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Sample records for aluminum hlw glasses

  1. HIGH ALUMINUM HLW GLASSES FOR HANFORD'S WTP

    International Nuclear Information System (INIS)

    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 Al2O3 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

    International Nuclear Information System (INIS)

    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

    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

  4. 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

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

    International Nuclear Information System (INIS)

    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 m2 and the DuraMelter 1200 (DMI200) HLW Pilot Melter with a melt surface area of 1.2 m2. The newly developed glass formulations had waste loadings as high as 50 wt%, with corresponding Al2O3 concentration in the glass of 26.63 wt%. The new glass formulations showed glass production rates as high as 1900 kg/(m2.day) under nominal melter operating conditions. The demonstrated glass production rates are much higher than the current requirement of 800 kg/(m2.day) and anticipated future enhanced Hanford Tank Waste Treatment and Immobilization Plant (WTP) requirement of 1000 kg/(m2.day).

  6. HIGH ALUMINUM HLW (HIGH LEVEL WASTE ) GLASSES FOR HANFORDS WTP (WASTE TREATMENT PROJECT)

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; BOWAN BW; JOSEPH I; GAN H; KOT WK; MATLACK KS; PEGG IL

    2010-01-04

    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{sup 2} and the DuraMelter 1200 (DMI200) HLW Pilot Melter with a melt surface area of 1.2 m{sup 2}. The newly developed glass formulations had waste loadings as high as 50 wt%, with corresponding Al{sub 2}O{sub 3} concentration in the glass of 26.63 wt%. The new glass formulations showed glass production rates as high as 1900 kg/(m{sup 2}.day) under nominal melter operating conditions. The demonstrated glass production rates are much higher than the current requirement of 800 kg/(m{sup 2}.day) and anticipated future enhanced Hanford Tank Waste Treatment and Immobilization Plant (WTP) requirement of 1000 kg/(m{sup 2}.day).

  7. 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

  8. Defense HLW Glass Degradation Model

    International Nuclear Information System (INIS)

    The purpose of this report is to document the development of a model for calculating the release rate for radionuclides and other key elements from high-level radioactive waste (HLW) glasses under exposure conditions relevant to the performance of the repository. Several glass compositions are planned for the repository, some of which have yet to be identified (i.e., glasses from Hanford and Idaho National Engineering and Environmental Laboratory). The mechanism for glass dissolution is the same for these glasses and the glasses yet to be developed for the disposal of DOE wastes. All of these glasses will be of a quality consistent with the glasses used to develop this report

  9. Defense HLW Glass Degradation Model

    Energy Technology Data Exchange (ETDEWEB)

    D. Strachan

    2004-10-20

    The purpose of this report is to document the development of a model for calculating the release rate for radionuclides and other key elements from high-level radioactive waste (HLW) glasses under exposure conditions relevant to the performance of the repository. Several glass compositions are planned for the repository, some of which have yet to be identified (i.e., glasses from Hanford and Idaho National Engineering and Environmental Laboratory). The mechanism for glass dissolution is the same for these glasses and the glasses yet to be developed for the disposal of DOE wastes. All of these glasses will be of a quality consistent with the glasses used to develop this report.

  10. 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.

  11. INCONEL 690 CORROSION IN WTP (WASTE TREATMENT PLANT) HLW (HIGH LEVEL WASTE) GLASS MELTS RICH IN ALUMINUM & BISMUTH & CHROMIUM OR ALUMINUM/SODIUM

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; FENG Z; GAN H; PEGG IL

    2009-11-05

    Metal corrosion tests were conducted with four high waste loading non-Fe-limited HLW glass compositions. The results at 1150 C (the WTP nominal melter operating temperature) show corrosion performance for all four glasses that is comparable to that of other typical borosilicate waste glasses, including HLW glass compositions that have been developed for iron-limited WTP streams. Of the four glasses tested, the Bi-limited composition shows the greatest extent of corrosion, which may be related to its higher phosphorus content. Tests at higher suggest that a moderate elevation of the melter operating temperature (up to 1200 C) should not result in any significant increase in Inconel corrosion. However, corrosion rates did increase significantly at yet higher temperatures (1230 C). Very little difference was observed with and without the presence of an electric current density of 6 A/inch{sup 2}, which is the typical upper design limit for Inconel electrodes. The data show a roughly linear relationship between the thickness of the oxide scale on the coupon and the Cr-depletion depth, which is consistent with the chromium depletion providing the material source for scale growth. Analysis of the time dependence of the Cr depletion profiles measured at 1200 C suggests that diffusion of Cr in the Ni-based Inconel alloy controls the depletion depth of Cr inside the alloy. The diffusion coefficient derived from the experimental data agrees within one order of magnitude with the published diffusion coefficient data for Cr in Ni matrices; the difference is likely due to the contribution from faster grain boundary diffusion in the tested Inconel alloy. A simple diffusion model based on these data predicts that Inconel 690 alloy will suffer Cr depletion damage to a depth of about 1 cm over a five year service life at 1200 C in these glasses.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-24

    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.

  14. 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.

  15. Empirical Model For Formulation Of Crystal-Tolerant HLW Glasses

    International Nuclear Information System (INIS)

    Historically, high-level waste (HLW) glasses have been formulated with a low liquideus temperature (TL), or temperature at which the equilibrium fraction of spinel crystals in the melt is below 1 vol % (T0.01), nominally below 1050 C. These constraints cannot prevent the accumulation of large spinel crystals in considerably cooler regions (∼ 850 C) of the glass discharge riser during melter idling and significantly limit the waste loading, which is reflected in a high volume of waste glass, and would result in high capital, production, and disposal costs. A developed empirical model predicts crystal accumulation in the riser of the melter as a function of concentration of spinel-forming components in glass, and thereby provides guidance in formulating crystal-tolerant glasses that would allow high waste loadings by keeping the spinel crystals small and therefore suspended in the glass.

  16. EMPIRICAL MODEL FOR FORMULATION OF CRYSTAL-TOLERANT HLW GLASSES

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATYAS J; HUCKLEBERRY AR; VIENNA JD; RODRIGUEZ CA

    2012-03-07

    Historically, high-level waste (HLW) glasses have been formulated with a low liquideus temperature (T{sub L}), or temperature at which the equilibrium fraction of spinel crystals in the melt is below 1 vol % (T{sub 0.01}), nominally below 1050 C. These constraints cannot prevent the accumulation of large spinel crystals in considerably cooler regions ({approx} 850 C) of the glass discharge riser during melter idling and significantly limit the waste loading, which is reflected in a high volume of waste glass, and would result in high capital, production, and disposal costs. A developed empirical model predicts crystal accumulation in the riser of the melter as a function of concentration of spinel-forming components in glass, and thereby provides guidance in formulating crystal-tolerant glasses that would allow high waste loadings by keeping the spinel crystals small and therefore suspended in the glass.

