INSTRON machine with an axial loading. Results showed that crush force and the number of layers were related to the enhancement of energy absorption before the collapse of columnar tubes. The energy absorption properties of the crushed hybrid-composite columnar tubes improved significantly with the addition of layers in the overwrap. Microscopic analysis on the modes of epoxy-glass fiber laminate failure was conducted by using scanning electron microscopy.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  14. Expected behavior of HLW glass in storage

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    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

  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

    Energy Technology Data Exchange (ETDEWEB)

    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. 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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, W. L.; Chemical Engineering

    2006-09-30

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Jain, V.; Pan, Y.M.

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  5. Corrosion behaviour of the WAK-HLW glass

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-26

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-02-01

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

  9. 12 Flasktransport of vitrified High Level Waste (HLW)

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. 12 Flasktransport of vitrified High Level Waste (HLW)

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  13. The borosilicate glass for 'PAMELA'

    International Nuclear Information System (INIS)

    Schiewer, E.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

    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. Immobilization of hazardous and radioactive waste into glass structures

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1997-01-01

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

  16. Solidification of HLLW by glass-ceramic process

    International Nuclear Information System (INIS)

    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

  17. Effect of magnesium aluminum silicate glass on the thermal shock resistance of BN matrix composite ceramics

    NARCIS (Netherlands)

    Cai, Delong; Jia, Dechang; Yang, Zhihua; Zhu, Qishuai; Ocelik, Vaclav; Vainchtein, Ilia D.; De Hosson, Jeff Th M.; Zhou, Yu

    The effects of magnesium aluminum silicate (MAS) glass on the thermal shock resistance and the oxidation behavior of h-BN matrix composites were systematically investigated at temperature differences from 600 degrees C up to 1400 degrees C. The retained strength rate of the composites rose with the

  18. New roots to formation of nanostructures on glass surface through anodic oxidation of sputtered aluminum

    Directory of Open Access Journals (Sweden)

    Satoru Inoue, Song-Zhu Chu, Kenji Wada, Di Li and Hajime Haneda

    2003-01-01

    Full Text Available New processes for the preparation of nanostructure on glass surfaces have been developed through anodic oxidation of sputtered aluminum. Aluminum thin film sputtered on a tin doped indium oxide (ITO thin film on a glass surface was converted into alumina by anodic oxidation. The anodic alumina gave nanometer size pore array standing vertically on the glass surface. Kinds of acids used in the anodic oxidation changed the pore size drastically. The employment of phosphoric acid solution gave several tens nanometer size pores. Oxalic acid cases produced a few tens nanometer size pores and sulfuric acid solution provided a few nanometer size pores. The number of pores in a unit area could be changed with varying the applied voltage in the anodization and the pore sizes could be increased by phosphoric acid etching. The specimen consisting of a glass substrate with the alumina nanostructures on the surface could transmit UV and visible light. An etched specimen was dipped in a TiO2 sol solution, resulting in the impregnation of TiO2 sol into the pores of alumina layer. The TiO2 sol was heated at ~400 °C for 2 h, converting into anatase phase TiO2. The specimens possessing TiO2 film on the pore wall were transparent to the light in UV–Visible region. The electro deposition technique was applied to the introduction of Ni metal into pores, giving Ni nanorod array on the glass surface. The removal of the barrier layer alumina at the bottom of the pores was necessary to attain smooth electro deposition of Ni. The photo catalytic function of the specimens possessing TiO2 nanotube array was investigated in the decomposition of acetaldehyde gas under the irradiation of UV light, showing that the rate of the decomposition was quite large.

  19. 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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  20. 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

    International Nuclear Information System (INIS)

    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.

  1. Experimental Study on the Feasibility of Using Water Glass and Aluminum Sulfate to Treat Complications in High Liquid Limit Soil Subgrade

    Directory of Open Access Journals (Sweden)

    Wen-hui Zhang

    2015-01-01

    Full Text Available The feasibility of using water glass and aluminum sulfate to treat high liquid limit soil subgrade diseases is studied through laboratory experiments, and the following results were observed. After improving the high liquid limit clay with water glass and aluminum sulfate, the liquid limit decreases, the plastic limit increases, and the plasticity index decreases. Compared with untreated soil, the clay content of the improved soil decreases, while the silt and coarse contents increase. The absolute and relative expansion rates of the improved soil are both lower than those of the untreated soil. With the same number of dry and wet cycles, the decreased degrees of cohesion and internal friction angle of the improved soil are, respectively, one-half and one-third of those of the untreated soil. After three dry and wet cycles, the California bearing ratio (CBR of the untreated soil does not meet the requirements of specifications. However, after being cured for seven days and being subjected to three dry and wet cycles, the CBR of the improved soil, with 4% water glass solution and 0.4% aluminum sulfate, meets the requirements of specifications.

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

    International Nuclear Information System (INIS)

    Hara, Shigeo; Terai, Ryohei; Yamanaka, Hiroshi

    1983-01-01

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

  3. FLOWSHEET FOR ALUMINUM REMOVAL FROM SLUDGE BATCH 6

    International Nuclear Information System (INIS)

    Pike, J.; Gillam, J.

    2008-01-01

    Samples of Tank 12 sludge slurry show a substantially larger fraction of aluminum than originally identified in sludge batch planning. The Liquid Waste Organization (LWO) plans to formulate Sludge Batch 6 (SB6) with about one half of the sludge slurry in Tank 12 and one half of the sludge slurry in Tank 4. LWO identified aluminum dissolution as a method to mitigate the effect of having about 50% more solids in High Level Waste (HLW) sludge than previously planned. Previous aluminum dissolution performed in a HLW tank in 1982 was performed at approximately 85 C for 5 days and dissolved nearly 80% of the aluminum in the sludge slurry. In 2008, LWO successfully dissolved 64% of the aluminum at approximately 60 C in 46 days with minimal tank modifications and using only slurry pumps as a heat source. This report establishes the technical basis and flowsheet for performing an aluminum removal process in Tank 51 for SB6 that incorporates the lessons learned from previous aluminum dissolution evolutions. For SB6, aluminum dissolution process temperature will be held at a minimum of 65 C for at least 24 days, but as long as practical or until as much as 80% of the aluminum is dissolved. As planned, an aluminum removal process can reduce the aluminum in SB6 from about 84,500 kg to as little as 17,900 kg with a corresponding reduction of total insoluble solids in the batch from 246,000 kg to 131,000 kg. The extent of the reduction may be limited by the time available to maintain Tank 51 at dissolution temperature. The range of dissolution in four weeks based on the known variability in dissolution kinetics can range from 44 to more than 80%. At 44% of the aluminum dissolved, the mass reduction is approximately 1/2 of the mass noted above, i.e., 33,300 kg of aluminum instead of 66,600 kg. Planning to reach 80% of the aluminum dissolved should allow a maximum of 81 days for dissolution and reduce the allowance if test data shows faster kinetics. 47,800 kg of the dissolved

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    Magnin, M.

    2009-09-01

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

  6. Technetium Chemistry in HLW

    International Nuclear Information System (INIS)

    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

  7. Study of lanthanum aluminum silicate glasses for passive and active optical fibers

    Science.gov (United States)

    Schuster, K.; Litzkendorf, D.; Grimm, S.; Kobelke, J.; Schwuchow, A.; Ludwig, A.; Leich, M.; Jetschke, S.; Dellith, J.; Auguste, J.-L.; Leparmentier, S.; Humbert, G.; Werner, G.

    2013-03-01

    We report on SiO2-Al2O3-La2O3 glasses - with and without Yb2O3 - suitable for nonlinear and fiber laser applications. We also present successful supercontinuum generation and fiber laser operation around 1060 nm in step-index fibers. We have optimized the glass compositions in terms of thermal and optical requirements for both a high La2O3 (24 mol%) and Yb2O3(6 mol%) concentration. The aluminum concentration was adjusted to about 21 mol% Al2O3 to increase the solubility of lanthanum and ytterbium in the glass beyond possible MCVD based techniques. The glasses have been characterized by dilatometrical methods to find transition temperatures from 860 to 880°C and thermal expansion coefficients between 4.1 and 7.0 × 10-6 K-1. Structured step index fibers with a SiO2-Al2O3-La2O3 core and silica cladding have been realized showing a fiber loss minimum of about 500 dB/km at 1200 nm wavelength. The chromatic dispersion could be adjusted to shift the zero dispersion wavelength (ZDW) close to the pump wavelength of 1550 nm in a supercontinuum generation setup. First fiber laser experiments show an efficiency of about 41 % with a remarkably reduced photodarkening compared to MCVD based fibers.

  8. Safety assessment of HLW geological disposal system

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

    Cunnane, J.C.

    1994-03-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Kruger, A.A.; Jain, V.

    2009-01-01

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

  13. 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/

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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)

  15. ALUMINUM BOX BUNDLING PRESS

    Directory of Open Access Journals (Sweden)

    Iosif DUMITRESCU

    2015-05-01

    Full Text Available In municipal solid waste, aluminum is the main nonferrous metal, approximately 80- 85% of the total nonferrous metals. The income per ton gained from aluminum recuperation is 20 times higher than from glass, steel boxes or paper recuperation. The object of this paper is the design of a 300 kN press for aluminum box bundling.

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

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; VIENNA JD; KIM DS; JAIN V

    2009-05-27

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

  17. 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

    International Nuclear Information System (INIS)

    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

  18. Aluminum elution and precipitation in glass vials: effect of pH and buffer species.

    Science.gov (United States)

    Ogawa, Toru; Miyajima, Makoto; Wakiyama, Naoki; Terada, Katsuhide

    2015-02-01

    Inorganic extractables from glass vials may cause particle formation in the drug solution. In this study, the ability of eluting Al ion from borosilicate glass vials, and tendencies of precipitation containing Al were investigated using various pHs of phosphate, citrate, acetate and histidine buffer. Through heating, all of the buffers showed that Si and Al were eluted from glass vials in ratios almost the same as the composition of borosilicate glass, and the amounts of Al and Si from various buffer solutions at pH 7 were in the following order: citrate > phosphate > acetate > histidine. In addition, during storage after heating, the Al concentration at certain pHs of phosphate and acetate buffer solution decreased, suggesting the formation of particles containing Al. In citrate buffer, Al did not decrease in spite of the high elution amount. Considering that the solubility profile of aluminum oxide and the Al eluting profile of borosilicate glass were different, it is speculated that Al ion may be forced to leach into the buffer solution according to Si elution on the surface of glass vials. When Al ions were added to the buffer solutions, phosphate, acetate and histidine buffer showed a decrease of Al concentration during storage at a neutral range of pHs, indicating the formation of particles containing Al. In conclusion, it is suggested that phosphate buffer solution has higher possibility of forming particles containing Al than other buffer solutions.

  19. Sodium aluminum-iron phosphate glass-ceramics for immobilization of lanthanide oxide wastes from pyrochemical reprocessing of spent nuclear fuel

    Science.gov (United States)

    Stefanovsky, S. V.; Stefanovsky, O. I.; Kadyko, M. I.; Nikonov, B. S.

    2018-03-01

    Sodium aluminum (iron) phosphate glass ceramics containing of up to 20 wt.% rare earth (RE) oxides simulating pyroprocessing waste were produced by melting at 1250 °C followed by either quenching or slow cooling to room temperature. The iron-free glass-ceramics were composed of major glass and minor phosphotridymite and monazite. The iron-bearing glass-ceramics were composed of major glass and minor monazite and Na-Al-Fe orthophosphate at low waste loadings (5-10 wt.%) and major orthophosphate and minor monazite as well as interstitial glass at high waste loadings (15-20 wt.%). Slowly cooled samples contained higher amount of crystalline phases than quenched ones. Monazite is major phase for REs. Leach rates from the materials of major elements (Na, Al, Fe, P) are 10-5-10-7 g cm-2 d-1, RE elements - lower than 10-5 g cm-2 d-1.

  20. The effect of electron irradiation on the structure and iron speciation in sodium aluminum (iron) phosphate glasses

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovsky, S.V., E-mail: serge.stefanovsky@yandex.ru [Frumkin Institute of Physical Chemistry and Electrochemistry RAS (Russian Federation); Presniakov, I.A.; Sobolev, A.V.; Glazkova, I.S. [Lomonosov Moscow State University, Department of Chemistry (Russian Federation); Kadyko, M.I.; Stefanovsky, O.I. [Frumkin Institute of Physical Chemistry and Electrochemistry RAS (Russian Federation)

    2016-08-01

    The effect of 8 MeV electron irradiation on the structure of glasses in the series 40 Na{sub 2}O, (20-x) Al{sub 2}O{sub 3}, x Fe{sub 2}O{sub 3}, 40 P{sub 2}O{sub 5} (mol.%) and on the iron speciation in these samples was studied by FTIR and Mössbauer spectroscopic techniques. Irradiation up to a dose of 1.0 MGy has no appreciable effects on the character of the bonds within anionic motif of the glass network. Electron irradiation increases the fraction of aluminum in octahedral coordination. Iron in both unirradiated and irradiated glasses is present mainly as Fe(III) (60–75% of the total amount) in the glasses and partly as Fe(II) and the ratio of two forms remains constant up to a dose of 1.0 MGy.

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

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K.; Edwards, T.

    2012-05-08

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

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

    International Nuclear Information System (INIS)

    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

  3. DEFENSE HIGH LEVEL WASTE GLASS DEGRADATION

    International Nuclear Information System (INIS)

    Ebert, W.

    2001-01-01

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

  4. Improvement of database on glass dissolution

    International Nuclear Information System (INIS)

    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)

  5. Liquidus Temperature Data for DWPF Glass

    International Nuclear Information System (INIS)

    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

  6. Embedded adhesive connection for laminated glass plates

    DEFF Research Database (Denmark)

    Hansen, Jens Zangenberg; Poulsen, S.H.; Bagger, A.

    2012-01-01

    The structural behavior of a new connection design, the embedded adhesive connection, used for laminated glass plates is investigated. The connection consists of an aluminum plate encapsulated in-between two adjacent triple layered laminated glass plates. Fastening between glass and aluminum...... usage in a design situation. The embedded connection shows promising potential as a future fastening system for load-carrying laminated glass plates....

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    1995-03-01

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

  9. Flexural Behavior of RC Members Using Externally Bonded Aluminum-Glass Fiber Composite Beams

    Directory of Open Access Journals (Sweden)

    Ki-Nam Hong

    2014-03-01

    Full Text Available This study concerns improvement of flexural stiffness/strength of concrete members reinforced with externally bonded, aluminum-glass fiber composite (AGC beams. An experimental program, consisting of seven reinforced concrete slabs and seven reinforced concrete beams strengthened in flexure with AGC beams, was initiated under four-point bending in order to evaluate three parameters: the cross-sectional shape of the AGC beam, the glass fiber fabric array, and the installation of fasteners. The load-deflection response, strain distribution along the longitudinal axis of the beam, and associated failure modes of the tested specimens were recorded. It was observed that the AGC beam led to an increase of the initial cracking load, yielding load of the tension steels and peak load. On the other hand, the ductility of some specimens strengthened was reduced by more than 50%. The A-type AGC beam was more efficient in slab specimens than in beam specimens and the B-type was more suitable for beam specimens than for slabs.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  14. Colloid formation during waste glass corrosion

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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)

  19. Durable Glass For Thousands Of Years

    International Nuclear Information System (INIS)

    Jantzen, C.

    2009-01-01

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

  20. DURABLE GLASS FOR THOUSANDS OF YEARS

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C.

    2009-12-04

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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  3. Glass leaching performance

    International Nuclear Information System (INIS)

    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

  4. Friction behavior of glass and metals in contact with glass in various environments

    Science.gov (United States)

    Buckley, D. H.

    1973-01-01

    Sliding friction experiments have been conducted for heat-resistant glass and metals in contact with glass. These experiments were conducted in various environments including vacuum, moist air, dry air, octane, and stearic acid in hexadecane. Glass exhibited a higher friction force in moist air than it did in vacuum when in sliding contact with itself. The metals, aluminum, iron, and gold, all exhibited the same friction coefficient when sliding on glass in vacuum as glass sliding on glass. Gold-to-glass contacts were extremely sensitive to the environment despite the relative chemical inertness of gold.

  5. The solubilities of significant organic compounds in HLW tank supernate solutions -- FY 1995 progress report

    International Nuclear Information System (INIS)

    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

  6. Experimental Investigation of Thermal Properties in Glass Fiber Reinforced with Aluminium

    Science.gov (United States)

    Irudaya raja, S. Joseph; Vinod Kumar, T.; Sridhar, R.; Vivek, P.

    2017-03-01

    A test method of a Guarded heat flow meter are used to measure the thermal conductivity of glass fiber and filled with a aluminum powder epoxy composites using an instrument in accordance with ASTM. This experimental study reveals that the incorporation of aluminum and glass fiber reinforced results in enhancement of thermal conductivity of epoxy resin and thereby improves its heat transfer capability. Fiber metal laminates are good candidates for advanced automobile structural applications due to their high categorical mechanical and thermal properties. The most consequential factor in manufacturing of these laminates is the adhesive bonding between aluminum and FRP layers. Here several glass-fiber reinforced aluminum were laminates with different proportion of bonding adhesion were been manufactured. It was observed that the damage size is more preponderant in laminates with poor interfacial adhesion compared to that of laminates with vigorous adhesion between aluminum and glass layers numerically calculated ones and it is found that the values obtained for various composite models using experimental testing method.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

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

    International Nuclear Information System (INIS)

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

    2002-03-01

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

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

    International Nuclear Information System (INIS)

    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

  10. R and D on HLW Partitioning in Russia

    International Nuclear Information System (INIS)

    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)

  11. Study of the structural role of gallium and aluminum in 45S5 bioactive glasses by molecular dynamics simulations.

    Science.gov (United States)

    Malavasi, Gianluca; Pedone, Alfonso; Menziani, Maria Cristina

    2013-04-18

    The structural properties of phosphosilicate glasses based on the 45S5 Bioglass doped with gallium and aluminum (46.2 SiO2·24.3Na2O·26.9CaO·2.6P2O5·1.0X2O3, X = Ga or Al) are investigated by means of classical molecular dynamics simulations. Structural features of the two compositions are compared with those of the original 45S5 Bioglass in order to relate them to the different known bioactivities of these materials. Differences in the coordination environments of Ga and Al, network connectivity, and ion aggregation reveal a microscopic model of these glasses which supports the interpretation of the experimental data and provides new insight into the different biological behaviors of Ga- and Al-containing phosphosilicate glasses. Although Ga is found predominantly in a 4-fold coordination environment, small amounts of 5- and 6-fold coordinated atoms have been detected depending on the interatomic potential model employed. This suggests its possible intermediate role in phosphosilicate glasses. On the contrary, Al plays a network former role and leads to glasses with a more polymerized structure. Interestingly, the results show an increased propensity for aggregation of the Ca(2+) and PO4(3-) ions in the Al-containing phosphosilicate glasses with respect to the Ga-containing ones. This leads to insoluble calcium-phosphate-rich regions not detected in the bioactive glasses.

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

    International Nuclear Information System (INIS)

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

    1983-05-01

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

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

    International Nuclear Information System (INIS)

    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.'

  14. Optical properties of single doped Cr3+ and co-doped Cr3+-Nd3+ aluminum tantalum tellurite glasses

    International Nuclear Information System (INIS)

    Rodriguez-Mendoza, U.R.; Speghini, A.; Jaque, D.; Zambelli, M.; Bettinelli, M.

    2004-01-01

    The optical properties for single doped Cr 3+ and co-doped Cr 3+ -Nd 3+ aluminum tantalum tellurite glasses have been studied as a function of temperature. For the single doped glass, the existence of two bands in the emission spectra at low temperature indicates the presence of two different sites for the Cr 3+ ions, labelled as usual as low- and high-field sites. The broad band centred in the Near Infrared region, corresponds to low-field sites transition 4 T 2 → 4 A 2 , and the narrow band centred at approximately 715 nm to the high-field sites transition 2 E→ 4 A 2 . The emission intensity for both high- and low-field sites shows a strong decrease with increasing temperature, with the emission for the former sites vanishing at RT. In both cases the quenching observed with the increase of temperature can be ascribed to the presence of non-radiative relaxation mechanisms. Experimental observations for the co-doped glass show that both radiative and non-radiative energy transfer processes from Cr 3+ to Nd 3+ are present

  15. Fiber glass reinforced structural materials for aerospace application

    Science.gov (United States)

    Bartlett, D. H.

    1968-01-01

    Evaluation of fiber glass reinforced plastic materials concludes that fiber glass construction is lighter than aluminum alloy construction. Low thermal conductivity and strength makes the fiber glass material useful in cryogenic tank supports.

  16. Final Report - High Level Waste Vitrification System Improvements, VSL-07R1010-1, Rev 0, dated 04/16/07

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Gan, H.; Pegg, I. L.; Gong, W.; Champman, C. C.; Joseph, I.; Matlack, K. S.

    2013-11-13

    This report describes work conducted to support the development and testing of new glass formulations that extend beyond those that have been previously investigated for the Hanford Waste Treatment and Immobilization Plant (WTP). The principal objective was to investigate maximization of the incorporation of several waste components that are expected to limit waste loading and, consequently, high level waste (HLW) processing rates and canister count. The work was performed with four waste compositions specified by the Office of River Protection (ORP); these wastes contain high concentrations of bismuth, chromium, aluminum, and aluminum plus sodium. The tests were designed to identify glass formulations that maximize waste loading while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM100) tests designed to examine the effects of enhanced glass formulations, increased glass processing temperature, increased crystallinity, and feed solids content on waste processing rate and product quality.