  17. A comparison of HLW-glass and PWR-borate waste glass

    Science.gov (United States)

    Luo, Shanggeng; Sheng, Jiawei; Tang, Baolong

    2001-09-01

    Glass can incorporate a wide variety of wastes ranging from high level wastes (HLW) to low and intermediate level wastes (LILW). A comparison of HLW-Glass and PWR-borate waste glass is given in this paper. The HLW glass formulation named GC-12/9B and 90-19/U can incorporate 16-20 wt% HLW at 1100°C or 1150°C. The borate waste glass named SL-1 can incorporate 45 wt% borate waste generated from PWR. Their physical properties, characteristic temperatures, chemical durability and leach behavior are summarized here. The comparison indicates: the PWR-glass SL-1 can incorporate up to 45 wt% waste oxides at lower melting temperature (1000°C) in agreement with minimum additive waste stabilization (MAWS) approach; owing to the PWR-borate glass contain less Si and more B and Na, its mass loss is higher than HWR-glass; both HLW-glass and PWR-borate glass have favorable chemical durability and the same leaching phenomena, i.e., Na is mostly depleted, but Ca, Mg, Al and Ti are enriched in the leached surface layer.

  18. A comparison of HLW-glass and PWR-borate waste glass

    International Nuclear Information System (INIS)

    Glass can incorporate a wide variety of wastes ranging from high level wastes (HLW) to low and intermediate level wastes (LILW). A comparison of HLW-Glass and PWR-borate waste glass is given in this paper. The HLW glass formulation named GC-12/9B and 90-19/U can incorporate 16-20 wt% HLW at 1100 deg. C or 1150 deg. C. The borate waste glass named SL-1 can incorporate 45 wt% borate waste generated from PWR. Their physical properties, characteristic temperatures, chemical durability and leach behavior are summarized here. The comparison indicates: the PWR-glass SL-1 can incorporate up to 45 wt% waste oxides at lower melting temperature (1000 deg. C) in agreement with minimum additive waste stabilization (MAWS) approach; owing to the PWR-borate glass contain less Si and more B and Na, its mass loss is higher than HWR-glass; both HLW-glass and PWR-borate glass have favorable chemical durability and the same leaching phenomena, i.e., Na is mostly depleted, but Ca, Mg, Al and Ti are enriched in the leached surface layer

  19. Fundamental study on HLW glass corrosion and mineralization

    International Nuclear Information System (INIS)

    A large number of studies on aqueous corrosion of HLW glass have shown that the glass react with water to form more stable solid phases (alteration-phases or secondary phases). The process of alteration-phase formation is expected to play an important role in the radionuclide release from the glass, because it can affect both the glass dissolution rate and the retention of radio nuclides in the phases. Recent studies have indicated that analcime (zeolite) forms during aqueous corrosion of the glass in certain conditions, and the analcime formation can accelerate the glass corrosion by consuming orthosilicic acid (H4SiO4) from the solution. On the other hand, the alteration-phases such as zeolite and smectite are expected to have a retention capacity for some radio nuclides by sorption or incorporation. Therefore, a sound understanding of the alteration phase formation is expected to be essential for validation of the long-term performance. The purpose of this study is to understand, qualitatively and quantitatively, the alteration-phase formation and associated elemental release during aqueous corrosion of HLW glass. Static corrosion tests were performed with a simulated HLW glass, P0798 glass, in NaOH solutions at elevated temperatures, in order to accelerate the reaction, as a function of temperature, time and NaOH concentration. Crystalline alteration-phases formed in the corroded glass were analyzed by use of XRD, and the solution concentrations of dissolved elements were measured by use of ICP-MS. The results indicated that; 1) Analcime or Na-beidellite or both of them form during the corrosion depending on the conditions, 2) Si rich amorphous phases are contained in the alteration-phases, 3) In addition to solution pH, solution concentrations of Na and K sensitively affect formation of analcime and Na-beidellite, 4) Analcime formation accelerates the glass corrosion, 5) Most of Cs in the glass is retained in the alteration-phases by sorption onto Na

  20. Chemical compatibility of HLW borosilicate glasses with actinides

    International Nuclear Information System (INIS)

    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 230C and 1150C. (orig.)

  1. NMR investigation of cation distribution in HLW wasteform glass

    International Nuclear Information System (INIS)

    Magic-angle-spinning NMR has been used to establish the structural roles of various cations added to the borosilicate glass which is used for the vitrification of high-level nuclear waste (HLW). Representative surrogate oxides with nominal valencies of +1, +2 and +3 have been studied which span the range of oxides from modifier to intermediate and conditional glass former. NMR has been carried out on those nuclei which are accessible and the species observed have been correlated with the physical and chemical behaviour. The controlling factor is the manner in which the alkali cations partition between the various network groups, changing the distribution of silicon Qn species and the boron N4 ratio. Identifiable super-structural units are also present in these glasses. The aqueous corrosion rate increases with Q3 content, as does the weight loss due to evaporation from the melt. The activation energy for DC conduction scales with N4. Values of N4 obtained for these glasses deviate significantly from those predicted by the currently accepted model (Dell and Bray) and are strongly affected by the modifier or intermediate nature of the surrogate oxide and also by its effect on the distribution of non bridging oxygens between the silicate and borate polyhedra. (authors)

  2. 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)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.

    2013-01-16

    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

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

    International Nuclear Information System (INIS)

    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

  5. Corrosion Behavior of Simulated HLW Glass in the Presence of Magnesium Ion

    Directory of Open Access Journals (Sweden)

    Toshikatsu Maeda

    2011-01-01

    Full Text Available Static leach tests were conducted for simulated HLW glass in MgCl2 solution for up to 92 days to investigate the dissolution mechanism of HLW glass under coastal repository condition. Under the condition that magnesium ion exists in leachate, the dissolution rate of the glass did not decrease with time during leaching, while the rate decreased when the magnesium ion depleted in the leachate. In addition, altered layer including magnesium and silica was observed at the surface of the glass after the leach tests. The present results imply that dissolution of the glass is accompanied with formation of magnesium silicate consuming silica, a glass network former. As a consequence, the glass dissolved with an initial high dissolution rate.