  17. Metallic glass coating on metals plate by adjusted explosive welding technique

    International Nuclear Information System (INIS)

    Liu, W.D.; Liu, K.X.; Chen, Q.Y.; Wang, J.T.; Yan, H.H.; Li, X.J.

    2009-01-01

    Using an adjusted explosive welding technique, an aluminum plate has been coated by a Fe-based metallic glass foil in this work. Scanning electronic micrographs reveal a defect-free metallurgical bonding between the Fe-based metallic glass foil and the aluminum plate. Experimental evidence indicates that the Fe-based metallic glass foil almost retains its amorphous state and mechanical properties after the explosive welding process. Additionally, the detailed explosive welding process has been simulated by a self-developed hydro-code and the bonding mechanism has been investigated by numerical analysis. The successful welding between the Fe-based metallic glass foil and the aluminum plate provides a new way to obtain amorphous coating on general metal substrates.

  18. HLW vitrification in France industrial experience and glass quality

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

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

    International Nuclear Information System (INIS)

    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.)

  2. Leaching behavior of simulated high-level waste glass

    International Nuclear Information System (INIS)

    Kamizono, Hiroshi

    1987-03-01

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

  3. White upconversion luminescence in Tm3+/Ho3+/Yb3+ triply doped K+-Na+ ion-exchanged aluminum germanate glass channel waveguide

    Science.gov (United States)

    Liu, Xiao; Chen, Baojie; Pun, Edwin Yue Bun; Lin, Hai

    2013-01-01

    Rare-earth ions doped K+-Na+ ion-exchanged aluminum germanate (NMAG) glass channel waveguides have been designed and fabricated. Under 980 nm laser pumping, an intense upconversion white light transmission trace was observed in Tm3+/Ho3+/Yb3+ triply doped NMAG glass channel waveguide and a high-brightness light spot was achieved from the output end of the fiber connected to the waveguide channel. The fluorescent colors were diverse and located within or near the white region in CIE chromaticity diagram under various pumping powers. These admirable results indicate that Tm3+/Ho3+/Yb3+ triply doped NMAG channel waveguide is a promising light source for medical and high-precision processing illumination.

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

    International Nuclear Information System (INIS)

    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

  5. Active geothermal systems as natural analogs of HLW repositories

    International Nuclear Information System (INIS)

    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)

  6. Retention of Halogens in Waste Glass

    Energy Technology Data Exchange (ETDEWEB)

    Hrma, Pavel R.

    2010-05-01

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

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

    International Nuclear Information System (INIS)

    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).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-17

    The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms within the DOE complex. These wastes can contain relatively high concentrations of sulfate, which has low solubility in borosilicate glass. This is a significant issue for low-activity waste (LAW) glass and is projected to have a major impact on the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Sulfate solubility has also been a limiting factor for recent high level waste (HLW) sludge processed at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). The low solubility of sulfate in glass, along with melter and off-gas corrosion constraints, dictate that the waste be blended with lower sulfate concentration waste sources or washed to remove sulfate prior to vitrification. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerate mission completion.The objective of the current scope being pursued by SHU is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DWPF and WTP, allowing for enhanced waste loadings and waste throughput at these facilities. A series of targeted glass compositions was selected to resolve data gaps in the model and is identified as Stage III. SHU fabricated these glasses and sent samples to SRNL for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for the Stage III, simulated HLW glasses fabricated by SHU in support of the sulfate solubility model development.

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

    International Nuclear Information System (INIS)

    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

  10. HLW Disposal System Development

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-14

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

  12. Magnetic resonance in spin glasses, superconductivity of thin aluminum films and models for transport properties of one dimensional systems

    International Nuclear Information System (INIS)

    Elliott, J.H.

    1983-01-01

    This thesis reports on three separate investigations in solid state physics. The first is electron paramagnetic resonance in the spin glass Ag:Mn. EPR measurements were performed at two resonance frequencies, concentrating on temperatures above the glass transition temperature. The measured linewidth appears to diverge at T/sub g/ for low resonance frequencies. These results will be compared with recently proposed phenomenological and microscopic theories. The second topic reported in this thesis is the superconducting transition of thin aluminum films. These films were investigated as a function of grain size and thickness. The transition temperature was enhanced over the bulk value, in agreement with many previous investigations of granular aluminum. The third topic reported in this thesis is an extension of the variable rate hopping theory applied in one dimension to N-ME-Qn(TCNQ) 2 . This model is a classical one used to explain both the dc and ac electrical conductivity of organic conductors. The temperature dependence of the model does not agree with experiment at low temperatures. Tunneling has been added to the hopping. This increases the conductivity at low temperatures, and results in excellent agreement with the experimental conductivity over the measured temperature range. The model also predicts that the frequency dependence of the conductivity varies as ω/sup .5/ at low frequencies. This long time tail prediction agrees with the measured dielectric constant of N-Me-iso-Qn(TCNQ) 2

  13. Final Report Melter Tests With AZ-101 HLW Simulant Using A Duramelter 100 Vitrification System VSL-01R10N0-1, Rev. 1, 2/25/02

    International Nuclear Information System (INIS)

    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

  14. Novel Synthesis of Calcium Oxide-Aluminum Oxide Glasses

    Science.gov (United States)

    Weber, J. K. Richard; Tangeman, Jean A.; Key, Thomas S.; Hiera, Kirsten J.; Paradis, Paul-Francois; Ishikawa, Takehiko; Yu, Jianding; Yoda, Shinichi

    2002-05-01

    Binary Al2O3:CaO glasses containing 36-50 mole% Al2O3 were synthesized by containerless processing of liquids in nitrogen using aerodynamic and a pressurized electrostatic-aerodynamic levitator. The critical cooling rate for glass formation RC under containerless conditions was ca. 70 K/s. The Vickers hardness of the glasses was 775-785; and the infrared transmission extended to approximately 5500 nm. The work function of the 36 mole% Al2O3 composition was 3.7 eV at 1100 K.

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

    International Nuclear Information System (INIS)

    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.'

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  18. DESIGN AND APPLICATION OF TRANSPARENT AND TRANSLUCENT ENAMELS ON ALUMINUM

    Directory of Open Access Journals (Sweden)

    H. AHMADI MOGHADDAM

    2012-09-01

    Full Text Available Transparent and opaque glass enamels for aluminum plates were designed with a minimum or with no heavy atom oxides such as lead and bismuth oxides. The thermal properties of the enamels were studied by DTA and their stability as measured by the difference of glass transition and crystallization onset temperatures was determined. Bending and rapid deformation (impact tests indicated the interfacial adhesion. The enamel/aluminum interfacial qualities were viewed and examined by scanning electron microscopy (SEM. A large amount of NaF and P2O5 in their formulation created opaque enamels. The three methods of melt dipping, pouring, and sintering were used to apply layers of enamels on aluminum plates. The novelty of the pouring and spreading method and its advantages over other methods, were in the use of lower stability and higher melting point enamels, without thermally/mechanically damaging the aluminum. Observations suggested that the interfacial contact and adhesion properties were good, particularly with the transparent or glassy state enamels.

  19. Study of viscosity of aluminum melt during glass transition by molecular dynamics and Green–Kubo formula

    International Nuclear Information System (INIS)

    Kirova, E M; Pisarev, V V

    2016-01-01

    Molecular dynamics study of shear viscosity behavior of liquid aluminum is performed. The embedded atom method potential is used at the simulation of isobaric cooling. The viscosity is calculated using the Green–Kubo formula. The stress autocorrelation functions are obtained in the range 300-1200 K. The calculated kinematic viscosity is in agreement with the experimental data for the temperatures above melting temperature. The steep change of the shear viscosity is found below 650 K which we associate with the glass transition and is in a good agreement with the temperature which is obtained using the calorimetric criterion Kolotova et al (2015 J. Non-Cryst. Solids 429 98). The viscosity coefficient can not be calculated using the direct atomistic simulations below that temperature. (paper)

  20. R and D programme for HLW disposal in Japan

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    1986-09-01

    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

  2. Counter current decantation washing of HLW sludge

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  8. Survey of glass plutonium contents and poison selection

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Accelerated Leach Testing of GLASS: ALTGLASS Version 3.0

    Energy Technology Data Exchange (ETDEWEB)

    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.

  10. Accelerated Leach Testing of GLASS: ALTGLASS Version 3.0

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    McKenzie, W.F.

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    McKenzie, W.F.

    1990-01-01

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

  13. Safety of HLW shipments

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  15. Characterization of the microstructure of zirconolite-based glass-ceramics

    International Nuclear Information System (INIS)

    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)

  16. FINAL REPORT MELTER TESTS WITH AZ-101 HLW SIMULANT USING A DURAMELTER 100 VITRIFICATION SYSTEM VSL-01R10N0-1 REV 1 2/25/02

    Energy Technology Data Exchange (ETDEWEB)

    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

  17. Corrosion of aluminum components and remedial measures

    International Nuclear Information System (INIS)

    Sheikh, S.T.; Khalique, A.; Malik, F.A.

    2006-01-01

    Aluminum has versatile physical properties, mechanical strength, corrosion resistance, and is used in special applications like aerospace, automobiles and other strategic industries. The outdoor exposed structural components of aluminum have very good corrosion resistance due to the thick oxide layer (0.2 -0.4 micro). This study involves the corrosion of aluminum based components, though aluminum is protected by an oxide layer but due to extreme weather and environmental conditions the oxide layer was damaged. The corroded product was removed, pits or cavities formed due to the material removal were filled with epoxy resins and acrylic-based compounds containing fibreglass as reinforcement. Optimum results were obtained with epoxy resins incorporated with 5% glass fibers. The inner surface of the components was provided further protection with a cellulose nitrate compound. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-01

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

  19. Hafnium in peralkaline and peraluminous boro-aluminosilicate glass, and glass subcomponents: a solubility study

    International Nuclear Information System (INIS)

    Davis, Linda L.; Darab, John G.; Qian, Maoxu; Zhao, Donggao; Palenik, Christopher S.; Li, Hong; Strachan, Denis M.; Li, Liyu

    2003-01-01

    A relationship between the solubility of hafnia (HfO2) and the host glass composition was explored by determining the solubility limits of HfO2 in peralkaline and peraluminous borosilicate glasses in the system SiO2-Al2O3-B2O3-Na2O, and in glasses in the system SiO2-Na2O-Al2O3 in air at 1450 C. The only Hf-bearing phase to crystallize in the peralkaline borosilicate melts is hafnia, while in the boron-free melts sodium-hafnium silicates crystallize. All peraluminous borosilicate melts crystallize hafnia, but the slightly peraluminous glasses also have sector-zoned hafnia crystals that contain Al and Si. The more peraluminous borosilicate glasses also crystallize a B-containing mullite. The general morphology of the hafnia crystals changes as peralkalinity (Na2O/(Na2O+Al2O3)) decreases, as expected in melts with increasing viscosity. In all of the glasses with Na2O > Al2O3, the solubility of hafnia is linearly and positively correlated with Na2O/(Na2O + Al2O3) or Na2O - Al2O3 (excess sodium), despite the presence of 5 to 16 mol% B2O3. The solubility of hafnia is higher in the sodium-aluminum borosilicate glasses than in the sodium-aluminosilicate glasses, suggesting that the boron is enhancing the effect that excess sodium has on the incorporation of Hf into the glass structure. The results of this solubility study are compared to other studies of high-valence cation solubility in B-free silicate melts. From this, for peralkaline B-bearing glasses, it is shown that, although the solubility limits are higher, the solution behavior of hafnia is the same as in B-free silicate melts previously studied. By comparison, also, it is shown that in peraluminous melts, there must be a different solution mechanism for hafnia: different than for peralkaline sodium-aluminum borosilicate glasses and different than for B-free silicate melts studied by others

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  1. PAIRWISE BLENDING OF HIGH LEVEL WASTE

    International Nuclear Information System (INIS)

    CERTA, P.J.

    2006-01-01

    The primary objective of this study is to demonstrate a mission scenario that uses pairwise and incidental blending of high level waste (HLW) to reduce the total mass of HLW glass. Secondary objectives include understanding how recent refinements to the tank waste inventory and solubility assumptions affect the mass of HLW glass and how logistical constraints may affect the efficacy of HLW blending

  2. Mixing up-conversion excitation behaviors in Er3+/Yb3+-codoped aluminum germanate glasses for visible waveguide devices

    International Nuclear Information System (INIS)

    Gong, H.; Lin, L.; Zhao, X.; Pun, E.Y.B.; Yang, D.L.; Lin, H.

    2010-01-01

    A mixing up-conversion excitation phenomenon in Er 3+ /Yb 3+ -codoped aluminum germanate (Na 2 O-MgO-Al 2 O 3 -GeO 2 , NMAG for short) glasses for K + -Na + ion-exchanged waveguides was observed and characterized. The green and red up-conversion luminescence of Er 3+ is due to a two-photon excitation process under low-power excitation of a 974 nm diode laser, however, with increasing the pumping power, the green emission turns to follow a combination of two- and three-photon excitation effects while the red one still agrees with a two-photon excitation law. Under high-power pumping, owing to potential thermal effect, the population ratio between the 2 H 11/2 and 4 S 3/2 levels adjusts acutely, which results in a distinct exhibition in 2 H 11/2 → 4 I 15/2 and 4 S 3/2 → 4 I 15/2 emission transitions. Green transmission trace has been observed in K + -Na + ion-exchanged Er 3+ /Yb 3+ -codoped NMAG glass waveguide and it provides an original reference in developing visible waveguide amplifiers and lasers.

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

    International Nuclear Information System (INIS)

    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)

  4. water alteration processes and kinetics of basaltic glasses, natural analogue of nuclear glasses

    International Nuclear Information System (INIS)

    Techer, I.; Advocat, Th.; Vernaz, E.; Lancelot, J.R.; Liotard, J.M.

    1997-01-01

    Dissolution experiments of a basaltic glass were carried out at 90 deg C for different reaction progresses. The initial dissolution rate was compared with values obtained for rhyolitic glass and the R7T7 nuclear glass. The activation energy was also determined by computing literature data. The results provide similar reactional mechanism for basaltic and nuclear glasses. Dissolution rates measured under saturation conditions were compared to theoretical dissolution rates. These ones were calculated using two kinetic models: the first rate equation is the Grambow's law which only takes into account ortho-silica acid activity; the second rate equation was proposed by Daux et al., where silica and aluminum are combined to formulate the affinity. The comparison between experimental and theoretical results point out that these two models are not appropriate to describe the alteration kinetic of basaltic glasses. (authors)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  7. Improvement of the mode II interface fracture toughness of glass fiber reinforced plastics/aluminum laminates through vapor grown carbon fiber interleaves.

    Science.gov (United States)

    Ning, Huiming; Li, Yuan; Hu, Ning; Cao, Yanping; Yan, Cheng; Azuma, Takesi; Peng, Xianghe; Wu, Liangke; Li, Jinhua; Li, Leilei

    2014-06-01

    The effects of acid treatment, vapor grown carbon fiber (VGCF) interlayer and the angle, i.e., 0° and 90°, between the rolling stripes of an aluminum (Al) plate and the fiber direction of glass fiber reinforced plastics (GFRP) on the mode II interlaminar mechanical properties of GFRP/Al laminates were investigated. The experimental results of an end notched flexure test demonstrate that the acid treatment and the proper addition of VGCF can effectively improve the critical load and mode II fracture toughness of GFRP/Al laminates. The specimens with acid treatment and 10 g m -2 VGCF addition possess the highest mode II fracture toughness, i.e., 269% and 385% increases in the 0° and 90° specimens, respectively compared to those corresponding pristine ones. Due to the induced anisotropy by the rolling stripes on the aluminum plate, the 90° specimens possess 15.3%-73.6% higher mode II fracture toughness compared to the 0° specimens. The improvement mechanisms were explored by the observation of crack propagation path and fracture surface with optical, laser scanning and scanning electron microscopies. Moreover, finite element analyses were carried out based on the cohesive zone model to verify the experimental fracture toughness and to predict the interface shear strength between the aluminum plates and GFRP laminates.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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)

  12. Determination of the corrosion mechanisms of high level waste containing glass

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  14. Optoelectronic properties of expanding thermal plasma deposited textured zinc oxide : effect of aluminum doping

    NARCIS (Netherlands)

    Groenen, R.; Kieft, E.R.; Linden, J.L.; Sanden, van de M.C.M.

    2006-01-01

    Aluminum-doped zinc oxide films exhibiting a rough surface morphol. are deposited on glass substrates utilizing expanding thermal plasma. Spectroscopic ellipsometry is used to evaluate optical and electronic film properties. The presence of aluminum donors in doped films is confirmed by a shift in

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

    Science.gov (United States)

    Christiansen, Mary U.

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

  16. Preparation of Aluminum Nanomesh Thin Films from an Anodic Aluminum Oxide Template as Transparent Conductive Electrodes

    Science.gov (United States)

    Li, Yiwen; Chen, Yulong; Qiu, Mingxia; Yu, Hongyu; Zhang, Xinhai; Sun, Xiao Wei; Chen, Rui

    2016-02-01

    We have employed anodic aluminum oxide as a template to prepare ultrathin, transparent, and conducting Al films with a unique nanomesh structure for transparent conductive electrodes. The anodic aluminum oxide template is obtained through direct anodization of a sputtered Al layer on a glass substrate, and subsequent wet etching creates the nanomesh metallic film. The optical and conductive properties are greatly influenced by experimental conditions. By tuning the anodizing time, transparent electrodes with appropriate optical transmittance and sheet resistance have been obtained. The results demonstrate that our proposed strategy can serve as a potential method to fabricate low-cost TCEs to replace conventional indium tin oxide materials.

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  18. Radiation hardening in sol-gel derived Er3+-doped silica glasses

    International Nuclear Information System (INIS)

    Hari Babu, B.; León Pichel, Mónica; Ollier, Nadège; El Hamzaoui, Hicham; Bigot, Laurent; Savelii, Inna; Bouazaoui, Mohamed; Poumellec, Bertrand; Lancry, Matthieu; Ibarra, Angel

    2015-01-01

    The aim of the present paper is to report the effect of radiation on the Er 3+ -doped sol-gel silica glasses. A possible application of these sol-gel glasses could be their use in harsh radiation environments. The sol-gel glasses are fabricated by densification of erbium salt-soaked nanoporous silica xerogels through polymeric sol-gel technique. The radiation-induced attenuation of Er 3+ -doped sol-gel silica is found to increase with erbium content. Electron paramagnetic resonance studies reveal the presence of E′ δ point defects. This happens in the sol-gel aluminum-silica glass after an exposure to γ-rays (kGy) and in sol-gel silica glass after an exposure to electrons (MGy). The concentration levels of these point defects are much lower in γ-ray irradiated sol-gel silica glasses. When the samples are co-doped with Al, the exposure to γ-ray radiation causes a possible reduction of the erbium valence from Er 3+ to Er 2+ ions. This process occurs in association with the formation of aluminum oxygen hole centers and different intrinsic point defects

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

    International Nuclear Information System (INIS)

    Kamizono, H.

    1990-01-01

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

  20. Viscosity of glasses containing simulated Savannah River Plant waste

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1978-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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)

  3. Characterization and Processing Behavior of Heated Aluminum-Polycarbonate Composite Build Plates for the FDM Additive Manufacturing Process

    Directory of Open Access Journals (Sweden)

    Sherri L. Messimer

    2018-02-01

    Full Text Available One of the most essential components of the fused deposition modeling (FDM additive manufacturing (AM process is the build plate, the surface upon which the part is constructed. These are typically made from aluminum or glass, but there are clear disadvantages to both and restrictions on which materials can be processed on them successfully. This study examined the suitability of heated aluminum-polycarbonate (AL-PC composite print beds for FDM, looking particularly at the mechanical properties, thermal behavior, deformation behavior, bonding strength with deposited material, printing quality, and range of material usability. Theoretical examination and physical experiments were performed for each of these areas; the results were compared to similar experiments done using heated aluminum and aluminum-glass print beds. Ten distinct materials (ABS, PLA, PET, HIPS, PC, TPU, PVA, nylon, metal PLA, and carbon-fiber PLA were tested for printing performance. The use of a heated AL-PC print bed was found to be a practical option for most of the materials, particularly ABS and TPU, which are often challenging to process using traditional print bed types. Generally, the results were found to be equivalent to or superior to tempered glass and superior to standard aluminum build plates in terms of printing capability.

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  5. Novel adaptations to zinc-silicate glass polyalkenoate cements: the unexpected influences of germanium based glasses on handling characteristics and mechanical properties.

    Science.gov (United States)

    Dickey, B T; Kehoe, S; Boyd, D

    2013-07-01

    Aluminum-free glass polyalkenoate cements (GPC) have been hindered for use as injectable bone cements by their inability to balance handling characteristics with mechanical integrity. Currently, zinc-based, aluminum-free GPCs demonstrate compression strengths in excess of 60MPa, but set in c. 1-2 min. Previous efforts to extend the setting reaction have remained clinically insufficient and are typically accompanied by a significant drop in strength. This work synthesized novel glasses based on a zinc silicate composition with the inclusion of GeO2, ZrO2, and Na2O, and evaluated the setting reaction and mechanical properties of the resultant GPCs. Germanium based GPCs were found to have working times between 5 and 10 min, setting times between 14 and 36 min, and compression strengths in excess of 30 MPa for the first 30 days. The results of this investigation have shown that the inclusion of GeO2, ZrO2, and Na2O into the glass network have produced, for the first time, an aluminum-free GPC that is clinically viable as injectable bone cements with regards to handling characteristics and mechanical properties. Crown Copyright © 2013. Published by Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    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)

  7. Solid oxide fuel cell having a glass composite seal

    Science.gov (United States)

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

    2013-04-16

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    International Nuclear Information System (INIS)

    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.)

  10. Effect of alumina on the dissolution rate of glasses

    International Nuclear Information System (INIS)

    Palavit, G.; Montagne, L.

    1997-01-01

    Small alumina addition to silicate glasses improves their chemical durability, but a large amount of alumina can also be beneficial to obtain a high dissolution rate. This paper describes the effect of Al 3+ on the early stage of glass alteration, in relation with its coordination in the glass and also with the reactions involved (hydrolysis and ionic exchange). We describe briefly nuclear magnetic resonance tools available to characterize the aluminum environments in the glasses. The rote of alumina on the dissolution rate of phosphate glasses is also discussed in order to show that the effect of Al 3+ is dependant upon the nature of the glass matrix. (author)

  11. Lithium-aluminum-zinc phosphate glasses activated with Tb3+ and Tb3+/Eu3+ for green laser medium, reddish-orange and white phosphor applications

    Science.gov (United States)

    Francisco-Rodriguez, H. I.; Lira, A.; Soriano-Romero, O.; Meza-Rocha, A. N.; Bordignon, S.; Speghini, A.; Lozada-Morales, R.; Caldiño, U.

    2018-05-01

    A spectroscopic analysis of Tb3+ and Tb3+/Eu3+ doped lithium-aluminum-zinc phosphate glasses is performed through their absorbance and photoluminescence spectra, and decay time profiles. Laser parameter values (stimulated emission cross section, effective bandwidth, gain bandwidth and optical gain) were obtained for the terbium 5D4 → 7F5 green emission from the Tb3+ singly-doped glass (LAZT) excited at 350 nm to judge the suitability of the glass phosphor for fiber lasers. A quantum yield of (47.68 ± 0.49)% was measured for the 5D4 level luminescence. Upon 350 nm excitation the LAZT glass phosphor emits green light with a color purity of 65.6% and chromaticity coordinates (0.285, 0.585) very close to those (0.29, 0.60) of European Broadcasting Union illuminant green. The Tb3+/Eu3+codoped glass emission color can be tuned from reddish-orange of 1865 K upon 318 nm excitation to warm white of 3599 K and neutral white of 4049 K upon 359 and 340 nm excitations, respectively. Upon Tb3+ excitation at 340 nm Eu3+ is sensitized by Tb3+ through a non-radiative energy transfer with an efficiency of 0.23-0.26. An electric dipole-dipole interaction might be the dominant mechanism in the Tb3+ to Eu3+ energy transfer taking place into Tb3+ - Eu3+ clusters.

  12. MIIT: International in-situ testing of nuclear-waste glasses: Performance of SRS simulated waste glass after five years of burial at the Waste Isolation Pilot Plant (WIPP)

    International Nuclear Information System (INIS)

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

    1991-01-01

    In July of 1986, the first in-situ test involving burial of simulated high-level waste (HLW) forms conducted in the United States was started. This program, called the Materials Interface Interactions Test or MIIT, comprises the largest, most cooperative field-testing venture in the international waste management community. In July of 1991, the experimental portion of the 5-year MIIT study was completed on schedule. During this time interval, many in-situ measurements were performed, thousands of brine analyses conducted, and hundreds of waste glass and package components exhumed and evaluated after 6 mo., 1 yr., 2 yr. and 5 yr. burial periods. Although analyses are still in progress, the performance of SRS waste glass based on all data currently available has been seen to be excellent thus far. Initial analyses and assessment of Savannah River (SR) waste glass after burial in WIPP at 90 degrees C for 5 years are presented in this document

  13. Fabrication and characterization of MCC approved testing material - ATM-1 glass

    International Nuclear Information System (INIS)

    Wald, J.W.

    1985-10-01

    The Materials Characterization Center Approved Testing Material ATM-1 is a borosilicate glass that incorporates nonradioactive constituents and uranium to represent 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 to which depleted uranium has been added as UO 2 . Three separate lots of ATM-1 glass have been fabricated, designated ATM-1a, ATM-1b, and ATM-1c. Limited analyses and microstructural evaluations were conducted on each type. Each lot of ATM-1 glass was produced from a feedstock melted in an air atmosphere at between 1150 to 1200 0 C and cast into stress annealed rectangular bars. Bars of ATM-1a were nominally 1.3 x 1.3 x 7.6 cm (approx.36 g each), bars of ATM-1b were nominally 2 x 2.5 x 17.5 cm (approx.190 g each) and bars of ATM-1c were nominally 1.9 x 1.9 x 15 cm (approx.170 g each). Thirteen bars of ATM-1a, 14 bars of ATM-1b, and 6 bars of ATM-1c were produced. Twelve random samples from each of lots ATM-1a, ATM-1b, and ATM-1c were analyzed. The concentrations (except for U and Cs) were obtained by Inductively-Coupled Argon Plasma Atomic Emission Spectroscopy analysis. Cesium analysis was performed by Atomic Absorption Spectroscopy, while uranium was analyzed by Pulsed Laser Fluorometry. X-ray diffraction analysis of four samples indicated that lot ATM-1a had no detectable crystalline phases (<3 wt %), while ATM-1b and ATM-1c contained approx.3 to 5 wt % iron-chrome spinel crystals. These concentrations of secondary spinel component are not considered uncommon. Scanning electron microscopy examination of fracture surfaces revealed only a random, apparently crystalline, second phase (1-10 μm diam) and a random distribution of small voids or bubbles (approx.1 μm nominal diam)

  14. Alkaline hydrothermal stabilization of Cr(VI) in soil using glass and aluminum from recycled municipal solid wastes.

    Science.gov (United States)

    Gattullo, Concetta Eliana; D'Alessandro, Caterina; Allegretta, Ignazio; Porfido, Carlo; Spagnuolo, Matteo; Terzano, Roberto

    2018-02-15

    Hexavalent chromium was stabilized in soil by using a mixture of glass and aluminum recovered from municipal solid wastes under alkaline hydrothermal conditions. Cr(VI) concentration was reduced by 94-98% already after 7days of treatment. After the same period, more than 90% of total Cr was stabilized in highly recalcitrant and scarcely mobile chemical forms, with 50% in the residual fraction (when the samples were treated at 1/10w/w mixture/soil ratio). Longer treatments increased Cr stabilization. X-ray microanalyses revealed that Cr was stabilized in geopolymeric structures within large aluminosilicate mineral aggregates (containing both amorphous and crystalline phases). 3D microstructural analyses showed a limited compaction of the soil with still a 20% internal porosity in the neoformed aggregates. Increased pH and salinity after the treatment can be restored by simple soil amendments and washing. Copyright © 2017. Published by Elsevier B.V.

  15. Survey of life-cycle costs of glass-paper HEPA filters

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  17. The properties and structure of Sn---Ca---P---O---F glasses

    International Nuclear Information System (INIS)

    Ding, J.Y.; Shih, P.Y.; Yung, S.W.; Hsu, K.L.; Chin, T.S.

    2003-01-01

    Low melting modified calcium stannous fluorophosphate glasses based on a basic composition 40P 2 O 5 -25SnO-30SnF 2 -5CaF 2 (in mol%) with glass transition temperature T g ranging 220-240 deg. C have been prepared. The effect of Al(OH) 3 or SiO 2 addition on the properties of these glasses has also been determined. Aluminum and silicon compounds decrease the dissolution rate in water and thermal expansion coefficient while increase the density, T g and softening temperature. The effect of Al(OH) 3 addition on the increase of chemical durability is better for these glasses, due to the partial crystallization effect in stannous calcium silicofluorophosphate glasses. With an addition of 4 wt.% Al(OH) 3 , the dissolution rate of the glass in 30 deg. C water decreases from 1.0x10 -5 to 1.3x10 -7 g cm -2 min. The structure of Al(OH) 3 added glasses was studied by 27 Al and 31 P MAS NMR and Fourier transform IR (FTIR) spectroscopies to explain the relationship between glass properties and composition. The 27 Al spectra show three different aluminum environments [Al(OP) 6 , Al(OP) 5 and Al(OP) 4 ], and Al(OP) 6 species seem to be predominant in these glasses. The formation of P---O-Al covalent bond and more strongly ionic bonds increase the strength of the glass network in stannous calcium aluminofluorophosphate glasses. FTIR spectra indicate the presence of P--F bond in the short range glass structure

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  1. Remote Fiber Laser Cutting System for Dismantling Glass Melter - 13071

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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)

  3. Ceramic fiber-reinforced monoclinic celsian phase glass-ceramic matrix composite material

    Science.gov (United States)

    Bansal, Narottam P. (Inventor); Dicarlo, James A. (Inventor)

    1994-01-01

    A hyridopolysilazane-derived ceramic fiber reinforced monoclinic celsian phase barium aluminum silicate glass-ceramic matrix composite material is prepared by ball-milling an aqueous slurry of BAS glass powder and fine monoclinic celsian seeds. The fibers improve the mechanical strength and fracture toughness and with the matrix provide superior dielectric properties.

  4. TWRS HLW interim storage facility search and evaluation

    Energy Technology Data Exchange (ETDEWEB)

    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.

  5. Immobilisation of radio cesium loaded ammonium molybdo phosphate in glass matrices

    International Nuclear Information System (INIS)

    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)

  6. Effect of aluminum and silicon reactants and process parameters on glass-ceramic waste form characteristics for immobilization of high-level fluorinel-sodium calcined waste

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1993-06-01

    In this report, the effects of aluminum and silicon reactants, process soak time and the initial calcine particle size on glass-ceramic waste form characteristics for immobilization of the high-level fluorinel-sodium calcined waste stored at the Idaho Chemical Processing Plant (ICPP) are investigated. The waste form characteristics include density, total and normalized elemental leach rates, and microstructure. Glass-ceramic waste forms were prepared by hot isostatically pressing (HIPing) a pre-compacted mixture of pilot plant fluorinel-sodium calcine, Al, and Si metal powders at 1050 degrees C, 20,000 psi for 4 hours. One of the formulations with 2 wt % Al was HIPed for 4, 8, 16 and 24 hours at the same temperature and pressure. The calcine particle size range include as calcined particle size smaller than 600 μm (finer than -30 mesh, or 215 μm Mass Median Diameter, MMD) and 180 μm (finer than 80 mesh, or 49 μm MMD)

  7. Challenges in development of matrices for vitrification of old legacy waste and high-level radioactive waste generated from reprocessing of AHWR and FBR spent fuel

    International Nuclear Information System (INIS)

    Kaushik, C.P.

    2012-01-01

    Majority of radioactivity in entire nuclear fuel cycle is concentrated in HLW. A three step strategy for management of HLW has been adopted in India. This involves immobilization of waste oxides in stable and inert solid matrices, interim retrievable storage of the conditioned waste product under continuous cooling and disposal in deep geological formations. Glass has been accepted as most suitable matrix world-wide for immobilization of HLW, because of its attractive features like ability to accommodate wide range of waste constituents, modest processing temperatures, adequate chemical, thermal and radiation stability. Borosilicate glass matrix developed by BARC in collaboration with CGCRI has been adopted in India for immobilization of HLW. In view of compositional variation of HLW from site to site, tailor make changes in the glass formulations are often necessary to incorporate all the waste constituents and having the product of desirable characteristics. The vitrified waste products made with different glass formulations and simulated waste need to be characterized for chemical durability, thermal stability, homogeneity etc. before finalizing a suitable glass formulation. The present extended abstract summarises the studies carried out for development of glass formulations for vitrification of legacy waste and futuristic waste likely to be generated from AHWR and FBR having wide variations in their compositions. The presently stored HLW at Trombay is characterized by significant concentrations of uranium, sodium and sulphate in addition to fission products, corrosion products and small amount of other actinides

  8. HLW Long-term Management Technology Development

    International Nuclear Information System (INIS)

    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

  9. Effect of irradiation on the evolution of alteration layer formed during nuclear glass leaching

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  11. Invisible Display in Aluminum

    DEFF Research Database (Denmark)

    Prichystal, Jan Phuklin; Hansen, Hans Nørgaard; Bladt, Henrik Henriksen

    2005-01-01

    Bang & Olufsen a/s has been working with ideas for invisible integration of displays in metal surfaces. Invisible integration of information displays traditionally has been possible by placing displays behind transparent or semitransparent materials such as plastic or glass. The wish for an integ......Bang & Olufsen a/s has been working with ideas for invisible integration of displays in metal surfaces. Invisible integration of information displays traditionally has been possible by placing displays behind transparent or semitransparent materials such as plastic or glass. The wish...... for an integrated display in a metal surface is often ruled by design and functionality of a product. The integration of displays in metal surfaces requires metal removal in order to clear the area of the display to some extent. The idea behind an invisible display in Aluminum concerns the processing of a metal...

  12. Strategic management of HLW repository projects

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  14. Fluorescent lighting with aluminum nitride phosphors

    Science.gov (United States)

    Cherepy, Nerine J.; Payne, Stephen A.; Seeley, Zachary M.; Srivastava, Alok M.

    2016-05-10

    A fluorescent lamp includes a glass envelope; at least two electrodes connected to the glass envelope; mercury vapor and an inert gas within the glass envelope; and a phosphor within the glass envelope, wherein the phosphor blend includes aluminum nitride. The phosphor may be a wurtzite (hexagonal) crystalline structure Al.sub.(1-x)M.sub.xN phosphor, where M may be drawn from beryllium, magnesium, calcium, strontium, barium, zinc, scandium, yttrium, lanthanum, cerium, praseodymium, europium, gadolinium, terbium, ytterbium, bismuth, manganese, silicon, germanium, tin, boron, or gallium is synthesized to include dopants to control its luminescence under ultraviolet excitation. The disclosed Al.sub.(1-x)M.sub.xN:Mn phosphor provides bright orange-red emission, comparable in efficiency and spectrum to that of the standard orange-red phosphor used in fluorescent lighting, Y.sub.2O.sub.3:Eu. Furthermore, it offers excellent lumen maintenance in a fluorescent lamp, and does not utilize "critical rare earths," minimizing sensitivity to fluctuating market prices for the rare earth elements.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  19. Anti-scratch AlMgB14 Gorilla® Glass coating

    Science.gov (United States)

    Putrolaynen, V. V.; Grishin, A. M.; Rigoev, I. V.

    2017-10-01

    Hard aluminum-magnesium boride (BAM) films were fabricated onto Corning® Gorilla® Glass by radio-frequency magnetron sputtering of a single stoichiometric AlMgB14 target. BAM films exhibit a Vickers hardness from 10 to 30 GPa and a Young's modulus from 80 to 160 GPa depending on applied loading forces. Deposited hard coating increases the critical load at which glass substrate cracks. The adhesion energy of BAM films on Gorilla® Glass is 6.4 J/m2.

  20. Aluminum-free glass-ionomer bone cements with enhanced bioactivity and biodegradability

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Filipa O.; Pires, Ricardo A., E-mail: rpires@dep.uminho.pt; Reis, Rui L.

    2013-04-01

    Al-free glasses of general composition 0.340SiO{sub 2}:0.300ZnO:(0.250-a-b)CaO:aSrO:bMgO:0.050Na{sub 2}O:0.060P{sub 2}O{sub 5} (a, b = 0.000 or 0.125) were synthesized by melt quenching and their ability to form glass-ionomer cements was evaluated using poly(acrylic acid) and water. We evaluated the influence of the poly(acrylic acid) molecular weight and glass particle size in the cement mechanical performance. Higher compressive strength (25 ± 5 MPa) and higher compressive elastic modulus (492 ± 17 MPa) were achieved with a poly(acrylic acid) of 50 kDa and glass particle sizes between 63 and 125 μm. Cements prepared with glass formulation a = 0.125 and b = 0.000 were analyzed after immersion in simulated body fluid; they presented a surface morphology consistent with a calcium phosphate coating and a Ca/P ratio of 1.55 (similar to calcium-deficient hydroxyapatite). Addition of starch to the cement formulation induced partial degradability after 8 weeks of immersion in phosphate buffer saline containing α-amylase. Micro-computed tomography analysis revealed that the inclusion of starch increased the cement porosity from 35% to 42%. We were able to produce partially degradable Al-free glass-ionomer bone cements with mechanical performance, bioactivity and biodegradability suitable to be applied on non-load bearing sites and with the appropriate physical characteristics for osteointegration upon partial degradation. Zn release studies (concentrations between 413 μM and 887 μM) evidenced the necessity to tune the cement formulations to reduce the Zn concentration in the surrounding environment. Highlights: ► We developed partially degradable, bioactive, Al-free glass-ionomer cements (GICs). ► Enhanced mechanical behavior was achieved using 63–125 μm glass particle size range. ► The highest mechanical resistance was obtained using poly(acrylic acid) of 50 kDa. ► Biodegradation was successfully tuned to start 8 weeks after GIC preparation. ► Zn

  1. Aluminum-free glass-ionomer bone cements with enhanced bioactivity and biodegradability

    International Nuclear Information System (INIS)

    Gomes, Filipa O.; Pires, Ricardo A.; Reis, Rui L.

    2013-01-01

    Al-free glasses of general composition 0.340SiO 2 :0.300ZnO:(0.250-a-b)CaO:aSrO:bMgO:0.050Na 2 O:0.060P 2 O 5 (a, b = 0.000 or 0.125) were synthesized by melt quenching and their ability to form glass-ionomer cements was evaluated using poly(acrylic acid) and water. We evaluated the influence of the poly(acrylic acid) molecular weight and glass particle size in the cement mechanical performance. Higher compressive strength (25 ± 5 MPa) and higher compressive elastic modulus (492 ± 17 MPa) were achieved with a poly(acrylic acid) of 50 kDa and glass particle sizes between 63 and 125 μm. Cements prepared with glass formulation a = 0.125 and b = 0.000 were analyzed after immersion in simulated body fluid; they presented a surface morphology consistent with a calcium phosphate coating and a Ca/P ratio of 1.55 (similar to calcium-deficient hydroxyapatite). Addition of starch to the cement formulation induced partial degradability after 8 weeks of immersion in phosphate buffer saline containing α-amylase. Micro-computed tomography analysis revealed that the inclusion of starch increased the cement porosity from 35% to 42%. We were able to produce partially degradable Al-free glass-ionomer bone cements with mechanical performance, bioactivity and biodegradability suitable to be applied on non-load bearing sites and with the appropriate physical characteristics for osteointegration upon partial degradation. Zn release studies (concentrations between 413 μM and 887 μM) evidenced the necessity to tune the cement formulations to reduce the Zn concentration in the surrounding environment. Highlights: ► We developed partially degradable, bioactive, Al-free glass-ionomer cements (GICs). ► Enhanced mechanical behavior was achieved using 63–125 μm glass particle size range. ► The highest mechanical resistance was obtained using poly(acrylic acid) of 50 kDa. ► Biodegradation was successfully tuned to start 8 weeks after GIC preparation. ► Zn release should be

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

    International Nuclear Information System (INIS)

    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)

  3. Spent fuel and HLW transportation the French experience

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  5. 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

    International Nuclear Information System (INIS)

    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

  6. Quartz glass behavior at ultraviolet spectrum region

    International Nuclear Information System (INIS)

    Braga, A.F.B.; Barbosa, L.C.; Evora, C.A.P.D.

    1990-01-01

    Melted quartz ingots were produced from raw materials of different sources. Behavior studies of these three different quartz glass were made at the eletromagnetic spectrum ultraviolet region. The atomic absorption spectroscopy was used as an analysis technique of the alkaline, transition and aluminum metal traces. It was found that the alkaline, transition and aluminum metals impurities present a great influence on the melted quartz spectral behavior at the ultraviolet region. It was stated that measurments at this spectrum region constitute an important characterization technique to natural quartz as well as melted quartz. (author) [pt

  7. Dissolution of lanthanide alumino-silicate oxynitride glasses

    Science.gov (United States)

    Bois, L.; Barré, N.; Guillopé, S.; Guittet, M. J.; Gautier-Soyer, M.; Duraud, J. P.; Trocellier, P.; Verdier, P.; Laurent, Y.

    2000-01-01

    The aqueous corrosion behavior of lanthanide aluminosilicate glasses has been studied under static conditions ( T=96°C, duration=1 and 3 months, glass surface area/leachate volume, S/ V=0.3 cm -1) by means of solution and solid analyses. It was found that these glasses exhibit a high chemical durability. The influence of yttrium, magnesium and nitrogen, which are supposed to improve the mechanical properties, on the chemical durability, has been investigated. After a one-month experiment, lanthanum and yttrium releases were found to be about 10 -7 mol l -1, while silicon and aluminum releases were about 10 -5 mol l -1. Yttrium seems to improve the chemical durability. The presence of nitrogen does not seem to modify the glass constituents releases, but seems to improve the surface state of the altered glass. XPS experiments reveal that lanthanum and yttrium are more concentrated near the surface (20-30 Å) of the glass after the leaching test.