  6. ORIGEN-S (α,n) neutron source spectra in borosilicate glass containing HLW

    International Nuclear Information System (INIS)

    There is growing interest in the methodology and computational software for evaluating the (α,n) source spectra produced in mixtures of high-level waste (HLW) and borosilicate glass. The need for this development has been seen in previous work involving the analysis of HLW in borosilicate glass. Descriptions and applications of the ORIGEN-S method of computing neutron source spectra by both (α,n) reactions and spontaneous fission of UO2 spent fuel have been reported previously. This summary presents a significant expansion of the ORIGEN-S (α,n) model to include alpha interactions with the light elements of borosilicate glass. The Battelle/Office of Nuclear Waste Isolation requested this model extension. There is an associated interest in the use of Oak Ridge National Lab. shielding codes for analyzing HLW systems

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

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

    Mougnaud, S.; Tribet, M.; Rolland, S.; Renault, J.-P.; Jégou, C.

    2015-07-01

    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.

  9. On the chemical stability of HLW glasses under irradiation

    International Nuclear Information System (INIS)

    Two articles published recently in Science on tie behavior of glasses under irradiation by heavy ions have received a degree of publicity not normally accorded specialized technical reports. This is due to some speculative statements made by the authors of those articles indicating that irradiation greatly enhanced the leaching rate of the glasses, which effect had not been observed in other studies because they had failed to take into account the influence of recoil atoms on this mechanism. The political impact of treating such topics in the daily press makes it imperative that all the facts be correct. For this reason, the statement contained in the two articles are analyzed in detail in this contribution, in which the authors show that the phenomena described in the papers under review are in no way new. In fact highly atypical, very simple and partly unknown types of glass had been used and examined by methods making simple extrapolation to long storage periods unrealistic. There are other studies conducted under far more realistic conditions which, for high level glasses of typical compositions, have not furnished any evidence of the effects postulated. In any case, the assertion is false that the influence of recoil atoms had been overlooked in glasses used for waste solidification and that the simulated storage periods achieved had been too short. (orig.)

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

    International Nuclear Information System (INIS)

    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

  11. SENSITIVITY OF HANFORD IMMOBILIZED HIGH LEVEL WASTE (IHLW) GLASS MASS TO CHROMIUM AND ALUMINUM PARTITIONING ASSUMPTIONS

    International Nuclear Information System (INIS)

    The strategy for the treatment of the Hanford Site tank wastes involves water and caustic washing of the tank waste sludges to reduce sludge mass and the corresponding mass of high-level waste (HLW) glass that will be generated by the Waste Treatment and Immobilization Plant (WTP). During fiscal year (FY) 2003 CH2M HILL Hanford Group, Inc. (CH2M HILL) developed revised water wash and caustic leach factors for chromium (RPP-10222) and aluminum (RPP-11079) to estimate the waste treatment behavior of the tank waste compositions. Subsequently, the U.S. Department of Energy, Office of River Protection (ORP) requested that CH2M HILL evaluate the potential impacts to the HLW glass mass due to these revised water wash and caustic leach factors. ORP plans to use the results of this study in conjunction with separate information regarding the process impacts of implementing oxidative leaching at the WTP to determine whether oxidative leaching is adequate to mitigate potential increases in HLW glass production or whether additional strategies are required. The purpose of this sensitivity study of immobilized HLW glass mass to chromium and aluminum partitioning assumptions was to: (1) Identify the impacts of the revised water wash and caustic leach factors for chromium and aluminum on the mass of HLW glass. (2) Understand the effect of oxidative leaching on the mass of HLW glass. (3) Identify the major influences for HLW glass mass and waste blending. (4) Characterize the degree of pretreatment (water washing, caustic leaching, and oxidative leaching) assumed for different source tanks. (5) Identify candidate tanks for opportunistic sampling and testing to confirm the inventory and better understand the behavior of chromium during retrieval, staging, and subsequent processing. The study concluded that: (1) Application of the revised chromium and aluminum wash and leach factors will increase the HLW glass mass by about 60 to 100 percent (using the relaxed glass properties model

  12. Advanced HLW management strategies employing both synroc and borosilicate glass waste-forms

    International Nuclear Information System (INIS)

    Recent resurgence of interest in waste partitioning permits the consideration of advanced strategies for Righ-level waste (HLW) management based on exploitation of Synroc in conjunction with borosilicate glass. The synergies resulting from the complementary of these waste-forms and their respective process technologies opens up the opportunity to reduce the overall volume of conditioned HLW for geological disposal. The paper provides a summary of the salient features of Synroc and discusses strategies for the conditioning of partitioned wastes from the reprocessing of UOX and MOX fuels from nuclear power generation. The discussion will also explore potential in U.S. defence waste remediation and disposition of excess fissile materials such as Pu. (authors)

  13. 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.

  14. Study of the release of various elements from HLW glasses in contact with porous media

    International Nuclear Information System (INIS)

    When a borosilicate glass incorporating HLW is embedded in a static porous media the leached elements diffuse into the media. The release of the various elements from the glass is governed by different mechanisms. Soluble elements (such as Cs or Tc) will be released as a direct consequence of the glass degradation. The release of elements which are less soluble than silica, on the contrary, will be dominated by the solubility limit and by the subsequent diffusion in the porous media. An integral experiment has been devised which allows one to measure the leaching rate and the diffusion of the various species in the porous media at the same time. 8 refs., 6 figs., 1 tab

  15. Database development of glass dissolution and radionuclide migration for performance analysis of HLW repository in Japan

    International Nuclear Information System (INIS)

    Japan Nuclear Cycle Development Institute (JNC, the successor of Power Reactor and Nuclear Fuel Development Corporation (PNC)) has published the second progress report for high-level radioactive waste (HLW) disposal in Japan (H12 report) in November, 1999. This report is important to obtain the confidence of HLW disposal system and to establish the implementation body in 2000. JNC has developed databases of glass dissolution and radionuclide migration for performance analysis of the engineered barrier system (EBS) and the geosphere for H12 report. The databases developed for H12 report are of dissolution rates of high-level radioactive vitrified waste, thermochemical data of radioactive elements (JNC-TDB), sorption/diffusion data in the EBS and the geosphere. The database development has been focused on the repository conditions; reducing conditions and compacted/intact system, e.g., actinide (IV)/(III), derivation of sorption coefficients from diffusion experiments rather than batch sorption experiments. The JNC-TDB and sorption database have been developed under the auspices of international experts. The quality of these databases has been checked through independent individual experiments; glass leaching, solubility, batch sorption, diffusion experiments and through coupled leaching experiments by using the fully high-level radioactive glass and plutonium-doped glass which were sandwiched between compacted bentonite saturated with water. The maximum concentration of insoluble elements dissolved from the glass has also been investigated to check the quality of the JNC-TDB by comparison with solubility prediction. Based on these studies, JNC has determined the transport parameters for H12 report; dissolution rate of glass for a soluble radioactive element (Cs), solubility for insoluble radioactive elements (e.g., actinides, Tc), distribution coefficients and effective diffusion coefficients in the EBS and the geosphere