  8. 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

    International Nuclear Information System (INIS)

    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

  9. Chemical Composition Analysis of INEEL Phase 3 Glasses: Task Technical and QA Plan

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  11. A Strategy for Maintenance of the Long-Term Performance Assessment of Immobilized Low-Activity Waste Glass

    Energy Technology Data Exchange (ETDEWEB)

    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).

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  14. Corrosion properties of aluminum based alloys deposited by ion beam assisted deposition

    International Nuclear Information System (INIS)

    Enders, B.; Krauss, S.; Wolf, G.K.

    1994-01-01

    The replacement of cadmium coatings by other protective measures is an important task because of the environmentally detrimental properties of cadmium. Therefore, aluminum and aluminum alloy coatings containing elements such as silicon or magnesium with more positive or negative positions in the galvanic series in relation to pure aluminum were deposited by ion beam assisted deposition onto glass and low carbon steel. Pure aluminum films were deposited onto low carbon steel in order to study the influence of the ion-to-atom arrival ratio and the angle of ion incidence on the corrosion properties. For examination of the pitting behavior as a function of the concentration of alloying element, quasipotentiostatic current-potential and potentiostatic current-time plots were measured in chlorine-containing acetate buffer. It is shown that these alloys can protect steel substrates under uniform and pitting corrosion conditions considerably better than pure aluminum coatings. ((orig.))

  15. Structure-Property Relationships and the Mixed Network Former Effect in Boroaluminosilicate Glasses

    DEFF Research Database (Denmark)

    Zheng, Qiuju; Potuzak, Marcel; Mauro, John C.

    compositions by substituting Al2O3 for SiO2. We also investigate the various roles of sodium in the glasses including charge compensation of tetrahedral aluminum and boron atoms and formation of non-bridging oxygen. We find that mechanical properties (density, elastic moduli, and hardness), glass transition......Boroaluminosilicate glasses are important materials for various applications, e.g., liquid crystal display substrates, glass fibers for reinforcement, and thermal shock-resistant glass containers. The complicated structural speciation in these glasses leads to a mixed network former effect yielding...... nonlinear variation in many macroscopic properties. It is therefore crucial to investigate and understand structure-property correlations in boroaluminosilicate glasses. Here we study the structure-property relationships of a range of sodium boroaluminosilicate glasses from peralkaline to peraluminous...

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

    International Nuclear Information System (INIS)

    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)

  17. Design and validation of the THMC China-Mock-Up test on buffer material for HLW disposal

    Directory of Open Access Journals (Sweden)

    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.

  18. Plasmonic Glasses and Films Based on Alternative Inexpensive Materials for Blocking Infrared Radiation.

    Science.gov (United States)

    V Besteiro, Lucas; Kong, Xiang-Tian; Wang, Zhiming; Rosei, Federico; Govorov, Alexander O

    2018-04-16

    The need for energy-saving materials is pressing. This Letter reports on the design of energy-saving glasses and films based on plasmonic nanocrystals that efficiently block infrared radiation. Designing such plasmonic composite glasses is nontrivial and requires taking full advantage of both material and geometrical properties of the nanoparticles. We compute the performance of solar plasmonic glasses incorporating a transparent matrix and specially shaped nanocrystals. This performance depends on the shape and material of such nanocrystals. Glasses designed with plasmonic nanoshells are shown to exhibit overall better performances as compared to nanorods and nanocups. Simultaneously, scalable synthesis of plasmonic nanoshells and nanocups is technologically feasible using gas-phase fabrication methods. The computational simulations were performed for noble metals (gold and silver) as well as for alternative plasmonic materials (aluminum, copper, and titanium nitride). Inexpensive plasmonic materials (silver, copper, aluminum, and titanium nitride) show an overall good performance in terms of the commonly used figures of merit of industrial glass windows. Together with numerical data for specific materials, this study includes a set of general rules for designing efficient plasmonic IR-blocking media. The plasmonic glasses proposed herein are good candidates for the creation of cheap optical media, to be used in energy-saving windows in warm climates' housing or temperature-sensitive infrastructure.

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

    International Nuclear Information System (INIS)

    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

  20. Fluorescence yield in rare-earth-doped sol-gel silicate glasses

    Energy Technology Data Exchange (ETDEWEB)

    Silversmith, A.J., E-mail: asilvers@hamilton.ed [Physics Department, Hamilton College, 198 College Hill Road, Clinton, NY 13323 (United States); Nguyen, Nguyen T.T.; Campbell, D.L. [Physics Department, Hamilton College, 198 College Hill Road, Clinton, NY 13323 (United States); Boye, D.M.; Ortiz, C.P. [Davidson College, Davidson, NC 28035 (United States); Hoffman, K.R. [Whitman College, Walla Walla, WA 99362 (United States)

    2009-12-15

    We have used trivalent terbium to investigate the mechanism behind fluorescence enhancement by Al{sup 3+} co-doping. Our results indicate that rare-earth (RE) ions cluster together in aluminum-rich regions of the glass, and behave as if they were dispersed uniformly throughout these regions when the ratio of Al to RE is {approx}10 or greater. We also studied the effects of adding chemical drying agents to the precursor solution for the synthesis of sol-gel-derived silicate glasses. Such glasses can be treated at significantly higher annealing temperatures without degradation of optical quality, and have the density of melt glass. Fluorescence yield from doped RE ions improves markedly with the addition of the drying agents, and the denser glasses are not subject to rehydration.

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

    International Nuclear Information System (INIS)

    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

  2. Advances on aluminum first-surface solar reflectors

    Science.gov (United States)

    Almanza, Rafael; Chen, Jiefeng; Mazari, Marcos

    1992-11-01

    Aluminum first surface mirrors have some advantages over second surface mirrors as has been discussed. At this stage of development some advantages are obtained: the first advantage was using two electron guns, one for aluminum evaporation permitting us to eliminate or to minimize the pinholes and the other to allow the evaporation of SiO without any mirror contamination as it was before due to the air when the chamber was opened to introduce the SiO, despite having only one e-gun in the laboratory. The second advantage was a better adherence between the aluminum film and the Si2O3, this last substance obtained with an oxidation of SiO with some oxygen inside the evaporation chamber (10-4 Torr). This improvement was due to the use of two e-guns that permit us not to open the chamber. These mirrors are actually under test in the environmental chamber for accelerated weather evaluations. One important aspect is the cleaning of the glass substrate. The chromic mixture cleaning is one of the most effective.

  3. Aluminum precipitation from Hanford DSSF

    International Nuclear Information System (INIS)

    Borgen, D.; Frazier, P.; Staton, G.

    1994-01-01

    A series of pilot scale tests using simulated Double Shell Slurry Feed (DSSF) showed that well-settled aluminum precipitate can be produced in Hanford double shell tank (DST) high level waste by slow neutralization with carbon dioxide. This pretreatment could provide an early grout feed and free tank space, as well as facilitate downstream processes such as ion exchange by providing a less caustic feed. A total of eight test runs were completed using a 10-ft tall 3-in i.d. glass column. The 10-ft height corresponds to about one third of the vertical height of a DST, hence providing a reasonable basis for extrapolating the observed precipitate settling and compaction to the actual waste tank environment. Four runs (three with a simplified simulant and one with a chemically complete simulant) produced well settled precipitates averaging 1.5 to 2 feet high. Aluminum gel rather than settled precipitate resulted from one test where neutralization was too rapid

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1993-06-01

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

  8. Low Temperature Curing of Hydrogen Silsesquioxane Surface Coatings for Corrosion Protection of Aluminum

    DEFF Research Database (Denmark)

    Lampert, Felix; Jensen, Annemette Hindhede; Møller, Per

    2016-01-01

    Hydrogen Silsesquioxane (HSQ) has shown to be a promising precursor for corrosion protective glass coatings for metallic substrates due to the excellent barrier properties of the films, especially in the application of protective coatings for aluminum in the automotive industry where high chemica...

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

    International Nuclear Information System (INIS)

    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

  10. Sealing glasses for titanium and titanium alloys

    Science.gov (United States)

    Brow, Richard K.; McCollister, Howard L.; Phifer, Carol C.; Day, Delbert E.

    1997-01-01

    Barium lanthanoborate sealing-glass compositions are provided comprising various combinations (in terms of mole-%) of boron oxide (B.sub.2 O.sub.3), barium oxide (BaO), lanthanum oxide (La.sub.2 O.sub.3), and at least one other oxide selected from the group consisting of aluminum oxide (Al.sub.2 O.sub.3), calcium oxide (CaO), lithium oxide (Li.sub.2 O), sodium oxide (Na.sub.2 O), silicon dioxide (SiO.sub.2), or titanium dioxide (TiO.sub.2). These sealing-glass compositions are useful for forming hermetic glass-to-metal seals with titanium and titanium alloys having an improved aqueous durability and favorable sealing characteristics. Examples of the sealing-glass compositions are provided having coefficients of thermal expansion about that of titanium or titanium alloys, and with sealing temperatures less than about 900.degree. C., and generally about 700.degree.-800.degree. C. The barium lanthanoborate sealing-glass compositions are useful for components and devices requiring prolonged exposure to moisture or water, and for implanted biomedical devices (e.g. batteries, pacemakers, defibrillators, pumps).

  11. Advanced waste form and Melter development for treatment of troublesome high-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (also with high Al2O3 concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both

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

    International Nuclear Information System (INIS)

    Ledieu, A.

    2004-10-01

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

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

    International Nuclear Information System (INIS)

    Ledieu, A.

    2004-10-01

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

  14. Preparation and spectroscopic properties of Yb-doped and Yb-Al-codoped high silica glasses

    International Nuclear Information System (INIS)

    Qiao Yanbo; Wen Lei; Wu Botao; Ren Jinjun; Chen Danping; Qiu Jianrong

    2008-01-01

    Yb-doped and Yb-Al-codoped high silica glasses have been prepared by sintering nanoporous glasses. The absorption, fluorescent spectra and fluorescent lifetimes have been measured and the emission cross-section and minimum pump intensities were calculated. Codoping aluminum ions enhanced the fluorescence intensity of Yb-doped high silica glass obviously. The emission cross-sections of Yb-doped and Yb-Al-codoped high silica glasses were 0.65 and 0.82 pm 2 , respectively. The results show that Yb-Al-codoped high silica glass has better spectroscopic properties for a laser material. The study of high silica glass doped with ytterbium is helpful for its application in Yb laser systems, especially for high-power and high-repetition lasers

  15. Development of geological disposal system; localization of element cost data and cost evaluation on the HLW repository

    Energy Technology Data Exchange (ETDEWEB)

    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)

  16. Effects of aluminum on root growth and absorption of nutrients by ...

    African Journals Online (AJOL)

    USER

    2010-06-28

    Jun 28, 2010 ... In Taiwan, pineapple (Ananas comosus (L.) Merr) is one of important fruits in sacrifice and daily consumption (Chang, 1995). Most of the pineapples were ..... culture. J. Plant Nutr. 17: 953-962. Nichol BE, Oliveira LA, Glass ADM, Siddiqi MY (1993). The effects of aluminum on the influx of calcium, potassium, ...

  17. Structure of high alumina content Al2O3-SiO2 composition glasses.

    Science.gov (United States)

    Weber, Richard; Sen, Sabyasachi; Youngman, Randall E; Hart, Robert T; Benmore, Chris J

    2008-12-25

    The structure of binary aluminosilicate glasses containing 60-67 mol % Al2O3 were investigated using high-resolution 27Al NMR and X-ray and neutron diffraction. The glasses were made by aerodynamic levitation of molten oxides. The 67% alumina composition required a cooling rate of approximately 1600 degrees C s(1-) to form glass from submillimeter sized samples. NMR results show that the glasses contain aluminum in 4-, 5-, and 6-fold coordination in the approximate ratio 4:5:1. The average Al coordination increases from 4.57 to 4.73 as the fraction of octahedral Al increases with alumina content. The diffraction results on the 67% composition are consistent with a disordered Al framework with Al ions in a range of coordination environments that are substantially different from those found in the equilibrium crystalline phases. Analysis of the neutron and X-ray structure factors yields an average bond angle of 125 +/- 4 degrees between an Al ion and the adjoining cation via a bridging oxygen. We propose that the structure of the glass is a "transition state" between the alumina-rich liquid and the equilibrium mullite phase that are dominated by 4- and 6-coordinated aluminum ions, respectively.

  18. Can-in-canister cold demonstration in DWPF (U)

    International Nuclear Information System (INIS)

    Kuehn, N.H.

    1996-07-01

    The Department of Energy Fissile Materials Disposition Program is evaluating a number of options for disposition of weapons-usable plutonium surplus to national defense needs. One of the immobilization options is the Can-In-Canister approach. In this option small cans of a plutonium glass, which contains a neutron absorber, are placed on a support structure in a large Savannah River Site Defense Waste Processing Facility (DWPF) canister. The top is then welded onto the canister. This canister is filled with High Level Waste (HLW) glass at the DWPF. The HLW glass provides the radiation source for proliferation resistance. These canisters are to be placed in a Federal Repository. To provide information on the technical feasibility of this option prior to the Record of Decision on plutonium disposition, the Department of Energy Fissile Materials Disposition Program funded a demonstration in the DWPF. This demonstration was conducted before the start of radioactive operations. Two test canisters containing cans of surrogate (non- radioactive) plutonium glass were successfully filled with simulated HLW glass at the DWPF using standard pouring procedures. One canister had twenty cans of surrogate plutonium glass. The other had eight cans of surrogate plutonium glass. After the canisters were filled, the contents of the canisters were examined to provide data on the effect of the rack and cans on the filling of the DWPF canister, the effect of the pour on the surrogate plutonium glass and the effect of the rack and cans on the simulated HLW glass. There was no deformation of the support racks during the pour. The simulated HLW glass filled all the regions around the rack and cans and the regions between the cans and the wall of the canister. This report discusses the design of the racks and cans, the modification of the DWPF canisters to accommodate the rack and cans, the conditions during the pours and the results of the post pour analysis

  19. Waste vitrification: a historical perspective

    International Nuclear Information System (INIS)

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

    1982-02-01

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

  20. Nitrogen bonding in aluminum oxynitride films

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Paul W., E-mail: pwang@bradley.edu [Department of Physics, Bradley University, 1501 W. Bradley Ave., Peoria, IL 61625 (United States); Hsu, Jin-Cherng [Department of Physics, Fu Jen Catholic University, Hsinchuang, Taipei Hsien 24205, Taiwan (China); Lin, Yung-Hsin; Chen, Huang-Lu [Graduate Institute of Applied Science and Engineering, Fu Jen Catholic University, Hsinchuang, Taipei Hsien 24205, Taiwan (China)

    2010-04-15

    Assignment of oxidation states of N{sub 1s} in XPS spectra of aluminum oxynitride by curve fitting is difficult. The XPS curve fitting was previously discussed in the paper published in J. Non-Cryst. Solids, 224 (1998) 31, in which O{sub 1s} photoelectrons from GeO{sub 2} glass were used to illustrate how to fit the XPS spectra. Three different ways were pointed out to eliminate the ambiguity caused by curve fitting such as comparing the data to data from standard samples, investigating the continuous surface modifications caused by slowly sputtering the surface, and monitoring the continuous surface modifications due to gradual increases in surface species under heating, cooling, or irradiation. Our recent work in aluminum oxynitride films provides another example of how to fit the XPS spectra of N{sub 1s} by three different oxidation states of N{sup +}, N{sup 2+}, and N{sup 3+}, by comparison of the measured data to data from previously published results, and by the gradual changes of spectra as functions of the oxygen contents in the films. Three oxidation states in different nitrogen bonding in the aluminum oxynitride, AlO{sub 2}N, Al{sub 2}O{sub 5}N{sub 2}, and AlO{sub 3}N, were clearly deduced.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-30

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

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

    International Nuclear Information System (INIS)

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

    1995-04-01

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

  3. Augmenting Scheffe Linear Mixture Models With Squared and/or Crossproduct Terms

    International Nuclear Information System (INIS)

    Piepel, Gregory F.; Szychowski, Jeffrey M.; Loeppky, Jason L.

    2001-01-01

    A glass composition variation study (CVS) for high-level waste (HLW) stored at the Idaho National Engineering and Environmental Laboratory (INEEL) is being statistically designed and performed in phases over several years. The purpose of the CVS is to investigate and model how HLW-glass properties depend on glass composition within a glass composition region compatible with the expected range of INEEL HLW. The resulting glass property-composition models will be used to develop desirable glass formulations and other purposes. Phases 1 and 2 of the CVS have been completed so far, and are briefly described. The main focus of this paper is the CVS Phase 3 experimental design (test matrix). The Phase 3 experimental design was chosen to augment the Phase 1 and 2 data with additional data points, as well as to account for additional glass components of interest not studied in Phases 1 and/or 2. The paper describes how these Phase 3 experimental design augmentation challenges were addressed using the previous data, preliminary property-composition models, and statistical mixture experiment and optimal experimental design methods and software. The resulting Phase 3 experimental design of 30 simulated HAW glasses is presented and discussed

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

    International Nuclear Information System (INIS)

    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.)

  5. Measurement of ac electrical conductivity of molten glass by impedance measurement using co-axial cylinder electrode

    International Nuclear Information System (INIS)

    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. Aluminum-doped zinc oxide films as transparent conductive electrode for organic light-emitting devices

    International Nuclear Information System (INIS)

    Jiang, X.; Wong, F.L.; Fung, M.K.; Lee, S.T.

    2003-01-01

    Highly transparent conductive, aluminum-doped zinc oxide (ZnO:Al) films were deposited on glass substrates by midfrequency magnetron sputtering of metallic aluminum-doped zinc target. ZnO:Al films with surface work functions between 3.7 and 4.4 eV were obtained by varying the sputtering conditions. Organic light-emitting diodes (OLEDs) were fabricated on these ZnO:Al films. A current efficiency of higher than 3.7 cd/A, was achieved. For comparison, 3.9 cd/A was achieved by the reference OLEDs fabricated on commercial indium-tin-oxide substrates

  7. Advanced waste form and melter development for treatment of troublesome high-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-02

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these "troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approached to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.

  8. Sludge Heel Removal By Aluminum Dissolution At Savannah River Site 12390

    International Nuclear Information System (INIS)

    Keefer, M.

    2012-01-01

    High Level Waste (HLW) at the Savannah River Site (SRS) is currently stored in aging underground storage tanks. This waste is a complex mixture of insoluble solids, referred to as sludge, and soluble salts. Continued long-term storage of these radioactive wastes poses an environmental risk. Operations are underway to remove and disposition the waste, clean the tanks and fill with grout for permanent closure. Heel removal is the intermediate phase of the waste retrieval and tank cleaning process at SRS, which is intended to reduce the volume of waste prior to treatment with oxalic acid. The goal of heel removal is to reduce the residual amount of radioactive sludge wastes to less than 37,900 liters (10,000 gallons) of wet solids. Reducing the quantity of residual waste solids in the tank prior to acid cleaning reduces the amount of acid required and reduces the amount of excess acid that could impact ongoing waste management processes. Mechanical heel removal campaigns in Tank 12 have relied solely on the use of mixing pumps that have not been effective at reducing the volume of remaining solids. The remaining waste in Tank 12 is known to have a high aluminum concentration. Aluminum dissolution by caustic leaching was identified as a treatment step to reduce the volume of remaining solids and prepare the tank for acid cleaning. Dissolution was performed in Tank 12 over a two month period in July and August, 2011. Sample results indicated that 16,440 kg of aluminum oxide (boehmite) had been dissolved representing 60% of the starting inventory. The evolution resulted in reducing the sludge solids volume by 22,300 liters (5900 gallons), preparing the tank for chemical cleaning with oxalic acid.

  9. SLUDGE HEEL REMOVAL BY ALUMINUM DISSOLUTION AT SAVANNAH RIVER SITE 12390

    Energy Technology Data Exchange (ETDEWEB)

    Keefer, M.

    2012-01-12

    High Level Waste (HLW) at the Savannah River Site (SRS) is currently stored in aging underground storage tanks. This waste is a complex mixture of insoluble solids, referred to as sludge, and soluble salts. Continued long-term storage of these radioactive wastes poses an environmental risk. Operations are underway to remove and disposition the waste, clean the tanks and fill with grout for permanent closure. Heel removal is the intermediate phase of the waste retrieval and tank cleaning process at SRS, which is intended to reduce the volume of waste prior to treatment with oxalic acid. The goal of heel removal is to reduce the residual amount of radioactive sludge wastes to less than 37,900 liters (10,000 gallons) of wet solids. Reducing the quantity of residual waste solids in the tank prior to acid cleaning reduces the amount of acid required and reduces the amount of excess acid that could impact ongoing waste management processes. Mechanical heel removal campaigns in Tank 12 have relied solely on the use of mixing pumps that have not been effective at reducing the volume of remaining solids. The remaining waste in Tank 12 is known to have a high aluminum concentration. Aluminum dissolution by caustic leaching was identified as a treatment step to reduce the volume of remaining solids and prepare the tank for acid cleaning. Dissolution was performed in Tank 12 over a two month period in July and August, 2011. Sample results indicated that 16,440 kg of aluminum oxide (boehmite) had been dissolved representing 60% of the starting inventory. The evolution resulted in reducing the sludge solids volume by 22,300 liters (5900 gallons), preparing the tank for chemical cleaning with oxalic acid.

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

    International Nuclear Information System (INIS)

    El-Alaily, N.A.

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    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

  12. Distribution of oxides in a Zr-Cu-Ni-Al-Nb-Si bulk metallic glass

    Energy Technology Data Exchange (ETDEWEB)

    Heinrich, Jochen; Busch, Ralf [Chair of Metallic Materials, Saarland University, PO Box 151150, 66041 Saarbruecken (Germany); Mueller, Frank; Huefner, Stefan [Chair of Experimental Physics, Saarland University, PO Box 151150, 66041 Saarbruecken (Germany)

    2010-07-01

    The course of oxide presence with distance from the sample surface and bonding partner was studied for the bulk metallic glass with the nominal composition Zr{sub 57.9}Cu{sub 15.4}Ni{sub 12.7}Al{sub 10.2}Nb{sub 2.8}Si{sub 1} (at%) by X-ray photoelectron spectroscopy (XPS). Investigated specimens are taken from vacuum quench-cast rods subjected to oxidation at room temperature and atmosphere. Binding energies were determined in various depths using ion beam ablation of up to 100 nanometers. XPS spectra confirm oxidation primarily of the pure zirconium and aluminum constituents, all other peaks correspond to metallic bonds. While the surface area shows a passivating zirconia layer a few nanometers thick, oxygen is bonded predominantly with aluminum inside the bulk. Since the concentration of oxygen is a crucial factor in the crystallization behavior of bulk metallic glass forming liquids on basis of oxygen affine metals, so far only high purity materials were thought to be suitable. The findings in this study, however, are promising for alloys with industrial grade elements with sufficient glass forming ability. Comparisons of the alloy with differing oxygen content support the conclusion that aluminum acts as an appropriate scavenger for both adsorbed and large amounts of intrinsic oxygen in zirconium based amorphous metals.

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

    International Nuclear Information System (INIS)

    Neumann, W.

    1980-01-01

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

  14. Execution techniques and approach for high level radioactive waste disposal in Japan: Demonstration of geological disposal techniques and implementation approach of HLW project

    International Nuclear Information System (INIS)

    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)

  15. Oxide glass structure evolution under swift heavy ion irradiation

    International Nuclear Information System (INIS)

    Mendoza, C.; Peuget, S.; Charpentier, T.; Moskura, M.; Caraballo, R.; Bouty, O.; Mir, A.H.; Monnet, I.; Grygiel, C.; Jegou, C.

    2014-01-01

    Highlights: • Structure of SHI irradiated glass is similar to the one of a hyper quenched glass. • D2 Raman band associated to 3 members ring is only observed in irradiated glass. • Irradiated state seems slightly different to an equilibrated liquid quenched rapidly. - Abstract: The effects of ion tracks on the structure of oxide glasses were examined by irradiating a silica glass and two borosilicate glass specimens containing 3 and 6 oxides with krypton ions (74 MeV) and xenon ions (92 MeV). Structural changes in the glass were observed by Raman and nuclear magnetic resonance spectroscopy using a multinuclear approach ( 11 B, 23 Na, 27 Al and 29 Si). The structure of irradiated silica glass resembles a structure quenched at very high temperature. Both borosilicate glass specimens exhibited depolymerization of the borosilicate network, a lower boron coordination number, and a change in the role of a fraction of the sodium atoms after irradiation, suggesting that the final borosilicate glass structures were quenched from a high temperature state. In addition, a sharp increase in the concentration of three membered silica rings and the presence of large amounts of penta- and hexacoordinate aluminum in the irradiated 6-oxide glass suggest that the irradiated glass is different from a liquid quenched at equilibrium, but it is rather obtained from a nonequilibrium liquid that is partially relaxed by very rapid quenching within the ion tracks

  16. HLW Feed Delivery AZ101 Batch Transfer to the Private Contractor Transfer and Mixing Process Improvements [Initial Release at Rev 2

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  18. Composition - structure - properties relationships of peraluminous glasses for nuclear waste containment

    International Nuclear Information System (INIS)

    Piovesan, Victor

    2016-01-01

    Part of the Research and Development program concerning high level nuclear waste conditioning aims to assess new glass formulations able to incorporate a high waste content with enhanced properties in terms of homogeneity, thermal stability, long term behavior and process ability. This study focuses on peraluminous glasses, defined by an excess of aluminum ions Al"3"+ in comparison with modifier elements such as Na"+, Li"+ or Ca"2"+. A Design of Experiment approach has been employed to determine relationships between composition of simplified peraluminous glasses (SiO_2 - B_2O_3 - Al_2O_3 - Na_2O - Li_2O - CaO - La_2O_3) and their physical properties such as viscosity, glass transition temperature and glass homogeneity. Moreover, some structural investigation (NMR) was performed in order to better understand the structural role of Na"+, Li"+ and Ca"2"+ and the structural organization of peraluminous glasses. Then, physical and chemical properties of fully simulated peraluminous glasses were characterized to evaluate transposition between simplified and fully simulated glasses and also to put forward the potential of peraluminous glasses for nuclear waste containment. (author) [fr

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

    Science.gov (United States)

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

    2017-05-01

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

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    Kuchinski, F.A.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    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'

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

    International Nuclear Information System (INIS)

    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

  5. Influence of Ag thickness of aluminum-doped ZnO/Ag/aluminum-doped ZnO thin films

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Hung-Wei, E-mail: hwwu@mail.ksu.edu.tw [Department of Computer and Communication, Kun Shan University, No. 949, Dawan Rd., Yongkang Dist., Tainan City 710, Taiwan (China); Yang, Ru-Yuan [Graduate Institute of Materials Engineering, National Pingtung University of Science and Technology, 1, Shuefu Rd., Neipu, Pingtung City 912, Taiwan (China); Hsiung, Chin-Min; Chu, Chien-Hsun [Department of Mechanical Engineering, National Pingtung University of Science and Technology, 1, Shuefu Rd., Neipu, Pingtung City 912, Taiwan (China)

    2012-10-01

    Highly conducting aluminum-doped ZnO (30 nm)/Ag (5-15 nm)/aluminum-doped ZnO (30 nm) multilayer thin films were deposited on glass substrate by rf magnetron sputtering (for top/bottom aluminum-doped ZnO films) and e-beam evaporation (for Ag film). The transmittance is more than 70% for wavelengths above 400 nm with the Ag layer thickness of 10 nm. The resistivity is 3.71 Multiplication-Sign 10{sup -4} {Omega}-cm, which can be decreased to 3.8 Multiplication-Sign 10{sup -5} {Omega}-cm with the increase of the Ag layer thickness to 15 nm. The Haacke figure of merit has been calculated for the films with the best value being 8 Multiplication-Sign 10{sup -3} {Omega}{sup -1}. It was shown that the multilayer thin films have potential for applications in optoelectronics. - Highlights: Black-Right-Pointing-Pointer High-quality Al-doped ZnO (AZO)/Ag/AZO Transparent Conducting Oxide films. Black-Right-Pointing-Pointer AZO films (30 nm) made by RF sputtering; E-beam evaporation for Ag film (5-15 nm). Black-Right-Pointing-Pointer Influence of Ag thickness on optical and electrical properties were analyzed. Black-Right-Pointing-Pointer High quality multilayer film with optimal intermediate Ag layer thickness of 10 nm. Black-Right-Pointing-Pointer 3.71 Multiplication-Sign 10{sup -4} {Omega}-cm resistivity, 91.89% transmittance at 470 nm obtained and reproducible.

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  7. SNF/HLW Transfer System Description Document

    International Nuclear Information System (INIS)

    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

  8. Final Report - IHLW PCT, Spinel T1%, Electrical Conductivity, and Viscosity Model Development, VSL-07R1240-4

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Piepel, Gregory F.; Landmesser, S. M.; Pegg, I. L.; Heredia-Langner, Alejandro; Cooley, Scott K.; Gan, H.; Kot, W. K.

    2013-11-13

    This report is the last in a series of currently scheduled reports that presents the results from the High Level Waste (HLW) glass formulation development and testing work performed at the Vitreous State Laboratory (VSL) of the Catholic University of America (CUA) and the development of IHLW property-composition models performed jointly by Pacific Northwest National Laboratory (PNNL) and VSL for the River Protection Project-Waste Treatment and Immobilization Plant (RPP-WTP). Specifically, this report presents results of glass testing at VSL and model development at PNNL for Product Consistency Test (PCT), one-percent crystal fraction temperature (T1%), electrical conductivity (EC), and viscosity of HLW glasses. The models presented in this report may be augmented and additional validation work performed during any future immobilized HLW (IHLW) model development work. Completion of the test objectives is addressed.

  9. Light transmission coefficients by subwavelength aluminum gratings with dielectric layers

    Energy Technology Data Exchange (ETDEWEB)

    Blinov, L. M., E-mail: lev39blinov@gmail.com; Lazarev, V. V.; Yudin, S. G.; Artemov, V. V.; Palto, S. P.; Gorkunov, M. V. [Russian Academy of Sciences, Shubnikov Institute of Crystallography (Crystallography and Photonics Federal Research Center) (Russian Federation)

    2016-11-15

    Spectral positions of plasmon resonances related to boundaries between a thin aluminum layer and dielectrics (air, glass, VDF–TrFE 65/35 ferroelectric copolymer, and indium tin oxide (ITO)) have been determined in the transmission spectra of aluminum gratings of three types with 30 × 30 μm{sup 2} dimensions and 350-, 400-, and 450-nm line periods. Experimental results agree well with spectral positions of plasmon resonances calculated for the normal incidence of TM-polarized light. In addition, maximum values of transmission coefficients in the region of λ ≈ 900–950 nm have been determined for glass–Al–copolymer and glass–ITO–Al–copolymer structures. These values are close to 100%, which shows that the effective optical aperture is two times greater than the geometric areas of slits.

  10. Estimation on separation efficiency of aluminum from base-cap of spent fluorescent lamp in hammer crusher unit.

    Science.gov (United States)

    Rhee, Seung-Whee

    2017-09-01

    In order to separate aluminum from the base-cap of spent fluorescent lamp (SFL), the separation efficiency of hammer crusher unit is estimated by introducing a binary separation theory. The base-cap of SFL is composed by glass fragment, binder, ferrous metal, copper and aluminum. The hammer crusher unit to recover aluminum from the base-cap consists of 3stages of hammer crusher, magnetic separator and vibrating screen. The optimal conditions of rotating speed and operating time in the hammer crusher unit are decided at each stage. At the optimal conditions, the aluminum yield and the separation efficiency of hammer crusher unit are estimated by applying a sequential binary separation theory at each stage. And the separation efficiency between hammer crusher unit and roll crush system is compared to show the performance of aluminum recovery from the base-cap of SFL. Since the separation efficiency can be increased to 99% at stage 3, from the experimental results, it is found that aluminum from the base-cap can be sufficiently recovered by the hammer crusher unit. Copyright © 2017. Published by Elsevier Ltd.

  11. Development of new radiopaque glass fiber posts

    Energy Technology Data Exchange (ETDEWEB)

    Furtos, Gabriel, E-mail: gfurtos@yahoo.co.uk [Raluca Ripan Institute of Research in Chemistry, Babes-Bolyai University, Cluj-Napoca (Romania); Baldea, Bogdan [Dep. of Prosthodontics, Faculty of Dental Medicine, Timisoara (Romania); Silaghi-Dumitrescu, Laura [Raluca Ripan Institute of Research in Chemistry, Babes-Bolyai University, Cluj-Napoca (Romania)

    2016-02-01

    The aim of this study was to analyze the radiopacity and filler content of three experimental glass fiber posts (EGFP) in comparison with other glass/carbon fibers and metal posts from the dental market. Three EGFP were obtained by pultrusion of glass fibers in a polymer matrix based on 2,2-bis[4-(2-hydroxy-3-methacryloyloxypropoxy)-phenyl]propane (bis-GMA) and triethyleneglycol dimethacrylate (TEGDMA) monomers. Using intraoral sensor disks 27 posts, as well as mesiodistal sections of human molar and aluminum step wedges were radiographed for evaluation of radiopacity. The percentage compositions of fillers by weight and volume were investigated by combustion analysis. Two EGFP showed radiopacity higher than enamel. The commercial endodontic posts showed radiopacity as follows: higher than enamel, between enamel and dentin, and lower than dentin. The results showed statistically significant differences (p < 0.05) when evaluated with one-way ANOVA statistical analysis. According to combustion analyses, the filler content of the tested posts ranges between 58.84 wt.% and 86.02 wt.%. The filler content of the tested EGFP ranged between 68.91 wt.% and 79.04 wt.%. EGFP could be an alternative to commercial glass fiber posts. Future glass fiber posts are recommended to present higher radiopacity than dentin and perhaps ideally similar to or higher than that of enamel, for improved clinical detection. The posts with a lower radiopacity than dentin should be considered insufficiently radiopaque. The radiopacity of some glass fiber posts is not greatly influenced by the amount of filler. - Highlights: • AR glass fibers for dental applications • AR glass fibers have a great potential for obtaining radiopaque glass fiber posts. • Experimental AR glass fiber posts could be an alternative to commercial glass fiber posts for clinical application.

  12. Development of new radiopaque glass fiber posts

    International Nuclear Information System (INIS)

    Furtos, Gabriel; Baldea, Bogdan; Silaghi-Dumitrescu, Laura

    2016-01-01

    The aim of this study was to analyze the radiopacity and filler content of three experimental glass fiber posts (EGFP) in comparison with other glass/carbon fibers and metal posts from the dental market. Three EGFP were obtained by pultrusion of glass fibers in a polymer matrix based on 2,2-bis[4-(2-hydroxy-3-methacryloyloxypropoxy)-phenyl]propane (bis-GMA) and triethyleneglycol dimethacrylate (TEGDMA) monomers. Using intraoral sensor disks 27 posts, as well as mesiodistal sections of human molar and aluminum step wedges were radiographed for evaluation of radiopacity. The percentage compositions of fillers by weight and volume were investigated by combustion analysis. Two EGFP showed radiopacity higher than enamel. The commercial endodontic posts showed radiopacity as follows: higher than enamel, between enamel and dentin, and lower than dentin. The results showed statistically significant differences (p < 0.05) when evaluated with one-way ANOVA statistical analysis. According to combustion analyses, the filler content of the tested posts ranges between 58.84 wt.% and 86.02 wt.%. The filler content of the tested EGFP ranged between 68.91 wt.% and 79.04 wt.%. EGFP could be an alternative to commercial glass fiber posts. Future glass fiber posts are recommended to present higher radiopacity than dentin and perhaps ideally similar to or higher than that of enamel, for improved clinical detection. The posts with a lower radiopacity than dentin should be considered insufficiently radiopaque. The radiopacity of some glass fiber posts is not greatly influenced by the amount of filler. - Highlights: • AR glass fibers for dental applications • AR glass fibers have a great potential for obtaining radiopaque glass fiber posts. • Experimental AR glass fiber posts could be an alternative to commercial glass fiber posts for clinical application.

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

    International Nuclear Information System (INIS)

    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. Fabrication of novel fiber reinforced aluminum composites by friction stir processing

    Energy Technology Data Exchange (ETDEWEB)

    Arab, Seyyed Mohammad; Karimi, Saeed; Jahromi, Seyyed Ahmad Jenabali, E-mail: jahromi@shirazu.ac.ir; Javadpour, Sirus; Zebarjad, Seyyed Mojtaba

    2015-04-24

    In this study, chopped and attrition milled high strength carbon, E-glass, and S-glass fibers have been used as the reinforcing agents in an aluminum alloy (Al1100) considered as the matrix. The Surface Metal Matrix Composites (SMMCs) then are produced by Friction Stir Processing (FSP). Tensile and micro-hardness examinations represent a magnificent improvement in the hardness, strength, ductility and toughness for all of the processed samples. Scanning Electron Micrographs reveal a proper distribution of the reinforcements in the matrix and a change in the fracture behavior of the FSPed specimens. The synergetic effects of reinforcing by fibers and Severe Plastic Deformation (SPD) lead to an extra ordinary improvement in the mechanical properties.

  15. Measurement of stress-induced birefringence in glasses based on reflective laser feedback effect

    Science.gov (United States)

    Haisha, Niu; YanXiong, Niu; Jiyang, Li

    2017-02-01

    A glass birefringence measurement system utilizing the reflective laser feedback (RLF) effect is presented. The measurement principle is analyzed based on the equivalent cavity of a Fabry-Perot interferometer, and the experiments are conducted with a piece of quartz glass with applied extrusion force. In the feedback system, aluminum film used as a feedback mirror is affixed to the back of the sample. When the light is reflected back into the cavity, as the reinjected light is imprinted with the birefringence information in the sample, the gain and polarization states of the laser are modulated. The variation of optical power and polarization states hopping is monitored to obtain the magnitude of the stress. The system has advantages such as simplicity and low-cost with a precision of 1.9 nm. Moreover, by adjusting the position of the aluminum, large-area samples can be measured anywhere at any place.

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

    International Nuclear Information System (INIS)

    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

  17. Technetium Immobilization Forms Literature Survey

    Energy Technology Data Exchange (ETDEWEB)

    Westsik, Joseph H.; Cantrell, Kirk J.; Serne, R. Jeffrey; Qafoku, Nikolla

    2014-05-01

    Of the many radionuclides and contaminants in the tank wastes stored at the Hanford site, technetium-99 (99Tc) is one of the most challenging to effectively immobilize in a waste form for ultimate disposal. Within the Hanford Tank Waste Treatment and Immobilization Plant (WTP), the Tc will partition between both the high-level waste (HLW) and low-activity waste (LAW) fractions of the tank waste. The HLW fraction will be converted to a glass waste form in the HLW vitrification facility and the LAW fraction will be converted to another glass waste form in the LAW vitrification facility. In both vitrification facilities, the Tc is incorporated into the glass waste form but a significant fraction of the Tc volatilizes at the high glass-melting temperatures and is captured in the off-gas treatment systems at both facilities. The aqueous off-gas condensate solution containing the volatilized Tc is recycled and is added to the LAW glass melter feed. This recycle process is effective in increasing the loading of Tc in the LAW glass but it also disproportionally increases the sulfur and halides in the LAW melter feed which increases both the amount of LAW glass and either the duration of the LAW vitrification mission or the required supplemental LAW treatment capacity.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    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. Phase transformations and thermodynamics of aluminum-based metallic glasses

    Science.gov (United States)

    Gao, Changhua (Michael)

    This thesis examines the thermodynamics and associated kinetics and phase transformations of the glass forming Al-Ni-Gd and Al-Fe-Gd systems. In order to fully understand the unique glass forming ability (GFA) of Al-based metallic glasses, the ternary Al-Fe-Gd and Al-Ni-Gd systems in their Al-rich corners were examined experimentally to assist in a thermodynamic assessment. The solid-state phase equilibria are determined using XRD and TEM-EDS techniques. While this work basically confirms the solid-state equilibria in Al-Fe-Gd reported previously, the ternary phase in Al-Ni-Gd system has been identified to be Al15Ni3Gd2 rather than Al16Ni 3Gd reported in the literature. DTA analysis of 24 alloys in the Al-Fe-Gd system and 42 alloys in the Al-Ni-Gd system have yielded critical temperatures pertaining to the solid-liquid transition. Based on these data and information from the literature, a self-consistent thermodynamic database for these systems has been developed using the CALPHAD technique. Parameters describing the Gibbs free energy for various phases of the Al-Gd, Al-Fe-Gd and Al-Ni-Gd systems are manually optimized in this study. Once constructed, the database is used to calculate driving forces for nucleation of crystalline phases which can qualitatively explain the phase formation sequence during crystallization at low temperatures. It was also confirmed that alloy compositions with the lowest Gibbs free energy difference between the equilibrium state and undercooled liquid state exhibit better GFA than other chemistries. Based on 250°C isothermal devitrification phase transformations of 17 Al-Ni-Gd alloys, a phase formation sequence map is constructed. Fcc-Al nanocrystals are formed first in most of the alloys studied, but eutectic crystallization of a metastable phase and fcc-Al is also observed. Addition of Al or Ni promotes fcc-Al phase formation, while increasing Gd suppresses it. The continuous heating DSC scans revealed that crystallization in Al

  2. A dose of HLW reality

    International Nuclear Information System (INIS)

    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

  3. Metallic Glasses as Potential Reinforcements in Al and Mg Matrices: A Review

    Directory of Open Access Journals (Sweden)

    S. Jayalakshmi

    2018-04-01

    Full Text Available Development of metal matrix composites (MMCs with metallic glass/amorphous alloy reinforcements is an emerging research field. As reinforcements, metallic glasses with their high strength (up to ~2 GPa and high elastic strain limit (~2% can provide superior mechanical properties. Being metallic in nature, the glassy alloys can ensure better interfacial properties when compared to conventional ceramic reinforcements. Given the metastable nature of metallic glasses, lightweight materials such as aluminum (Al and magnesium (Mg with relatively lower melting points are suitable matrix materials. Synthesis of these advanced composites is a challenge as selection of processing method and appropriate reinforcement type (which does not allow devitrification of the metallic glass during processing is important. Non-conventional techniques such as high frequency induction sintering, bidirectional microwave sintering, friction stir processing, accumulative roll-bonding, and spark plasma sintering are being explored to produce these novel materials. In this paper, an overview on the synthesis and properties of aluminum and magnesium based composites with glassy reinforcement produced by various unconventional methods is presented. Evaluation of properties of the produced composites indicate: (i retention of amorphous state of the reinforcement after processing; (ii significant improvement in hardness and strength; (iii improvement/retention of ductility; and (iv high wear resistance and low coefficient of friction. Further, a comparative understanding of the properties highlights that the selection of the processing method is important in producing high performance composites.