  16. Development on Glass Formulation for Aluminum Metal and Glass Fiber

    International Nuclear Information System (INIS)

    Vitrification technology has been widely applied as one of effective processing methods for wastes generated in nuclear power plants. The advantage of vitrifying for low- and intermediate-level radioactive wastes has a large volume reduction and good durability for the final products. Recently, a filter using on HVAC(Heating Ventilating and Air Conditioning System) is composed with media (glass fiber) and separator (aluminum film) has been studied the proper treatment technology for meeting the waste disposal requirement. Present paper is a feasibility study for the filter vitrification that developing of the glass compositions for filter melting and melting test for physicochemical characteristic evaluation. The aluminum metal of film type is preparing with 0.5 cm size for proper mixing with glass frit, glass fiber is also preparing with 1 cm size within crucible. The glass compositions should be developed considering molten glass are related with wastes reduction. Glass compositions obtained from developing on glass formulation are mainly composed of SiO2and B2O3for aluminum metal. A variety of factors obtained from the glass formulation and melting test are reviewed, which is feeding rate and glass characteristics of final products such as durability for implementing the wastes disposal requirement.

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

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K.

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, A. A. [Department of Energy, Office of River Protection, Richland, WA (United States); Rodriguez, C. A. [Pacific Northwest National Laboratory, Richland, WA (United States); Matyas, J. [Pacific Northwest National Laboratory, Richland, WA (United States); Owen, A. T. [Pacific Northwest National Laboratory, Richland, WA (United States); Jansik, D. P. [Pacific Northwest National Laboratory, Richland, WA (United States); Lang, J. B. [Pacific Northwest National Laboratory, Richland, WA (United States)

    2012-11-12

    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 {mu}m, 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.

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. M.

    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

  1. PAIRWISE BLENDING OF HIGH LEVEL WASTE (HLW)

    Energy Technology Data Exchange (ETDEWEB)

    CERTA, P.J.

    2006-02-22

    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. Dose Calculations for the Co-Disposal WP-of HLW-Glass and the Triga SNF

    Energy Technology Data Exchange (ETDEWEB)

    G. Radulescu

    1999-08-02

    This calculation is prepared by the Monitored Geologic Repository (MGR) Waste Package Operations (WPO). The purpose of this calculation is to determine the surface dose rates of a codisposal waste package (WP) containing a centrally located Department of Energy (DOE) standardized 18-in. spent nuclear fuel (SNF) canister, loaded with the TRIGA (Training, Research, Isotopes, General Atomics) SNF. This canister is surrounded by five 3-m long canisters, loaded with Savannah River Site (SRS) high-level waste (HLW) glass. The results are to support the WP design and radiological analyses.

  3. 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.

  4. 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

  5. Effect of composition of titanium silicate glasses containing simulated Cs-loaded HLW on the leaching behavior

    International Nuclear Information System (INIS)

    The titanium silicate glasses containing simulated Cs-loaded HLW were prepared. MCC-1 method was used to test the chemical durability of the glasses. The simulated solid samples were immersed for 7 days in deionized water at 90 degree C. The ratio of the surface area of sample to the volume of deionized water is 10 m-1. The leachate was analyzed with ICP-AES and AAS. The results show that the leachate is mainly composed of Na+ and Si4+. The amount of ZrO2 in the glass is a major factor affecting the leaching processes. When the glass contains about 1.5%-4.5% ZrO2, the concentration of Na+ and Si4+ in the leachate is greatly decreased. The amount of TiO2 and SiO2 also affect the concentration of Na+ and Si4+ in the leachate, the suitable ratio of TiO2/SiO2 in the system is about 0.40

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

    International Nuclear Information System (INIS)

    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)

  7. 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

  8. Data Summary Report Small Scale Melter Testing Of HLW Algorithm Glasses Matrix1 Tests VSL-07S1220-1, Rev. 0, 7/25/07

    International Nuclear Information System (INIS)

    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 T1%, 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 melter

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

    International Nuclear Information System (INIS)

    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)

  10. Enhanced Sulfate Management in HLW Glass Formulations VSL12R2540-1 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); Kot, Wing [The Catholic University of America, Washington, DC (United States); Gan, Hao [The Catholic University of America, Washington, DC (United States); Matlack, Keith S. [The Catholic University of America, Washington, DC (United States)

    2012-11-13

    The Low Activity Waste (LAW) tanks that are scheduled to provide the Hanford Tank Waste Treatment and Immobilization Plant (WTP) with waste feeds contain significant amounts of sulfate. The sulfate content in the LAW feeds is sufficiently high that a separate molten sulfate salt phase may form on top of the glass melt during the vitrification process unless suitable glass formulations are employed and sulfate levels are controlled. Since the formation of the salt phase is undesirable from many perspectives, mitigation approaches had to be developed. Considerable progress has been made and reported by the Vitreous State Laboratory (VSL) in enhancing sulfate incorporation into LAW glass melts and developing strategies to manage and mitigate the risks associated with high-sulfate feeds.

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

    International Nuclear Information System (INIS)

    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 PuO2.xH2O. It is therefore concluded that PuO2.xH2O 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.)

  12. Aluminum/glass fibre and aluminum/carbon fibre hybrid laminates

    Directory of Open Access Journals (Sweden)

    Ana STAN

    2010-06-01

    Full Text Available The metal/fibre hybrid laminates consist of an alternation of 0.2 ÷ 0.5 mm metallic sheets(Aluminum or Titanium in Aeronautical Engineering and pre-pregs made of unidirectional carbon oraramid or glass fibre or of the two-dimensional fabric of these materials, bonded by a polymeradhesive (epoxy, especially. Compared with the monolithic metal foils, the essential quality of thesehybrid laminates is their superior resistance to fatigue, impact and crack propagation (existing ormade by notches. The paper presents some results regarding hybrid laminates aluminium-carbonfibre and aluminum-glass fibre achieved in the CEEX project X1C05 (2005.

  13. HIGH LEVEL WASTE (HLW) VITRIFICATION EXPERIENCE IN THE US: APPLICATION OF GLASS PRODUCT/PROCESS CONTROL TO OTHERHLW AND HAZARDOUS WASTES

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C; James Marra, J

    2007-09-17

    Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. At the Savannah River Site (SRS) actual HLW tank waste has successfully been processed to stringent product and process constraints without any rework into a stable borosilicate glass waste since 1996. A unique 'feed forward' statistical process control (SPC) has been used rather than statistical quality control (SQC). In SPC, the feed composition to the melter is controlled prior to vitrification. In SQC, the glass product is sampled after it is vitrified. Individual glass property models form the basis for the 'feed forward' SPC. The property models transform constraints on the melt and glass properties into constraints on the feed composition. The property models are mechanistic and depend on glass bonding/structure, thermodynamics, quasicrystalline melt species, and/or electron transfers. The mechanistic models have been validated over composition regions well outside of the regions for which they were developed because they are mechanistic. Mechanistic models allow accurate extension to radioactive and hazardous waste melts well outside the composition boundaries for which they were developed.

  14. Raman and Infrared Spectroscopy of Yttrium Aluminum Borate Glasses and Glass-ceramics

    Science.gov (United States)

    Bradley, J.; Brooks, M.; Crenshaw, T.; Morris, A.; Chattopadhyay, K.; Morgan, S.

    1998-01-01

    Raman spectra of glasses and glass-ceramics in the Y2O3-Al2O3-B2O3 system are reported. Glasses with B2O3 contents ranging from 40 to 60 mole percent were prepared by melting 20 g of the appropriate oxide or carbonate powders in alumina crucibles at 1400 C for 45 minutes. Subsequent heat treatments of the glasses at temperatures ranging from 600 to 800 C were performed in order to induce nucleation and crystallization. It was found that Na2CO3 added to the melt served as a nucleating agent and resulted in uniform bulk crystallization. The Raman spectra of the glasses are interpreted primarily in terms of vibrations of boron - oxygen structural groups. Comparison of the Raman spectra of the glass-ceramic samples with spectra of aluminate and borate crystalline materials reveal that these glasses crystallize primarily as yttrium aluminum borate, YAl3(BO3)4.

  15. 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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  18. 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

  19. Final Report - Sulfate Solubility in RPP-WTP HLW Glasses, VSL-06R6780-1, Rev. 0

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-03

    This report describes the results of work and testing specified by Test Specifications 24590-HLW-TSP-RT-01-006 Rev 1, Test Plans VSL-02T7800-1 Rev 1 and Test Exceptions 24590-HLW-TEF-RT-05-00007. The work and any associated testing followed established quality assurance requirements and were conducted as authorized. The descriptions provided in this report are an accurate account of both the conduct of the work and the data collected. Results required by the Test Plans are reported. Also reported are any unusual or anomalous occurrences that are different from the starting hypotheses. The test results and this report have been reviewed and verified.

  20. Qualification of a Radioactive High Aluminum Glass for Processing in the Defense Waste Processing Facility at the Savannah River Site

    International Nuclear Information System (INIS)

    At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a borosilicate glass for approximately eleven years. Currently the DWPF is immobilizing HLW sludge in Sludge Batch 4 (SB4). Each sludge batch is nominally two million liters of HLW and produces nominally five hundred stainless steel canisters 0.6 meters in diameter and 3 meters tall filled with the borosilicate glass. In SB4 and earlier sludge batches, the Al concentration has always been rather low, (less than 9.5 weight percent based on total dried solids). It is expected that in the future the Al concentrations will increase due to the changing composition of the HLW. Higher Al concentrations could introduce problems because of its known effect on the viscosity of glass melts and increase the possibility of the precipitation of nepheline in the final glass and decrease its durability. In 2006 Savannah River National Laboratory (SRNL) used DWPF processes to immobilize a radioactive HLW slurry containing 14 weight percent Al to ensure that this waste is viable for future DWPF processing. This paper presents results of the characterization of the high Al glass prepared in that demonstration. At SRNL, a sample of the processed high Al HLW slurry was mixed with an appropriate glass frit as performed in the DWPF to make a waste glass containing nominally 30% waste oxides. The glass was prepared by melting the frit and waste remotely at 1150 deg. C. The glass was then characterized by - determining the chemical composition of the glass including the concentrations of several actinide and U-235 fission products, - calculating the oxide waste loading of the glass based on the chemical composition and comparing it to that of the target - determining if the glass composition met the DWPF processing constraints such as glass melt viscosity and liquidus temperature along with a waste form affecting constraint that prevents

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

    International Nuclear Information System (INIS)

    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)

  2. Determination of long-lived fission products and actinides in Savannah River site HLW sludge and glass for waste acceptance

    International Nuclear Information System (INIS)

    Savannah River Site (SRS) is currently immobilizing the radioactive, caustic, high-level waste sludge in Tank 51 into a borosilicate glass for disposal in a geologic repository. A requirement for repository acceptance is that SRS report the concentrations of certain fission product and actinide radionuclides in the glass. This paper presents measurements of many of these concentrations in both Tank 51 sludge and the final glass. The radionuclides were measured by inductively coupled plasma - mass spectrometry and α, β, and γ counting methods. Examples of the radionuclides are Sr-90, Cs-137, U-238, Pu-239, and Cm-244. Concentrations in the glass are 3.1 times lower due to dilution of the sludge with a nonradioactive glass forming frit in the vitrification process. Results also indicated that in both the sludge and glass the relative concentrations of the long lived fission products insoluble in caustic area in proportion to their yields from the fission of U-235 in the SRS reactors. This allowed the calculation of a fission yield scaling factor. This factor in addition to the sludge dilution factor can be used to estimate concentrations of waste acceptance radionuclides that cannot be measured in the glass

  3. Determination of long-lived fission products and actinides in Savannah River Site HLW sludge and glass for waste acceptance

    International Nuclear Information System (INIS)

    Savannah River Site (SRS) is immobilizing the radioactive, high-level waste sludge in Tank 51 into a borosilicate glass for disposal in a geologic repository. A requirement for repository acceptance is that SRS report the concentrations of certain fission product and actinide radionuclides in the glass. This paper presents measurements of many of these concentrations in both Tank 51 sludge and the final glass. The radionuclides were measured by inductively coupled plasma mass spectrometry and α, β, and γ counting methods. Examples of the radionuclides are 90Sr, 137Cs, 238U and , 239Pu. Concentrations in the glass are 3.1 times lower due to dilution of the sludge with a nonradioactive glass forming frit in the vitrification process. Results also indicated that in both the sludge and glass the relative concentrations of the long lived fission products insoluble in caustic are in proportion to their yields from the fission of 235U waste in the SRS reactors. This allowed the calculation of a fission yield scaling factor. This factor in addition to the sludge dilution factor can be used to estimate concentrations of waste acceptance radionuclides that cannot be measured in the glass. Examples of these radionuclides are 79Se, 93Zr, and 107Pd. (author)

  4. Acoustic Emission Characteristics during fracture Process of Glass Fiber/Aluminum Hybrid Laminates

    International Nuclear Information System (INIS)