  4. Glass Polyalkenoate Cements Designed for Cranioplasty Applications: An Evaluation of Their Physical and Mechanical Properties

    Directory of Open Access Journals (Sweden)

    Basel A. Khader

    2016-03-01

    Full Text Available Glass polyalkenoate cements (GPCs have potential for skeletal cementation. Unfortunately, commercial GPCs all contain, and subsequently release, aluminum ions, which have been implicated in degenerative brain disease. The purpose of this research was to create a series of aluminum-free GPCs constructed from silicate (SiO2, calcium (CaO, zinc (ZnO and sodium (Na2O-containing glasses mixed with poly-acrylic acid (PAA and to evaluate the potential of these cements for cranioplasty applications. Three glasses were formulated based on the SiO2-CaO-ZnO-Na2O parent glass (KBT01 with 0.03 mol % (KBT02 and 0.06 mol % (KBT03 germanium (GeO2 substituted for ZnO. Each glass was then mixed with 50 wt % of a patented SiO2-CaO-ZnO-strontium (SrO glass composition and the resultant mixtures were subsequently reacted with aqueous PAA (50 wt % addition to produce three GPCs. The incorporation of Ge in the glass phase was found to result in decreased working (142 s to 112 s and setting (807 s to 448 s times for the cements manufactured from them, likely due to the increase in crosslink formation between the Ge-containing glasses and the PAA. Compressive (σc and biaxial flexural (σf strengths of the cements were examined at 1, 7 and 30 days post mixing and were found to increase with both maturation and Ge content. The bonding strength of a titanium cylinder (Ti attached to bone by the cements increased from 0.2 MPa, when placed, to 0.6 MPa, after 14 days maturation. The results of this research indicate that Germano-Silicate based GPCs have suitable handling and mechanical properties for cranioplasty fixation.

  5. Porous aluminum room temperature anodizing process in a fluorinated-oxalic acid solution

    Science.gov (United States)

    Dhahri, S.; Fazio, E.; Barreca, F.; Neri, F.; Ezzaouia, H.

    2016-08-01

    Anodizing of aluminum is used for producing porous insulating films suitable for different applications in electronics and microelectronics. Porous-type aluminum films are most simply realized by galvanostatic anodizing in aqueous acidic solutions. The improvement in application of anodizing technique is associated with a substantial reduction of the anodizing voltage at appropriate current densities as well as to the possibility to carry out the synthesis process at room temperature in order to obtain a self-planarizing dielectric material incorporated in array of super-narrow metal lines. In this work, the anodizing of aluminum to obtain porous oxide was carried out, at room temperature, on three different substrates (glass, stainless steel and aluminum), using an oxalic acid-based electrolyte with the addition of a relatively low amount of 0.4 % of HF. Different surface morphologies, from nearly spherical to larger porous nanostructures with smooth edges, were observed by means of scanning electron microscopy. These evidences are explained by considering the formation, transport and adsorption of the fluorine species which react with the Al3+ ions. The behavior is also influenced by the nature of the original substrate.

  6. Mineral surface processes responsible for the decreased retardation (or enhanced mobilization) of 137Cs from HLW tank discharges. 1998 annual progress report

    International Nuclear Information System (INIS)

    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.'

  7. Permitting plan for the high-level waste interim storage

    International Nuclear Information System (INIS)

    Deffenbaugh, M.L.

    1997-01-01

    This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-12-07

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

  10. Behavior of uranium and its surrogates in molten aluminosilicate glasses in contact with liquid metals

    International Nuclear Information System (INIS)

    Chevreux, Pierrick

    2016-01-01

    This study concerns an innovative process used for conditioning nuclear waste that contain metallic parts contaminated with actinides. High actinides concentrations are expected to be incorporated in the glass melt in contact with the molten metals. Among these metals, aluminum and/or stainless steel impose a strongly reducing environment to the glass melt involving redox reactions. These reactions modify actinides oxidation states and therefore change their solubilities in the glass and could also reduce them into the metallic form. In this work, we focus on the behavior of uranium and its surrogates, namely hafnium and neodymium, in aluminosilicate glasses from the Na 2 O-CaO-SiO 2 -Al 2 O 3 system melted in highly reducing conditions. The first step consists in comparing the hafnium and uranium solubilities in the glass as functions of redox conditions and glass composition. A methodology has been set up and a specific device has been used to control the oxygen fugacity and the alkali content of the glass. The results show that uranium is far less soluble in the glass than hafnium (Hf(IV)) in reducing conditions. The uranium solubility ranges from 4 to 7 wt% UO 2 for an oxygen fugacity below 10 -14 atm at 1250 C-1400 C. Uranium oxidation states have been investigated by X-ray absorption spectroscopy (XANES). It has been pointed out that U(IV) is the main form in the glass for such imposed oxygen fugacities. The second step of this work is to identify the glass-metal interaction mechanisms in order to determine the localization of uranium and its surrogates (Nd, Hf) in the glass-metal system. Mechanisms are mostly ruled by the presence of metallic aluminum and are similar for uranium, neodymium and hafnium. Glass-metal interaction kinetics demonstrate that uranium and its surrogates can temporarily be reduced into the metallic form for particular conditions. A re-oxidation occurs with time which is in good agreement with thermodynamics. Regarding uranium, the re

  11. Regional Geologic Evaluations for Disposal of HLW and SNF: The Pierre Shale of the Northern Great Plains

    Energy Technology Data Exchange (ETDEWEB)

    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).

  12. The improvement of surface roughness for OAP aluminum mirrors: from terahertz to ultraviolet

    Science.gov (United States)

    Peng, Jilong; Yu, Qian; Shao, Yajun; Wang, Dong; Yi, Zhong; Wang, Shanshan

    2018-01-01

    Aluminum reflector, especially OAP (Off-Axis Parabolic) reflector, has been widely used in terahertz and infrared systems for its low cost, lightweight, good machinability, small size, simple structure, and having the same thermal expansion and contraction with the system structure which makes it have a wide temperature adaptability. Thorlabs, Daheng and other large optical components companies even have Aluminum OAP sold on shelf. Most of the precision Aluminum OAP is fabricated by SPDT (single point diamond turing). Affected by intermittent shock, the roughness of aluminum OAP mirrors through conventional single-point diamond lathes is around 7 nm which limits the scope of application for aluminum mirrors, like in the high power density terahertz/infrared systems and visible/UV optical systems. In this paper, a continuous process frock is proposed, which effectively reduces the influence of turning impact on the mirror roughness. Using this process, an off-axis parabolic aluminum reflector with an effective diameter of 50 mm, off-axis angle of 90 degree is fabricated, and the performances are validated. Measurement by VEECO NT1100 optical profiler with 20× objects, the surface roughness achieves 2.3 nm, and the surface figure error is within λ/7 RMS (λ= 632.8 nm) tested by FISB Aμ Phase laser interferometer with the help of a standard flat mirror. All these technical specifications are close to the traditional glass-based reflectors, and make it possible for using Aluminum reflectors in the higher LIDT (laser induced damage threshold) systems and even for the micro sensor of ionospheric for vacuum ultraviolet micro nano satellites.

  13. Comparing technical concepts for disposal of Belgian vitrified HLW

    International Nuclear Information System (INIS)

    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

  14. A Two-Stage Layered Mixture Experiment Design for a Nuclear Waste Glass Application-Part 1

    International Nuclear Information System (INIS)

    Cooley, Scott K.; Piepel, Gregory F.; Gan, Hao; Kot, Wing; Pegg, Ian L.

    2003-01-01

    A layered experimental design involving mixture variables was generated to support developing property-composition models for high-level waste (HLW) glasses. The design was generated in two stages, each having unique characteristics. Each stage used a layered design having an outer layer, an inner layer, a center point, and some replicates. The layers were defined by single- and multi-variable constraints. The first stage involved 15 glass components treated as mixture variables. For each layer, vertices were generated and optimal design software was used to select alternative subsets of vertices and calculate design optimality measures. Two partial quadratic mixture models, containing 25 terms for the outer layer and 30 terms for the inner layer, were the basis for the optimal design calculations. Distributions of predicted glass property values were plotted and evaluated for the alternative subsets of vertices. Based on the optimality measures and the predicted property distributions, a ''best'' subset of vertices was selected for each layer to form a layered design for the first stage. The design for the second stage was selected to augment the first-stage design. The discussion of the second-stage design begins in this Part 1 and is continued in Part 2 (Cooley and Piepel, 2003b)

  15. Technical Status Report on the Effect of Phosphate and Aluminum on the Development of Amorphous Phase Separation in Sodium

    International Nuclear Information System (INIS)

    Cozzi, A.D.

    1998-01-01

    The objective of the Tank Focus Area ''Optimize Waste Loading'' task is to enhance the definition of the acceptable processing window for high-level waste vitrification plants. One possible manner in which the acceptable processing window may be enhanced is by reducing the uncertainty of various compositional/property models through a specifically defined experimental plan. A reduction in model uncertainty can reduce limitations on current acceptance constraints and may allow for a larger processing or operational window. Enhanced composition/property model predictions coupled with an increased waste loading may decrease the processing time and waste glass disposal costs (i.e., overall lifecycle costs). One of the compositional/property models currently being evaluated by the Tanks Focus Area is related to the development of amorphous phase separation in multi-component borosilicate glasses.Described in this report is the current status for evaluating the effect of phosphorus and alumina on both simple sodium borosilicate and high-level waste glasses on the formation of amorphous phase separation. The goal of this subtask is to increase the understanding of the formation of phase separation by adding significant amounts (3-5 wt. percent) of phosphorus and alumina to well-characterized glasses. Additional scope includes evaluating the effects of thermal history on the formation of amorphous phase separation and durability of select glasses.The development of data, understanding, and quantitative description for composition and kinetic effects on the development of amorphous phase separation will continue in FY99. This effort will provide insight into the compositional and thermal effects on phase stability and will lead to a better understanding of the methods used to predict the development of amorphous phase separation in HLW glasses

  16. Aluminum anode for aluminum-air battery - Part I: Influence of aluminum purity

    Science.gov (United States)

    Cho, Young-Joo; Park, In-Jun; Lee, Hyeok-Jae; Kim, Jung-Gu

    2015-03-01

    2N5 commercial grade aluminum (99.5% purity) leads to the lower aluminum-air battery performances than 4N high pure grade aluminum (99.99% purity) due to impurities itself and formed impurity complex layer which contained Fe, Si, Cu and others. The impurity complex layer of 2N5 grade Al declines the battery voltage on standby status. It also depletes discharge current and battery efficiency at 1.0 V which is general operating voltage of aluminum-air battery. However, the impurity complex layer of 2N5 grade Al is dissolved with decreasing discharge voltage to 0.8 V. This phenomenon leads to improvement of discharge current density and battery efficiency by reducing self-corrosion reaction. This study demonstrates the possibility of use of 2N5 grade Al which is cheaper than 4N grade Al as the anode for aluminum-air battery.

  17. Demonstration of pyrometallurgical processing for metal fuel and HLW

    International Nuclear Information System (INIS)

    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)

  18. Korean Reference HLW Disposal System

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Selection of a glass-ceramic formulation to immobilize fluorinel- sodium calcine

    International Nuclear Information System (INIS)

    Staples, B.A.; Wood, H.C.

    1994-12-01

    One option for immobilizing calcined high level wastes produced by nuclear fuel reprocessing activities at the Idaho Chemical Processing Plant (ICPP) is conversion to a glass-ceramic form through hot isostatic pressing. Calcines exist in several different chemical compositions, and thus candidate formulations have been developed for converting each to glass-ceramic forms which are potentially resistant to aqueous corrosion and stable enough to qualify for repository storage. Fluorinel/Na, a chemically complex calcine type, is one of the types being stored at ICPP, and development efforts have identified three formulations with potential for immobilizing it. These are a glass forming additive that uses aluminum metal to enhance reactivity, a second glass forming additive that uses titanium metal to enhance reactivity and a third that uses not only a combination of silicon and titanium metals but enough phosphorous pentoxide to form a calcium phosphate host phase in the glass-ceramic product. Glass-ceramics of each formulation performed well in restricted characterization tests. However, none of the three was subjected to rigorous testing that would provide information on whether each was processable, that is able to retain favorable characteristics over a practical range of processing conditions

  20. Study on a Preliminary Survey and Analysis of HLW Management Technology Suitable for Nuclear Industrial Environment in Korea

    International Nuclear Information System (INIS)

    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

  1. Thermal phase stability of some simulated Defense waste glasses

    International Nuclear Information System (INIS)

    May, R.P.

    1981-04-01

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

  2. Thermal phase stability of some simulated Defense waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    May, R.P.

    1981-04-01

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

  3. Efficient enhancement of bismuth NIR luminescence by aluminum and its mechanism in bismuth doped germanate laser glass

    DEFF Research Database (Denmark)

    Wang, L.P.; Tan, L.L.; Yue, Yuanzheng

    2016-01-01

    As a new member of laser glass family, bismuth-doped glasses have received rising interests due to the application of fiber amplifiers and laser sources in the new spectral range for the next-generation optical communication system. For practical application of the glasses, it must be considered ...

  4. Effect of the Callovian-Oxfordian clayey fraction on borosilicate glass alteration

    International Nuclear Information System (INIS)

    Debure, M.; Frugier, P.; GIN, S.; De Windt, L.; Michau, N.

    2012-01-01

    Document available in extended abstract form only. In France, high-level nuclear waste (HLW) is confined in a glass matrix packaged into stainless steel canister and carbon steel overpack. The HLW should be buried in a geological clay formation like, potentially, the Callovian-Oxfordian (COx) clay-stone located in the north-eastern Parisian basin. The COx clay-stone contains minerals that can feed the near-field with soluble Mg. Such minerals are carbonates (ankerite, dolomite) as well as clay minerals (chlorite, illite, interstratified illite/smectite). Previous laboratory experiments have proved that aqueous solutions of Mg salts could significantly increase the alteration rate of nuclear glass (Jollivet et al., 2012). This motivated to go a step further by studying the alteration of nuclear glass put in contact with Mg minerals. A first set of experiments have revealed that the rate of glass dissolution was increased with hydro-magnesite (4MgCO 3 .Mg(OH) 2 .4H 2 O, a chemically simple model mineral) and dolomite. In both cases, Mg coming from carbonate dissolution reacts with Si, provided by the glass, in order to form Mg silicates (Debure et al., 2012). In that case, Si consumption sustains glass alteration. Mg silicate precipitation also consumes protons; therefore the interdiffusion of alkali within the glass alteration layer eventually becomes a driving force that sustains Mg silicate precipitation. The second set of experiments, presented here, aimed at better characterizing the role of the COx clayey fraction. The separation of the clayey phases of the COx clay-stone has been made in collaboration with the LEM lab (Nancy, France) by a sequence of sieving, acidic dissolution of carbonates, NaCl washing and sedimentation (Rivard, 2011). According to XRD and infrared analyses, the clayey fraction was mainly composed of kaolinite, illite, interstratified illite/smectite and chlorite (plus a little residual amount of quartz). This first step aimed to remove

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  6. Aluminosilicate melts and glasses at 1 to 3 GPa: Temperature and pressure effects on recovered structural and density changes

    Science.gov (United States)

    Bista, S; Stebbins, Jonathan; Hankins, William B.; Sisson, Thomas W.

    2015-01-01

    In the pressure range in the Earth’s mantle where many basaltic magmas are generated (1 to 3 GPa) (Stolper et al. 1981), increases in the coordination numbers of the network-forming cations in aluminosilicate melts have generally been considered to be minor, although effects on silicon and particularly on aluminum coordination in non-bridging oxygen-rich glasses from the higher, 5 to 12 GPa range, are now well known. Most high-precision measurements of network cation coordination in such samples have been made by spectroscopy (notably 27Al and 29Si NMR) on glasses quenched from high-temperature, high-pressure melts synthesized in solid-media apparatuses and decompressed to room temperature and 1 bar pressure. There are several effects that could lead to the underestimation of the extent of actual structural (and density) changes in high-pressure/temperature melts from such data. For non-bridging oxygen-rich sodium and calcium aluminosilicate compositions in the 1 to 3 GPa range, we show here that glasses annealed near to their glass transition temperatures systematically record higher recovered increases in aluminum coordination and in density than samples quenched from high-temperature melts. In the piston-cylinder apparatus used, rates of cooling through the glass transition are measured as very similar for both higher and lower initial temperatures, indicating that fictive temperature effects are not the likely explanation of these differences. Instead, transient decreases in melt pressure during thermal quenching, which may be especially large for high initial run temperatures, of as much as 0.5 to 1 GPa, may be responsible. As a result, the equilibrium proportion of high-coordinated Al in this pressure range may be 50 to 90% greater than previously estimated, reaching mean coordination numbers (e.g., 4.5) that are probably high enough to significantly affect melt properties. New data on jadeite (NaAlSi2O6) glass confirm that aluminum coordination increase

  7. A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

    Energy Technology Data Exchange (ETDEWEB)

    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

  8. A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

    International Nuclear Information System (INIS)

    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

  9. Calibration of a laboratory spectrophotometer for specular light by means of stacked glass plates.

    Science.gov (United States)

    Allen, W. A.; Richardson, A. J.

    1971-01-01

    Stacked glass plates have been used to calibrate a laboratory spectrophotometer, over the spectral range 0.5-2.5 microns, for specular light. The uncalibrated instrument was characterized by systematic errors when used to measure the reflectance and transmittance of stacked glass plates. Calibration included first, a determination of the reflectance of a standard composed of barium sulfate paint deposited on an aluminum plate; second, the approximation of the reflectance and transmittance residuals between observed and computed values by means of cubic equations; and, finally, the removal of the systematic errors by a computer. The instrument, after calibration, was accurate to 1% when used to measure the reflectance and transmittance of stacked glass plates.

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

    Directory of Open Access Journals (Sweden)

    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.

  11. Effect of irradiation on differential thermal properties and crystallization behavior of some lithium borate glasses

    International Nuclear Information System (INIS)

    El-Alaily, N.A.; Mohamed, R.M.

    2001-01-01

    Differential thermal properties and the crystallization behavior of binary system Li 2 O-B 2 O 3 glasses were investigated. The effects of the presence of oxides of aluminum, lead or one of the transition metals TiO 2 or V 2 O 5 or Fe 2 O 3 in the parent glass were also studied. The effects of three different heat treatments on the crystalline structure of all the studied glasses were also investigated. The results showed that all glass samples were amorphous before the heat treatment, with the most common formed phase being tetraborate Li 2 B 8 O 13 (Li 2 O-4B 2 O 3 ). The exposure of the glass samples to either gamma rays or fast neutrons resulted in considerable changes in their thermal behavior. The results also showed that T g increases for all studied glasses when subjected to irradiation either by fast neutron or gamma rays, while T c decreased only at higher doses

  12. The dual effect of Mg on the long-term alteration rate of AVM nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Thien, Bruno M.J., E-mail: bruno.thien@psi.ch [Commissariat a l' Energie Atomique (CEA), Laboratoire d' Etude du Comportement a Long Terme des Materiaux, DTCD/DEN, Marcoule, 30207 Bagnols sur Ceze (France); Godon, Nicole; Ballestero, Anthony; Gin, Stephane [Commissariat a l' Energie Atomique (CEA), Laboratoire d' Etude du Comportement a Long Terme des Materiaux, DTCD/DEN, Marcoule, 30207 Bagnols sur Ceze (France); Ayral, Andre [Institut Europeen des Membranes, Universite de Montpellier, cc. 047, Place Eugene Bataillon, 34095 Montpellier (France)

    2012-08-15

    Inactive Mg-containing nuclear waste glasses simulating actual HLW glasses produced at the AVM facility since 1995 (Marcoule, France), were leached in aqueous solution in order to assess their long term behaviour. The focus was on the effect of Mg. Our findings show that the distribution of Mg between the gel and the secondary crystalline phases strongly influences the glass dissolution rate. The glasses were leached in initially pure water (T = 50 Degree-Sign C, surface/volume ratio (S/V) = 55 cm{sup -1}) with and without addition of Mg{sup 2+} in the solution. 'Mg-free' AVM glasses were also leached in initially pure water (50 Degree-Sign C, 200 cm{sup -1}) with and without addition of Mg{sup 2+} in the solution. Accurate identification of Mg-smectite secondary phases and gel composition calculations enable us to explain the different observed behaviours. Glass AVM 10 was the less altered glass in pure water. Its gel is more protective than the other probably because it is mainly balanced by Mg{sup 2+}. The addition of Mg{sup 2+} in the solution triggers the precipitation of smectite (not observed in pure water experiments), which consumes silicon from the gel, leading finally to a significant increase of the glass alteration. We also focused on the AVM 6 glass which was the most altered glass in pure water of available AVM glasses. Contrary to AVM 10, the gel of AVM 6 is mainly balanced by Na{sup +}. The addition of Mg{sup 2+} in the solution allows the replacement of Na by Mg within the gel. This reaction clearly improves the gel properties and allows the rate to decrease more rapidly, in spite of the precipitation of smectite (also observed in pure water experiments). Finally, the two glasses were altered in synthetic groundwater (SGW) with a high Mg-Ca content. As expected from the previous observations, AVM 10 was insensitive to the presence of alkaline earths in the leaching solution whereas AVM 6 glass exhibited a lower rate than in pure water

  13. Engineering Glass Passivation Layers -Model Results

    Energy Technology Data Exchange (ETDEWEB)

    Skorski, Daniel C.; Ryan, Joseph V.; Strachan, Denis M.; Lepry, William C.