    Fracture behaviors and acoustic emission (AE) characteristics of single-edge-notched monolithic aluminum plates and glass fiber/aluminum hybrid laminate plates have been investigated under tensile loads. AE signals from monolithic aluminum could be classified into two different types: signals with low frequency band and high frequency band. High frequency signals were detected in the post stage of loading beyond displacement of 0.45mm. For glass fiber/aluminum laminates, AE signals with high amplitude and long duration were additionally confirmed on FFT frequency analysis, which corresponded to macro-crack propagation and/or delamination between A1 and fiber layers. On the basis of the above AE analysis and fracture observation with optical microscopy and ultrasonic T scan, characteristic features of AE associated with fracture processes of single-edge-notched glass fiber/aluminum laminates were elucidated according to different fiber ply orientations

  5. Incorporation of Fines and Noble Metals into HLW Borosilicate Glass: Industrial Responses to a Challenging Issue - 13056

    Energy Technology Data Exchange (ETDEWEB)

    Chauvin, E.; Chouard, N.; Prod' homme, A. [AREVA, AREVA NC, Paris (France); Boudot, E. [AREVA, AREVA NC, La Hague (France); Gruber, Ph.; Pinet, O. [CEA Marcoule LCV, France (France); Grosman, R. [AREVA, SGN, Paris (France)

    2013-07-01

    During the early stages of spent fuel reprocessing, the fuel rods are cut and dissolved to separate the solid metallic parts of the rods (cladding and end pieces) from the radioactive nitric acid solution containing uranium, plutonium, minor actinides and fission products (FP). This solution contains small, solid particles produced during the shearing process. These small particles, known as 'fines', are then separated from the liquid by centrifugation. At the La Hague plant in France, the fines solution is transferred to the vitrification facilities to be incorporated into borosilicate glass along with the highly radioactive FP solution. These fines are also composed of Zr, Mo and other noble metals (i.e. Ru, Pd, Rh, etc.) that are added before vitrification to the the FP solution that already contained noble metals. As noble metals has the potential to modify the glass properties (including viscosity, electrical conductivity, etc.) and to be affected by sedimentation inside the melter, their behavior in borosilicate glass has been studied in depth over the years by the AREVA and CEA teams which are now working together in the Joint Vitrification Laboratory (LCV). At La Hague, the R7 vitrification facility started operation in 1989 using induction-heated metallic melter technology and was quickly followed by the T7 vitrification facility in 1992. Incorporating the fines into glass has been a challenge since operation began, and has given rise to several R and D studies resulting in a number of technological enhancements to improve the mixing capability of the melters (multiple bubbling technology and mechanical stirring in the mid-90's). Nowadays, the incorporation of fines into R7T7 glass is well understood and process adaptations are deployed in the La Hague facilities to increase the operating flexibility of the melters. The paper will briefly describe the fines production mechanisms, give details of the resulting fines characteristics, explain

  6. Development of a novel aluminum-free glass ionomer cement based on magnesium/strontium-silicate glasses

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong-Ae [Department of Biomaterials Science, College of Dentistry, Dankook University, Cheonan 330-714 (Korea, Republic of); Department of Nanobiomedical Science and BK21 Plus NBM Global Research Center for Regenerative Medicine, Dankook University Graduate School, Cheonan 330-714 (Korea, Republic of); Abo-Mosallam, Hany A. [Glass Research Department, National Research Centre, Dokki, Cairo (Egypt); Lee, Hye-Young [Department of Nanobiomedical Science and BK21 Plus NBM Global Research Center for Regenerative Medicine, Dankook University Graduate School, Cheonan 330-714 (Korea, Republic of); Institute of Tissue Regeneration Engineering (ITREN), Dankook University, Cheonan 330-714 (Korea, Republic of); Kim, Gyu-Ri [Department of Biomaterials Science, College of Dentistry, Dankook University, Cheonan 330-714 (Korea, Republic of); Department of Nanobiomedical Science and BK21 Plus NBM Global Research Center for Regenerative Medicine, Dankook University Graduate School, Cheonan 330-714 (Korea, Republic of); Kim, Hae-Won [Department of Biomaterials Science, College of Dentistry, Dankook University, Cheonan 330-714 (Korea, Republic of); Department of Nanobiomedical Science and BK21 Plus NBM Global Research Center for Regenerative Medicine, Dankook University Graduate School, Cheonan 330-714 (Korea, Republic of); Institute of Tissue Regeneration Engineering (ITREN), Dankook University, Cheonan 330-714 (Korea, Republic of); Lee, Hae-Hyoung, E-mail: haelee@dku.edu [Department of Biomaterials Science, College of Dentistry, Dankook University, Cheonan 330-714 (Korea, Republic of); Institute of Tissue Regeneration Engineering (ITREN), Dankook University, Cheonan 330-714 (Korea, Republic of)

    2014-09-01

    The effects of strontium substitution for magnesium in a novel aluminum-free multicomponent glass composition for glass ionomer cements (GICs) were investigated. A series of glass compositions were prepared based on SiO{sub 2}-P{sub 2}O{sub 5}-CaO-ZnO-MgO{sub (1-X)}-SrO{sub X}-CaF{sub 2} (X = 0, 0.25, 0.5 and 0.75). The mechanical properties of GICs prepared were characterized by compressive strength, flexural strength, flexural modules, and microhardness. Cell proliferation was evaluated indirectly by CCK-8 assay using various dilutions of the cement and rat mesenchyme stem cells. Incorporation of strontium instead of magnesium in the glasses has a significant influence on setting time of the cements and the properties. All mechanical properties of the GICs with SrO substitution at X = 0.25 were significantly increased, then gradually decreased with further increase of the amount of strontium substitution in the glass. The GIC at X = 0.25, also, showed an improved cell viability at low doses of the cement extracts in comparison with other groups or control without extracts. The results of this study demonstrate that the glass compositions with strontium substitution at low levels can be successfully used to prepare aluminum-free glass ionomer cements for repair and regeneration of hard tissues. - Highlights: • We developed multicomponent glass compositions for a novel aluminum-free glass ionomer cement (GIC). • The effects of MgO replacement with SrO in the glasses on the mechanical properties and cell proliferation were evaluated. • Substitution of MgO with SrO at low levels led to improvement of mechanical properties and cell viability of the cements. • Microstructural degradations in the cement matrix of the GICs with strontium at high levels were observed after aging.

  7. Development of a novel aluminum-free glass ionomer cement based on magnesium/strontium-silicate glasses

    International Nuclear Information System (INIS)

    The effects of strontium substitution for magnesium in a novel aluminum-free multicomponent glass composition for glass ionomer cements (GICs) were investigated. A series of glass compositions were prepared based on SiO2-P2O5-CaO-ZnO-MgO(1-X)-SrOX-CaF2 (X = 0, 0.25, 0.5 and 0.75). The mechanical properties of GICs prepared were characterized by compressive strength, flexural strength, flexural modules, and microhardness. Cell proliferation was evaluated indirectly by CCK-8 assay using various dilutions of the cement and rat mesenchyme stem cells. Incorporation of strontium instead of magnesium in the glasses has a significant influence on setting time of the cements and the properties. All mechanical properties of the GICs with SrO substitution at X = 0.25 were significantly increased, then gradually decreased with further increase of the amount of strontium substitution in the glass. The GIC at X = 0.25, also, showed an improved cell viability at low doses of the cement extracts in comparison with other groups or control without extracts. The results of this study demonstrate that the glass compositions with strontium substitution at low levels can be successfully used to prepare aluminum-free glass ionomer cements for repair and regeneration of hard tissues. - Highlights: • We developed multicomponent glass compositions for a novel aluminum-free glass ionomer cement (GIC). • The effects of MgO replacement with SrO in the glasses on the mechanical properties and cell proliferation were evaluated. • Substitution of MgO with SrO at low levels led to improvement of mechanical properties and cell viability of the cements. • Microstructural degradations in the cement matrix of the GICs with strontium at high levels were observed after aging

  8. 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.

  9. NMR studies of aluminum speciation in tellurite glasses

    International Nuclear Information System (INIS)

    Tellurite glasses have recently been investigated as hosts for rare earth metal ions for application in telecommunication systems. These glasses, when doped with erbium ions, exhibit favorable optical properties, including high emission cross-sections and quantum efficiencies, and relatively broad emission spectra. Many tellurite glasses also exhibit good optical transparency and thermal stability, which are desirable for fiber fabrication. We report results from solid state NMR studies of the aluminium coordination environments in Ta alumino-tellurite glasses. These data show the presence of 4-,5- and 6-coordinate aluminium in all glasses. The aluminium speciation is independent of the Ta/Al ratio, but is strongly affected by the Te O2 content. In glasses rich in Te O2, aluminium is primarily in octahedral coordination. At flower of Te O2, the aluminium gradually shifts towards a larger tetrahedral population. The percentage of 5-coordinate aluminium remains relatively constant in all Ta-Al-tellurite glasses. The average coordination number of aluminium is also affected by the thermal history of the glasses, as this value increases after annealing of rapidly quenched glasses. The significance of this rich structural chemistry will be discussed in relation to the desired physical properties of such glasses for application as optical components in telecommunication systems. (author)

  10. Formation of Nanoporous Anodic Alumina by Anodization of Aluminum Films on Glass Substrates.

    Science.gov (United States)

    Lebyedyeva, Tetyana; Kryvyi, Serhii; Lytvyn, Petro; Skoryk, Mykola; Shpylovyy, Pavlo

    2016-12-01

    Our research was aimed at the study of aluminum films and porous anodic alumina (PAA) films in thin-film РАА/Al structures for optical sensors, based on metal-clad waveguides (MCWG). The results of the scanning electron microscopy (SEM) and atomic force microscopy (AFM) studies of the structure of Al films, deposited by DC magnetron sputtering, and of PAA films, formed on them, are presented in this work.The study showed that the structure of the Al films is defined by the deposition rate of aluminum and the thickness of the film. We saw that under anodization in 0.3 M aqueous oxalic acid solution at a voltage of 40 V, the PAA film with a disordered array of pores was formed on aluminum films 200-600 nm thick, which were deposited on glass substrates with an ultra-thin adhesive Nb layer. The research revealed the formation of two differently sized types of pores. The first type of pores is formed on the grain boundaries of aluminum film, and the pores are directed perpendicularly to the surface of aluminum. The second type of pores is formed directly on the grains of aluminum. They are directed perpendicularly to the grain plains. There is a clear tendency to self-ordering in this type of pores. PMID:27083584

  11. Optical Properties of Au Nanoparticles Coated on Surface of Glass or Anodic Aluminum Oxide Template

    Institute of Scientific and Technical Information of China (English)

    FENG Jinyang; WU Can; MA Xiao; ZHANG Hongquan; ZHAO Xiujian

    2012-01-01

    Au nanoparticles coated on the surface of glass (Sample A) or on anodic aluminum oxide template surface (Sample B) were prepared using titanium dioxide sol-gel doped with chloroauric acid and with a reduction process.FE-SEM,UV-Vis spectrum and Fluorescence spectrum tests show that Au nanoparticles have been distributed randomly on the surface of glass,while deposition occurs on the surface of regular hollows for anodic aluminum oxide template.A sharp absorption peak appears at the wavelength of 536 nm for sample B,while there is a red shift,with a broader peak for sample A.A distinct fluorescence emission at the wavelength of 633 nm is detected for sample A,but no noticeable fluorescence emission has been found for Sample B.The results indicate that the microstructure and optical properties of Au nanoparticles can be modulated by different substrate.

  12. 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.

  13. Friction Stir Welding of Zr_(55)Al_(10)Ni_5Cu_(30) Bulk Metallic Glass to Crystalline Aluminum

    Institute of Scientific and Technical Information of China (English)

    Zuoxiang Qin; Cuihong Li; Haifeng Zhang; Zhongguang Wang; Zhuangqi Hu; Zhiqiang Liu

    2009-01-01

    The Zr_(55)Al_(10)Ni_5Cu_(30) bulk metallic glass plate were successfully welded to crystalline aluminum plates by using a friction stir welding (FSW) method. The welded zone was examined. No defects, cracks or pores were observed and no other crystalline phases except for aluminum were found in the welded joint. The strength of the joint is higher than that of aluminum. The glassy phase in the stir zone keeps the amorphous state, showing a successful welding. The storage modulus softens over the glass transition. And the weldability was discussed according to this phenomena.

  14. 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. PMID:24261406

  15. 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.

  16. Fabrication of Poly-Si Thin Film on Glass Substrate by Aluminum-induced Crystallization

    Institute of Scientific and Technical Information of China (English)

    XU Man; XIA Donglin; YANG Sheng; ZHAO Xiujian

    2006-01-01

    Amorphous silicon (a-Si) thin films were deposited on glass substrate by PECVD,and polycrystalline silicon (poly-Si) thin films were prepared by aluminum-induced crystallization (AIC). The effects of annealing temperature on the microstructure and morphology were investigated. The AIC poly-Si thin films were characterized by XRD, Raman and SEM. It is found that a-Si thin film has a amorphous structure after annealing at 400 ℃ for 20 min, a-Si films begin to crystallize after annealing at 450 ℃ for 20 min, and the crystallinity of a-Si thin films is enhanced obviously with the increment of annealing temperature.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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.