    2011-08-08

    The immobilization of radioactive waste into glass waste forms is a baseline process of nuclear waste management not only in the United States, but worldwide. The rate of radionuclide release from these glasses is a critical measure of the quality of the waste form. Over long-term tests and using extrapolations of ancient analogues, it has been shown that well designed glasses exhibit a dissolution rate that quickly decreases to a slow residual rate for the lifetime of the glass. The mechanistic cause of this decreased corrosion rate is a subject of debate, with one of the major theories suggesting that the decrease is caused by the formation of corrosion products in such a manner as to present a diffusion barrier on the surface of the glass. Although there is much evidence of this type of mechanism, there has been no attempt to engineer the effect to maximize the passivating qualities of the corrosion products. This study represents the first attempt to engineer the creation of passivating phases on the surface of glasses. Our approach utilizes interactions between the dissolving glass and elements from the disposal environment to create impermeable capping layers. By drawing from other corrosion studies in areas where passivation layers have been successfully engineered to protect the bulk material, we present here a report on mineral phases that are likely have a morphological tendency to encrust the surface of the glass. Our modeling has focused on using the AFCI glass system in a carbonate, sulfate, and phosphate rich environment. We evaluate the minerals predicted to form to determine the likelihood of the formation of a protective layer on the surface of the glass. We have also modeled individual ions in solutions vs. pH and the addition of aluminum and silicon. These results allow us to understand the pH and ion concentration dependence of mineral formation. We have determined that iron minerals are likely to form a complete incrustation layer and we plan

  14. REFINEMENT OF THE NEPHELINE DISCRIMINATOR: RESULTS OF A PHASE I STUDY

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-02-13

    The performance of a glass used for immobilization of high-level nuclear waste (HLW) is generally quantified by its resistance to chemical degradation, or durability. The durability of a HLW glass is dependent on its composition. If crystalline phases form within a glass during cooling, the composition of the residual glass network is altered, therefore affecting the durability of the glass. Crystallization of nepheline (NaAlSiO{sub 4}) has been shown to adversely impact the durability of HLW glasses since it removes glass forming species (in this case, Al and Si) from the glass network. The propensity for nepheline crystallization in a HLW glass increases with increasing concentrations of Al{sub 2}O{sub 3} and Na{sub 2}O in the glass. Nepheline crystallization is therefore of concern for processing of HLW at the Defense Waste Processing Facility (DWPF) since the sludge waste streams at the Savannah River Site (SRS) can contain high concentrations of Al{sub 2}O{sub 3} and Na{sub 2}O. Currently, a 'nepheline discriminator' is included as a process control constraint at the DWPF. The nepheline discriminator relates the concentrations of SiO{sub 2}, Na{sub 2}O and Al{sub 2}O{sub 3} (as weight percentages in glass) to a critical value of 0.62. The discriminator defines a boundary line on the SiO{sub 2}-Na{sub 2}O-Al{sub 2}O{sub 3} ternary diagram above which (or toward the SiO{sub 2} corner of the ternary) nepheline is not predicted to crystallize in the glass upon quenching or slow cooling. The current equation uses only the concentrations of the SiO{sub 2}, Na{sub 2}O and Al{sub 2}O{sub 3} components in the glass in predicting whether or not nepheline is likely to crystallize. However, several other components have been shown to impact the propensity for nepheline crystallization, including B{sub 2}O{sub 3} and CaO among others. Therefore, the potential exists to further refine the nepheline discriminator to include these components. In addition, recently

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

    International Nuclear Information System (INIS)

    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

  16. Leachates analysis of glass from black and white and color televisions sets

    Directory of Open Access Journals (Sweden)

    Radovan Kukla

    2012-01-01

    Full Text Available The aim of work was to determine the content of selected elements in the glass from color and black and white television (TV sets. The amount of back taken TV sets in the Czech Republic increases annualy, which is associated with higher production of the waste glass. Currently there is 1.4 television sets for each household and the number of it should increase in future, because of higher standard of living and new technologies used. Waste glass treatment or landfilling may present, because of composition of the waste glass threat to the environment. One of the indicators of the polution from waste glass is leachate analysis, which can show us the content of hazardous substances in the waste glass, which can be released to the environment. A qualitative analysis of leachate samples was carried out by UV-VIS spectrophotometer. The results showed concentration of potencionaly hazardous substances contained in leachate samples. This was especially content of aluminum, cadmium, chromium, copper, molybdenum, nickel, lead, tin and zinc. Results of analyzes of the aqueous extract of glass were confronted with the limits specified in the currently valid legislation. Based on the results there is clear that in the case of landfilling of the glass from television sets, there is possibility of the contamination of landfill leachate by the elements, which are presented in the glass.

  17. Effect of aluminum addition on the optical, morphology and electrical behavior of spin coated zinc oxide thin films

    Directory of Open Access Journals (Sweden)

    Amit Kumar Srivastava

    2011-09-01

    Full Text Available Aluminum-doped ZnO thin films of high optical transmittance (∼ 84-100% and low resistivity (∼ 2.3x10-2 Ωcm have been prepared on glass substrate by the spin coating and subsequent annealing at 500°C for 1h in air or vacuum. Effect of aluminum doping and annealing environment on morphology, optical transmittance and electrical resistivity of ZnO thin films has been studied with possible application as a transparent electrode in photovoltaic. The changes occurring due to aluminum addition include reduction in grain size, root mean square roughness, peak-valley separation, and sheet resistance with improvement in the optical transmittance to 84-100% in the visible range. The origin of low electrical resistivity lies in increase in i electron concentration following aluminum doping (being trivalent, formation of oxygen vacancies due to vacuum annealing, filling of cation site with additional zinc at solution stage itself and ii carrier mobility.

  18. Influence of clay environment and of corroding canister on the long-term behaviour of nuclear glasses

    International Nuclear Information System (INIS)

    GIn, St.

    1997-01-01

    This article reviews the current state of knowledge of R77 glass alteration mechanisms in the presence of the clay materials under consideration for use in an engineered barrier, and in the presence of metal canister corrosion products. These issues are addressed by a phenomenological approach and by modeling. In most cases, the codes are unable to account for the glass alteration data obtained in clay media. The main reason for this discrepancy lies in the glass alteration kinetics (a first-order law based exclusively on silicon) that do not take into account other elements including aluminum, zirconium or the rare earths, which are strongly mobilized in clay media. (author)

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  20. Packaging material and aluminum. Hoso zairyo to aluminum

    Energy Technology Data Exchange (ETDEWEB)

    Itaya, T [Mitsubishi Aluminum Co. Ltd., Tokyo (Japan)

    1992-02-01

    The present paper introduces aluminum foil packaging materials among the relation between packing materials and aluminum. The characteristics of aluminum foil in the packaging area are in its barrier performance, non-toxicity, tastelessness and odorlessness. Its excellent functions and processibility suit best as functional materials for food, medicine and industrial material packaging. While an aluminum foil may be used as a single packing material as in foils used in homes, many of it as a packaging material are used in combination with adhesives, papers or plastic films, or coated or printed. It is used as composite materials laminated or coated with other materials according to their use for the purpose of complementing the aluminum foil as the base material. Representative method to laminate aluminum foils include the wet lamination, dry lamination, thermally dissolved lamination and extruded lamination. The most important quality requirement in lamination is the adhesion strength, which requires a close attention in selecting the kinds of adhesive, laminating conditions, and aging conditions. 8 figs., 6 tabs.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-23

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

  3. Application of PCT to the EBR II ceramic waste form

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  4. Helium trapping in aluminum and sintered aluminum powders

    International Nuclear Information System (INIS)

    Das, S.K.; Kaminsky, M.; Rossing, T.

    1975-01-01

    The surface erosion of annealed aluminum and of sintered aluminum powder (SAP) due to blistering from implantation of 100-keV 4 He + ions at room temperature has been investigated. A substantial reduction in the blistering erosion rate in SAP was observed from that in pure annealed aluminum. In order to determine whether the observed reduction in blistering is due to enhanced helium trapping or due to helium released, the implanted helium profiles in annealed aluminum and in SAP have been studied by Rutherford backscattering. The results show that more helium is trapped in SAP than in aluminum for identical irradiation conditions. The observed reduction in erosion from helium blistering in SAP is more likely due to the dispersion of trapped helium at the large Al-Al 2 O 3 interfaces and at the large grain boundaries in SAP than to helium release

  5. Finite element modelling of aluminum alloy 2024-T3 under transverse impact loading

    Science.gov (United States)

    Abdullah, Ahmad Sufian; Kuntjoro, Wahyu; Yamin, A. F. M.

    2017-12-01

    Fiber metal laminate named GLARE is a new aerospace material which has great potential to be widely used in future lightweight aircraft. It consists of aluminum alloy 2024-T3 and glass-fiber reinforced laminate. In order to produce reliable finite element model of impact response or crashworthiness of structure made of GLARE, one can initially model and validate the finite element model of the impact response of its constituents separately. The objective of this study was to develop a reliable finite element model of aluminum alloy 2024-T3 under low velocity transverse impact loading using commercial software ABAQUS. Johnson-Cook plasticity and damage models were used to predict the alloy's material properties and impact behavior. The results of the finite element analysis were compared to the experiment that has similar material and impact conditions. Results showed good correlations in terms of impact forces, deformation and failure progressions which concluded that the finite element model of 2024-T3 aluminum alloy under low velocity transverse impact condition using Johnson-Cook plastic and damage models was reliable.

  6. Graphene-aluminum nanocomposites

    International Nuclear Information System (INIS)

    Bartolucci, Stephen F.; Paras, Joseph; Rafiee, Mohammad A.; Rafiee, Javad; Lee, Sabrina; Kapoor, Deepak; Koratkar, Nikhil

    2011-01-01

    Highlights: → We investigated the mechanical properties of aluminum and aluminum nanocomposites. → Graphene composite had lower strength and hardness compared to nanotube reinforcement. → Processing causes aluminum carbide formation at graphene defects. → The carbides in between grains is a source of weakness and lowers tensile strength. - Abstract: Composites of graphene platelets and powdered aluminum were made using ball milling, hot isostatic pressing and extrusion. The mechanical properties and microstructure were studied using hardness and tensile tests, as well as electron microscopy, X-ray diffraction and differential scanning calorimetry. Compared to the pure aluminum and multi-walled carbon nanotube composites, the graphene-aluminum composite showed decreased strength and hardness. This is explained in the context of enhanced aluminum carbide formation with the graphene filler.

  7. Use of molecular dynamics to model the structure of nuclear glasses

    International Nuclear Information System (INIS)

    Delaye, J.M.; Ghaleb, D.

    1997-01-01

    Born-Mayer-Huggins potentials (BMH) are generally used to model oxides glasses. In this article we show how they can be used, if they can be completed by three body terms, to reproduce correctly structures of alumino-borosilicate glasses. Taking into account the parameters adjusted for the basic matrix (SiO 2 , B 2 O 3 , Na 2 O, ZrO 2 , Al 2 O 3 ) of French light water reactor waste containment glass, we have analyzed equally the structure evolutions reproduced by BMH potentials according to the modification of the composition of the glass. The principal objective of this study being the appreciation of the capability of this potentials to simulate structure evolutions in the glass; this potentials representing neither the nature of chemical bonds nor the electronic state modifications that accompany changes of coordination numbers. Globally BHM potentials, despite their limited representativeness, reproduce some experimental observations. For example: - the gradual incorporation of the boron in the silicate network; - the shorter distances between network formers and non-bridging oxygen atoms; - the visualization of boron-enriched segregation zones in a high content boron composition without aluminum. (authors)

  8. Drop Calculations of HLW Canister and Pu Can-in-Canister

    International Nuclear Information System (INIS)

    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

  9. Development of gap filling technique in HLW repository

    International Nuclear Information System (INIS)

    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)

  10. Glass fiber reinforced concrete for terrestrial photovoltaic arrays

    Science.gov (United States)

    Maxwell, H.

    1979-01-01

    The use of glass-fiber-reinforced concrete (GRC) as a low-cost structural substrate for terrestrial solar cell arrays is discussed. The properties and fabrication of glass-reinforced concrete structures are considered, and a preliminary design for a laminated solar cell assembly built on a GRC substrate is presented. A total cost for such a photovoltaic module, composed of a Korad acrylic plastic film front cover, an aluminum foil back cover, an ethylene/vinyl acetate pottant/adhesive and a cotton fabric electrical isolator in addition to the GRC substrate, of $9.42/sq m is projected, which is less than the $11.00/sq m cost goal set by the Department of Energy. Preliminary evaluations are concluded to have shown the design capabilities and cost effectiveness of GRC; however, its potential for automated mass production has yet to be evaluated.

  11. Negative permittivity of ZnO thin films prepared from aluminum and gallium doped ceramics via pulsed-laser deposition

    DEFF Research Database (Denmark)

    Bodea, M. A.; Sbarcea, G.; Naik, G. V.

    2013-01-01

    Aluminum and gallium doped zinc oxide thin films with negative dielectric permittivity in the near infrared spectral range are grown by pulsed laser deposition. Composite ceramics comprising ZnO and secondary phase Al2O3 or Ga2O3 are employed as targets for laser ablation. Films deposited on glass...

  12. Dissolution model for a glass having an adherent insoluble surface layer

    International Nuclear Information System (INIS)

    Harvey, K.B.; Larocque, C.A.B.

    1990-01-01

    Waste form glasses that contain substantial quantities of iron, manganese, and aluminum oxides, such as the Savannah River SRL TDS-131 glass, form a thick, hydrated surface layer when placed in contact with water. The dissolution of such a glass has been modeled with the Savannah River Model. The authors showed previously that the equations of the Savannah River Model could be fitted to published experimental data if a time-dependent diffusion coefficient was assumed for species of diffusing through the surface layer. The Savannah River Model assumes that all of the material dissolved from the glass enters solution, whereas it was observed that substantial quantities of material were retained in the surface layer. An alternative model, presented contains a mass balance equation that allows material either to enter solution or to be retained in the surface layer. It is shown that the equations derived using this model can be fitted to the published experimental data assuming a constant diffusion coefficient for species diffusing through the surface layer

  13. Selective Adsorption of Sodium Aluminum Fluoride Salts from Molten Aluminum

    Energy Technology Data Exchange (ETDEWEB)

    Leonard S. Aubrey; Christine A. Boyle; Eddie M. Williams; David H. DeYoung; Dawid D. Smith; Feng Chi

    2007-08-16

    Aluminum is produced in electrolytic reduction cells where alumina feedstock is dissolved in molten cryolite (sodium aluminum fluoride) along with aluminum and calcium fluorides. The dissolved alumina is then reduced by electrolysis and the molten aluminum separates to the bottom of the cell. The reduction cell is periodically tapped to remove the molten aluminum. During the tapping process, some of the molten electrolyte (commonly referred as “bath” in the aluminum industry) is carried over with the molten aluminum and into the transfer crucible. The carryover of molten bath into the holding furnace can create significant operational problems in aluminum cast houses. Bath carryover can result in several problems. The most troublesome problem is sodium and calcium pickup in magnesium-bearing alloys. Magnesium alloying additions can result in Mg-Na and Mg-Ca exchange reactions with the molten bath, which results in the undesirable pickup of elemental sodium and calcium. This final report presents the findings of a project to evaluate removal of molten bath using a new and novel micro-porous filter media. The theory of selective adsorption or removal is based on interfacial surface energy differences of molten aluminum and bath on the micro-porous filter structure. This report describes the theory of the selective adsorption-filtration process, the development of suitable micro-porous filter media, and the operational results obtained with a micro-porous bed filtration system. The micro-porous filter media was found to very effectively remove molten sodium aluminum fluoride bath by the selective adsorption-filtration mechanism.

  14. Microarc Oxidation of the High-Silicon Aluminum AK12D Alloy

    Directory of Open Access Journals (Sweden)

    S. K. Kiseleva

    2015-01-01

    Full Text Available The aim of work is to study how the high-silicon aluminum AK12D alloy microstructure and MAO-process modes influence on characteristics (microhardness, porosity and thickness of the oxide layer of formed surface layer.Experimental methods of study:1 MAO processing of AK12D alloy disc-shaped samples. MAO modes features are concentration of electrolyte components – soluble water glass Na2SiO3 and potassium hydroxide (KOH. The content of two components both the soluble water glass and the potassium hydroxide was changed at once, with their concentration ratio remaining constant;2 metallographic analysis of AK12D alloy structure using an optical microscope «Olympus GX51»;3 image analysis of the system "alloy AK12D - MAO - layer" using a scanning electron microscope «JEOL JSM 6490LV»;4 hardness evaluation of the MAO-layers using a micro-hardness tester «Struers Duramin».The porosity, microhardness and thickness of MAO-layer formed on samples with different initial structures are analyzed in detail. Attention is paid to the influence of MAO process modes on the quality layer.It has been proved that the MAO processing allows reaching quality coverage with high microhardness values of 1200-1300HV and thickness up to 114 μm on high-silicon aluminum alloy. It has been found that the initial microstructure of alloy greatly affects the thickness of the MAO - layer. The paper explains the observed effect using the physical principles of MAO process and the nature of silicon particles distribution in the billet volume.It has been shown that increasing concentration of sodium silicate and potassium hydroxide in the electrolyte results in thicker coating and high microhardness.It has been revealed that high microhardness is observed in the thicker MAO-layers.Conclusions:1 The microstructure of aluminum AK12D alloy and concentration of electrolyte components - liquid glass Na2SiO3 and potassium hydroxide affect the quality of coating resulted from MAO

  15. Operating experience during high-level waste vitrification at the West Valley Demonstration Project

    International Nuclear Information System (INIS)

    Valenti, P.J.; Elliott, D.I.

    1999-01-01

    This report provides a summary of operational experiences, component and system performance, and lessons learned associated with the operation of the Vitrification Facility (VF) at the West Valley Demonstration Project (WVDP). The VF was designed to convert stored high-level radioactive waste (HLW) into a stable waste form (borosilicate glass) suitable for disposal in a federal repository. Following successful completion on nonradioactive test, HLW processing began in July 1995. Completion of Phase 1 of HLW processing was reached on 10 June 1998 and represented the processing of 9.32 million curies of cesium-137 (Cs-137) and strontium-90 (Sr-90) to fill 211 canisters with over 436,000 kilograms of glass. With approximately 85% of the total estimated curie content removed from underground waste storage tanks during Phase 1, subsequent operations will focus on removal of tank heel wastes

  16. PV glass curtain walls; Kenzai ittaigata taiyo denchi gaiheki no kaihatsu (glass curtain wall eno tekiyo)

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, T.; Iwai, T.; Ouchi, T.; Ito, T.; Nagai, T.; Shu, I. [Kajima Corp., Tokyo (Japan); Arai, T. [Showa Shell Sekiyu K.K., Tokyo (Japan); Ishikawa, N.; Tazawa, K.

    1997-12-20

    Reported in this article are PV (photovoltaic) modules now under development for integration into building walls. First of all, the power generating capability of PV modules and appropriate use of the generated power are studied, and the performance (resistance to fire or incombustibility, strength and durability, appearance and design, and dimensional standardization) that such outer wall materials are to be equipped with are determined. Next, module development, installation technique, computer graphics-aided facade designing, and real size module-using proof test are studied before installability, the power to be generated, and designs are finalized. In the development of modules, design evaluation involves the combining of various kinds of glass, solar cells, back sheets, and fillers, and the importance is confirmed of the prevention of insulation degradation around the modules. As for the methods of installation, the gasket method and aluminum sash method, etc., are tested. In the study of facade design, it is found that various expressions are possible by properly choosing gasket colors, module types, and kinds of glass to cover the openings. 1 ref., 6 figs., 3 tabs.

  17. Public Perspectives in the Japanese HLW Disposal Program

    International Nuclear Information System (INIS)

    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

  18. Public Perspectives in the Japanese HLW Disposal Program

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. Aluminum recovery as a product with high added value using aluminum hazardous waste

    International Nuclear Information System (INIS)

    David, E.; Kopac, J.

    2013-01-01

    Highlights: • Granular and compact aluminum dross were physically and chemically characterized. • A relationship between density, porosity and metal content from dross was established. • Chemical reactions involving aluminum in landfill and negative consequences are shown. • A processing method for aluminum recovering from aluminum dross was developed. • Aluminum was recovered as an value product with high grade purity such as alumina. -- Abstract: The samples of hazardous aluminum solid waste such as dross were physically and chemically characterized. A relationship between density, porosity and metal content of dross was established. The paper also examines the chemical reactions involving aluminum dross in landfill and the negative consequences. To avoid environmental problems and to recovery the aluminum, a processing method was developed and aluminum was recovered as an added value product such as alumina. This method refers to a process at low temperature, in more stages: acid leaching, purification, precipitation and calcination. At the end of this process aluminum was extracted, first as Al 3+ soluble ions and final as alumina product. The composition of the aluminum dross and alumina powder obtained were measured by applying the leaching tests, using atomic absorption spectrometry (AAS) and chemical analysis. The mineralogical composition of aluminum dross samples and alumina product were determined by X-ray diffraction (XRD) and the morphological characterization was performed by scanning electron microscopy (SEM). The method presented in this work allows the use of hazardous aluminum solid waste as raw material to recover an important fraction from soluble aluminum content as an added value product, alumina, with high grade purity (99.28%)

  20. Development and evaluation of candidate high-level waste forms

    International Nuclear Information System (INIS)

    Bernadzikowski, T.A.

    1981-01-01

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

  1. Evaluation of forms for the immobilization of high-level and transuranic wastes

    International Nuclear Information System (INIS)

    Schuman, R.P.; Cox, N.D.; Gibson, G.W.; Kelsey, P.V. Jr.

    1982-08-01

    A figure-of-merit (FOM) analysis has been made of a number of waste forms for solidifying both defense and commercial high-level reprocessing waste (HLW) and transuranic (TRU) wastes. The evaluation includes iron-enriched basalt (IEB), a fusion-produced glass-ceramic, which has not been included in other assessments. For HLW, concrete receives the highest FOM, but may not meet regulatory requirements; IEB and glass are the best choices of the materials that should easily meet regulatory requirements. Concrete waste forms are the best choice for TRU wastes, with IEB a close contender. 116 references, 3 figures, 112 tables

  2. Modelling of radionuclide migration and heat transport from an High-Level-Radioactive-Waste-repository (HLW) in Boom clay

    International Nuclear Information System (INIS)

    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

  3. Use of sugar-cane bagasse ash to produce glass-ceramic material in the system Ca O-SiO2-Na2O

    International Nuclear Information System (INIS)

    Teixeira, S.R.; Santos, G.T.A.; Magalhaes, R.S.; Rincon, J.Ma.; Romero, M.; Carvalho, C.L.

    2009-01-01

    A bottom ash was used as raw material to obtain glass which was crystallized to form glass-ceramic material. The characterization of the ash shows that it consists mainly of crystalline materials, predominantly quartz, with oxides of iron, potassium and aluminum as minor elements. The glass was obtained from the mixing of ash with calcium and sodium carbonates. The glass and the glass-ceramic were examined using differential thermal analysis (DTA), X-ray fluorescence (XRF), X-ray diffraction (XRD) and scanning electron microscopy (SEM). XRD and DTA data show that Wollastonita is the only crystalline phase present in the material crystallized at 1050 deg C. Part of the glass was synthesized at this temperature for one hour, resulting in a green/brown hard material glass-ceramic. The images of SEM show morphology of spherilithic growth indicating volumetric crystallization mechanism. (author)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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)

  7. Commercial waste treatment program annual progress report for FY 1983

    Energy Technology Data Exchange (ETDEWEB)

    McElroy, J.L.; Burkholder, H.C. (comps.)

    1984-02-01

    This annual report describes progress during FY 1983 relating to technologies under development by the Commercial Waste Treatment Program, including: development of glass waste form and vitrification equipment for high-level wastes (HLW); waste form development and process selection for transuranic (TRU) wastes; pilot-scale operation of a radioactive liquid-fed ceramic melter (LFCM) system for verifying the reliability of the reference HLW treatment proces technology; evaluation of treatment requirements for spent fuel as a waste form; second-generation waste form development for HLW; and vitrification process control and product quality assurance technologies.

  8. Experimental investigation of the effects of aqueous species on the dissolution kinetics of R7T7 glass; Etude experimentale de l`influence d`especes aqueuses sur la cinetique de dissolution du verre nucleaire R7T7

    Energy Technology Data Exchange (ETDEWEB)

    Gin, S

    1994-10-01

    This contribution to the study of aqueous corrosion of the French ``R7T7`` reference nuclear containment glass includes a bibliographic survey of prior investigations, highlighting the problems encountered in interpreting the interactions in systems containing clay materials in contact with the glass. An experimental methodology is proposed to investigate the effects of inorganic aqueous species separately from those of a few organic acids on the dissolution mechanisms and kinetics of R7T7 glass at 90 deg. C. The experimental results discussed support the idea that several glass network forming elements may have a kinetically limiting role. The most likely hypothesis to account for the absence of saturation conditions with respect to the glass in certain clay media involves the formation of complexes with kinetically limiting metallic elements such as aluminum released by glass corrosion. This work contributes to a better understanding of the basic mechanisms of nuclear glass dissolution in a geological repository environment. It facilitates the interpretation of glass alteration studies in realistic or actual solutions and may contribute to specifying near field chemical barriers in the form of additives (amorphous silica, aluminum hydroxides or phosphates) around the glass disposal package to enhance the stability of the glass matrix. (author). 148 refs., 40 figs., 32 tabs., 1 append.

  9. Safety studies of HLW-disposal in the Mors salt dome - Support to the salt option of the Pagis project

    International Nuclear Information System (INIS)

    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

  10. Application of Cu-polyimide flex circuit and Al-on-glass pitch adapter for the ATLAS SCT barrel hybrid

    CERN Document Server

    Unno, Y; Ikegami, Y; Iwata, Y; Kohriki, T; Kondo, T; Nakano, I; Ohsugi, T; Takashima, R; Tanaka, R; Terada, S; Ujiie, N

    2005-01-01

    We applied the surface build-up Cu-polyimide flex-circuit technology with laser vias to the ATLAS SCT barrel hybrid to be made in one piece from the connector to the electronics sections including cables. The hybrids, reinforced with carbon-carbon substrates, provide mechanical strength, thermal conductivity, low-radiation length, and stability in application-specific integrated circuit (ASIC) operation. By following the design rules, we experienced little trouble in breaking the traces. The pitch adapter between the sensor and the ASICs was made of aluminum traces on glass substrate. We identified that the generation of whiskers around the wire-bonding feet was correlated with the hardness of metallized aluminum. The appropriate hardness has been achieved by keeping the temperature of the glasses as low as room temperature during the metallization. The argon plasma cleaning procedure cleaned the contamination on the gold pads of the hybrids for successful wire bonding, although it was unsuccessful in the alu...

  11. Tailored white light emission in Eu3+/Dy3+ doped tellurite glass phosphors containing Al3+ ions

    Science.gov (United States)

    Walas, Michalina; Piotrowski, Patryk; Lewandowski, Tomasz; Synak, Anna; Łapiński, Marcin; Sadowski, Wojciech; Kościelska, Barbara

    2018-05-01

    Tellurite glass systems modified by addition of aluminum fluoride AlF3 have been successfully synthesized as host matrices for optically active rare earth ions RE3+ (RE3+ = Eu3+, Dy3+). Samples with different Eu3+ to Dy3+ molar ratio have been studied in order to determine possibility of white light emission via UV excitation. Structural investigations confirmed amorphous character of materials whereas spectroscopic studies brought more insight into glass network's nature. FTIR results shown presence of two features related to tellurite glass matrix (in 490-935 cm-1 spectral region) and another one (940-1250 cm-1) due to aluminum addition. Especially, Al-O and Te-O-Al bonds of AlO4 tetrahedrons have been found. AlO4 units are considered as glass formers that improve network's strength and thermal resistivity against devitrification. Based on XPS studies of Al3+ photoelectron band the existence of Al-O and also Al-F bonds have been examined. Moreover, signals originating from Eu3+ and Dy3+ have been found confirming their valence state. Luminescence results revealed possibility of simultaneous UV excitation of Eu3+ and Dy3+ ions. Excitation with λexc = 390 and 393 nm resulted in white light generation starting from warm white to neutral and cool white depending on Eu3+ concentration and used excitation wavelength. Additionally, increase of decay lifetime of Eu3+ induced by Al3+ presence have been revealed based on luminescence decay analysis. Thus, tellurite glass systems modified by AlF3 and doped with Eu3+/Dy3+ may be considered as promising candidates for white light emitting sources.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  15. Glass ceramics for sealing to high-thermal-expansion metals

    International Nuclear Information System (INIS)

    Wilder, J.A. Jr.

    1980-10-01

    Glass ceramics were studied, formulated in the Na 2 O CaO.P 2 O 5 , Na 2 O.BaOP 2 O 5 , Na 2 O.Al 2 O 3 .P 2 O 5 , and Li 2 O.BaO.P 2 O 5 systems to establish their suitability for sealing to high thermal expansion metals, e.g. aluminum, copper, and 300 series stainless steels. Glass ceramics in Na 2 O.CaO.P 2 O 5 and Na 2 O.BaO.P 2 O 5 systems have coefficients of thermal expansion in the range 140 x 10 -1 per 0 C less than or equal to α less than or equal to 225 x 10 -7 per 0 C and fracture toughness values generally greater than those of phosphate glasses; they are suitable for fabricating seals to high thermal expansion metals. Crystal phases include NaPo 3 , (NaPO 3 ) 3 , NaBa(PO 3 ) 3 , and NaCa(PO 3 ) 3 . Glass ceramics formed in the Na 2 O.Al 2 O 3 .P 2 O 5 systems have coefficients of thermal expansion greater than 240 x 10 -7 per 0 C, but they have extensive microcracking. Due to their low thermal expansion values (α less than or equal to 120 x 10 -7 per 0 C), glass ceramics in the Li 2 O.BaO.P 2 O 5 system are unsuitable for sealing to high thermal expansion metals

  16. Development of high-level waste solidification technology 1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Joon Hyung; Kim, Hwan Young; Kim, In Tae [and others

    1999-02-01

    Spent nuclear fuel contains useful nuclides as valuable resource materials for energy, heat and catalyst. High-level wastes (HLW) are expected to be generated from the R and D activities and reuse processes. It is necessary to develop vitrification or advanced solidification technologies for the safe long-term management of high level wastes. As a first step to establish HLW vitrification technology, characterization of HLWs that would arise at KAERI site, glass melting experiments with a lab-scale high frequency induction melter, and fabrication and property evaluation of base-glass made of used HEPA filter media and additives were performed. Basic study on the fabrication and characterization of candidate ceramic waste form (Synroc) was also carried out. These HLW solidification technologies would be directly useful for carrying out the R and Ds on the nuclear fuel cycle and waste management. (author). 70 refs., 29 tabs., 35 figs.

  17. TIME-TEMPERATURE-TRANSFORMATION DIAGRAMS FOR THE SLUDGE BATCH 3 - FRIT 418 GLASS SYSTEM

    International Nuclear Information System (INIS)

    Billings, A.; Edwards, Tommy

    2009-01-01

    As a part of the Waste Acceptance Product Specifications (WAPS) for Vitrified High-Level Waste Forms defined by the Department of Energy - Office of Environmental Management, the phase stability must be determined for each of the projected high-level waste (HLW) types at the Savannah River Site (SRS). Specifically, WAPS 1.4.1 requires the glass transition temperature (Tg) to be defined and time-temperature-transformation (TTT) diagrams to be developed. The Tg of a glass is an indicator of the approximate temperature where the supercooled liquid converts to a solid on cooling or conversely, where the solid begins to behave as a viscoelastic solid on heating. A TTT diagram identifies the crystalline phases that can form as a function of time and temperature for a given waste type or more specifically, the borosilicate glass waste form. In order to assess durability, the Product Consistency Test (PCT) was used and the durability results compared to the Environmental Assessment (EA) glass. The measurement of glass transition temperature and the development of TTT diagrams have already been performed for the seven Defense Waste Processing Facility (DWPF) projected compositions as defined in the Waste Form Compliance Plan (WCP). These measurements were performed before DWPF start-up and the results were incorporated in Volume 7 of the Waste Form Qualification Report (WQR). Additional information exists for other projected compositions, but overall these compositions did not consider some of the processing scenarios now envisioned for DWPF to accelerate throughput. Changes in DWPF processing strategy have required this WAPS specification to be revisited to ensure that the resulting phases have been bounded. Frit 418 was primarily used to process HLW Sludge Batch 3 (SB3) at 38% waste loading (WL) through the DWPF. The Savannah River National Laboratory (SRNL) fabricated a cache of glass from reagent grade oxides to simulate the SB3-Frit 418 system at a 38 wt % WL for glass

  18. Aluminum powder metallurgy processing

    Energy Technology Data Exchange (ETDEWEB)

    Flumerfelt, J.F.

    1999-02-12

    The objective of this dissertation is to explore the hypothesis that there is a strong linkage between gas atomization processing conditions, as-atomized aluminum powder characteristics, and the consolidation methodology required to make components from aluminum powder. The hypothesis was tested with pure aluminum powders produced by commercial air atomization, commercial inert gas atomization, and gas atomization reaction synthesis (GARS). A comparison of the GARS aluminum powders with the commercial aluminum powders showed the former to exhibit superior powder characteristics. The powders were compared in terms of size and shape, bulk chemistry, surface oxide chemistry and structure, and oxide film thickness. Minimum explosive concentration measurements assessed the dependence of explosibility hazard on surface area, oxide film thickness, and gas atomization processing conditions. The GARS aluminum powders were exposed to different relative humidity levels, demonstrating the effect of atmospheric conditions on post-atomization processing conditions. The GARS aluminum powders were exposed to different relative humidity levels, demonstrating the effect of atmospheric conditions on post-atomization oxidation of aluminum powder. An Al-Ti-Y GARS alloy exposed in ambient air at different temperatures revealed the effect of reactive alloy elements on post-atomization powder oxidation. The pure aluminum powders were consolidated by two different routes, a conventional consolidation process for fabricating aerospace components with aluminum powder and a proposed alternative. The consolidation procedures were compared by evaluating the consolidated microstructures and the corresponding mechanical properties. A low temperature solid state sintering experiment demonstrated that tap densified GARS aluminum powders can form sintering necks between contacting powder particles, unlike the total resistance to sintering of commercial air atomization aluminum powder.

  19. Characterization of Simulant LAW Envelope A, B, and C with Glass Formers

    International Nuclear Information System (INIS)

    Hansen, E.K.

    2000-01-01

    The River Protection Project-Waste Treatment Plant (RPP-WPT) pretreatment and immobilization processes being developed by the DOE Office of River Protection will decontaminate High Level Waste (HLW) Envelopes A and B supernates using crossflow filtration followed by cesium and technetium ion exchange. Envelope C will undergo Sr/TRU precipitation prior to filtration to remove chelated actinides. The decontaminated supernates, now called low activity waste (LAW), will be concentrated through the LAW Melter Feed Evaporator. The concentrated LAW Melter Feed will be mixed with glass forming minerals and chemicals in an in the LAW Melter Feed Preparation Tank. The resulting slurry is then transferred to a Melter Feed Tank from which it is fed to one of the joule-heated, refractory-lined melters. Characterization of the melter feed slurry is required to complete the design of the RPP-WPT slurry feed systems. This report discusses the results obtained from the task, ''Bench Scale Mixing - Characterization of Simulant LAW Envelope A (AN105), B (AZ101), and C (AN107) With Glass Formers''. This task characterized the physical and chemical properties (rheology, particle size, weight percent soluble and insoluble solids, and chemical composition) of simulated LAW Melter feeds made from the different envelopes mentioned above. The goal of this task was to provide data for the design of the RPP-WPT Melter feed system

  20. The Oxidation Products of Aluminum Hydride and Boron Aluminum Hydride Clusters

    Science.gov (United States)

    2016-01-04

    AFRL-AFOSR-VA-TR-2016-0075 The Oxidation Products of Aluminum Hydride and Boron Aluminum Hydride Clusters KIT BOWEN JOHNS HOPKINS UNIV BALTIMORE MD...2. REPORT TYPE Final Performance 3. DATES COVERED (From - To) 30-09-2014 to 29-09-2015 4. TITLE AND SUBTITLE The Oxidation Products of Aluminum ...Hydride and Boron Aluminum Hydride Clusters 5a.  CONTRACT NUMBER 5b.  GRANT NUMBER FA9550-14-1-0324 5c.  PROGRAM ELEMENT NUMBER 61102F 6. AUTHOR(S) KIT

  1. Aluminum Hydroxide

    Science.gov (United States)

    Aluminum hydroxide is used for the relief of heartburn, sour stomach, and peptic ulcer pain and to ... Aluminum hydroxide comes as a capsule, a tablet, and an oral liquid and suspension. The dose and ...

  2. Investigation of alumino-phosphate glasses for iodine conditioning

    International Nuclear Information System (INIS)

    Lemesle, T.

    2013-01-01

    Iodine 129 is a long-lived intermediate level radioactive waste, which is currently managed by isotopic dilution. In view of an alternative management by geological disposal, we aimed at developing phosphate glasses of the AgI-Ag 2 O-P 2 O 5 -Al 2 O 3 system, elaborated at low temperature and without iodine volatilization. Alumina is expected to induce crosslinking of the phosphate network and thus to improve the thermal and chemical properties. To define a glass composition that meets the specifications, we varied the level of iodine, the Ag 2 O/P 2 O 5 ratio and alumina content. For 1 g.cm -3 of iodine, SEM-EDS observations indicate that alumina solubility is limited to 0.5% mol., independently of Ag 2 O/P 2 O 5 ratio. The structural study by 31 P, 27 Al and 109 Ag MAS NMR, shows that aluminum adopts an octahedral coordination that effectively contributes to the crosslinking of the glassy network and iodine is incorporated without clustering. 31 P- 27 Al NMR correlations confirmed the presence of an alumino-phosphate network, and 31 P- 31 P correlations indicate that iodine does not change the connectivity of the glass network. The glass composition 28,8AgI-44,2Ag 2 O-26,5P 2 O 5 -0,5Al 2 O 3 presents the best compromise between the level of incorporation of iodine and the chemical durability, has a glass transition temperature of 123 C and an initial alteration rate in pure water at 50 C of 6 g.m -2 .d -1 . The long-term behavior of this glass is controlled by a post-alteration structure based on pyrophosphate, which holds nearly 80% of the initial iodine. (author) [fr

  3. Industrial high-rate (~14 nm/s) deposition of low resistive and transparent ZnOx:Al films on glass

    NARCIS (Netherlands)

    Illiberi, A.; Kniknie, B.; Deelen, J. van; Steijvers, H.L.A.H.; Habets, D.; Simons, P.J.P.M.; Janssen, A.C.; Beckers, E.H.A.

    2011-01-01

    Aluminum doped ZnOx (ZnOx:Al) films have been deposited on glass in an in-line industrial-type reactor by a metalorganic chemical vapor deposition process at atmospheric pressure. Tertiary-butanol has been used as oxidant for diethylzinc and trimethylaluminium as dopant gas. ZnOx:Al films can be

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

    International Nuclear Information System (INIS)

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

    1988-09-01

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

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

    International Nuclear Information System (INIS)

    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

  6. Is the Aluminum Hypothesis Dead?

    Science.gov (United States)

    2014-01-01

    The Aluminum Hypothesis, the idea that aluminum exposure is involved in the etiology of Alzheimer disease, dates back to a 1965 demonstration that aluminum causes neurofibrillary tangles in the brains of rabbits. Initially the focus of intensive research, the Aluminum Hypothesis has gradually been abandoned by most researchers. Yet, despite this current indifference, the Aluminum Hypothesis continues to attract the attention of a small group of scientists and aluminum continues to be viewed with concern by some of the public. This review article discusses reasons that mainstream science has largely abandoned the Aluminum Hypothesis and explores a possible reason for some in the general public continuing to view aluminum with mistrust. PMID:24806729

  7. Grouping of HLW in partitioning for B/T (burning and/or transmutation) treatment with neutron reactors based on three criteria

    International Nuclear Information System (INIS)

    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)

  8. 75 FR 70689 - Kaiser Aluminum Fabricated Products, LLC; Kaiser Aluminum-Greenwood Forge Division; Currently...

    Science.gov (United States)

    2010-11-18

    ... DEPARTMENT OF LABOR Employment and Training Administration [TA-W-70,376] Kaiser Aluminum Fabricated Products, LLC; Kaiser Aluminum- Greenwood Forge Division; Currently Known As Contech Forgings, LLC..., applicable to workers of Kaiser Aluminum Fabricated Products, LLC, Kaiser Aluminum-Greenwood Forge Division...

  9. The effect of zinc on the aluminum anode of the aluminum-air battery

    Science.gov (United States)

    Tang, Yougen; Lu, Lingbin; Roesky, Herbert W.; Wang, Laiwen; Huang, Baiyun

    Aluminum is an ideal material for batteries, due to its excellent electrochemical performance. Herein, the effect of zinc on the aluminum anode of the aluminum-air battery, as an additive for aluminum alloy and electrolytes, has been studied. The results show that zinc can decrease the anodic polarization, restrain the hydrogen evolution and increase the anodic utilization rate.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  11. Microstructural and Optical Properties of Porous Alumina Elaborated on Glass Substrate

    Science.gov (United States)

    Zaghdoudi, W.; Gaidi, M.; Chtourou, R.

    2013-03-01

    A transparent porous anodized aluminum oxide (AAO) nanostructure was formed on a glass substrate using the anodization of a highly pure evaporated aluminum layer. A parametric study was carried out in order to achieve a fine control of the microstructural and optical properties of the elaborated films. The microstructural and surface morphologies of the porous alumina films were characterized by x-ray diffraction and atomic force microscopy. Pore diameter, inter-pore separation, and the porous structure as a function of anodization conditions were investigated. It was then found that the pores density decreases with increasing the anodization time. Regular cylindrica