  19. 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.

  20. 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.

  1. High Temperature Mechanical Behavior of Aluminum- Cu50Zr50Metallic Glass Interface

    Science.gov (United States)

    Gupta, P.; Yedla, N.

    2016-02-01

    Molecular dynamics (MD) simulations are carried out to determine interface strength between aluminum (metal) and Cu50Zr50 (metallic glass) at temperature of 500 K and at strain rate of 108 s-1. Simulation box of size 100 Å (x) × 110 Å (y) × 50 Å (z) is used for the above studies. At first Al-Cu50Zr50 crystalline interface model is built with the base layer-Al of 50 A and the top layer-Cu50Zr50 of 55 Å along y-direction. Later Cu50Zr50 metallic glass is obtained by quenching at a cooling rate of 4 x 1012 Ks-1. NPT ensemble is used in metallic glass preparation simulation. The interface model is then equilibrated at 300 K for 500 ps to relieve the internal stresses. EAM (Embedded Atom Method) potential is used for modelling the interaction between Al-Cu-Zr atoms. The interface strength of Al-Cu50Zr50 model interface is determined by applying load in the directions normal (mode-I) and parallel (mode-II) to the interface. NVT ensemble is used for the deformation studies. In mode-I for perfect and cracked interface, the interface fractures in the Al-region via necking. Sticking of the Al-atoms to the metallic glass is observed in both the loading conditions. Also, multiple voids are nucleated at the interface.

  2. Toughness and fracture mechanisms of glass fiber/aluminum hybrid laminates under tensile loading

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Sung Choong [Konkuk University, Seoul (Korea, Republic of); Choi, Nak Sam; Chang, Young Wook [Hanyang University, Ansan (Korea, Republic of)

    2007-12-15

    Tensile properties and fracture toughness of monolithic aluminum (Al), glass fiber reinforced plastics (GFRPs) and glass fiber/aluminum hybrid laminates (GFMLs) were examined in relation to the fracture processes of plain coupon and single-edge-notched specimens. Elastic modulus and ultimate tensile strength of GFMLs showed characteristic dependences on the kind of Al, fiber orientation and the Al/fiber layer composition ratio. Fracture toughnesses KC and GC of A-GFML-UD were comparable to those of GFRP-UD and were much superior to monolithic Al. However, GFML with a transverse crack parallel to the fiber layer deteriorated largely in toughness. Microscopic observation of the fracture zone in the vicinity of the crack tip revealed various modes of micro-cracks in the respective layers as well as fiber fractures and delamination between fiber/Al layers. Such damage advances in GFMLs dependent on the orientation of the fiber layer and the Al/fiber composition ratio strongly influenced the strength and toughness of GFMLs.

  3. Toughness and fracture mechanisms of glass fiber/aluminum hybrid laminates under tensile loading

    International Nuclear Information System (INIS)

    Tensile properties and fracture toughness of monolithic aluminum (Al), glass fiber reinforced plastics (GFRPs) and glass fiber/aluminum hybrid laminates (GFMLs) were examined in relation to the fracture processes of plain coupon and single-edge-notched specimens. Elastic modulus and ultimate tensile strength of GFMLs showed characteristic dependences on the kind of Al, fiber orientation and the Al/fiber layer composition ratio. Fracture toughnesses KC and GC of A-GFML-UD were comparable to those of GFRP-UD and were much superior to monolithic Al. However, GFML with a transverse crack parallel to the fiber layer deteriorated largely in toughness. Microscopic observation of the fracture zone in the vicinity of the crack tip revealed various modes of micro-cracks in the respective layers as well as fiber fractures and delamination between fiber/Al layers. Such damage advances in GFMLs dependent on the orientation of the fiber layer and the Al/fiber composition ratio strongly influenced the strength and toughness of GFMLs

  4. Multi-channel transition emissions of Sm3+ in lithium yttrium aluminum silicate glasses and derived opalescent glass ceramics

    International Nuclear Information System (INIS)

    Highlights: • Multi-channel transitions of Sm3+ in aluminosilicate glasses are investigated. • Judd–Ofelt parameters are derived to be 4.12 × 10−20, 4.26 × 10−20 and 2.11 × 10−20 cm2. • Cross-section profiles σem for visible and NIR emissions are calculated by FL formula. • 44.9% internal quantum efficiency and 11.58% external quantum yield are exposed. • Orientational crystallization was observed after the heat-treatment testing. -- Abstract: Sm3+-doped lithium–yttrium–aluminum-silicate (LYAS) glasses have been fabricated and effective visible and near-infrared (NIR) emissions are exhibited. The predicated spontaneous emission probabilities (Arad) are derived to be 29.1, 149.2, 150.1 and 39.0 s−1 for conventional visible emissions assigned to 4G5/2 → 6HJ (J = 5/2, 7/2, 9/2 and 11/2) transitions, and Arad are obtained to be 24.8, 4.6 and 2.7 s−1 for the optical transitions 4G5/2 → 6FJ (J = 5/2, 7/2 and 9/2) corresponding to the NIR emissions, respectively. The maximum stimulated emission cross-sections (σem) are obtained to be 7.45 × 10−22 and 0.48 × 10−22 cm2 for visible (4G5/2 → 6H9/2) and NIR (4G5/2 → 6F9/2) emissions, respectively. Internal quantum efficiency for the 4G5/2 level and external quantum yield for visible emissions of Sm3+ are determined to be 44.9% and 11.58%, respectively. High temperature heat treatment testing was carried out on the LYAS glasses and orientational crystallization was observed. Visible transition emissions of Sm3+ with obvious Stark splitting indicate Sm3+ ions have been introduced into the YAG micro-crystal formed in the glass ceramics. Investigations on multi-channel radiative transition emissions of Sm3+ in LYAS glasses provide a new clue to develop tunable lasers, compact light sources and optoelectronic devices

  5. Penetration experiments in aluminum 1100 targets using soda-lime glass projectiles

    Science.gov (United States)

    Horz, Friedrich; Cintala, Mark J.; Bernhard, Ronald P.; Cardenas, Frank; Davidson, William E.; Haynes, Gerald; See, Thomas H.; Winkler, Jerry L.

    1995-01-01

    The cratering and penetration behavior of annealed aluminum 1100 targets, with thickness varied from several centimeters to ultra-thin foils less than 1 micrometer thick, were experimentally investigated using 3.2 mm diameter spherical soda-lime glass projectiles at velocities from 1 to 7 km/s. The objective was to establish quantitative, dimensional relationships between initial impact conditions (impact velocity, projectile diameter, and target thickness) and the diameter of the resulting crater or penetration hole. Such dimensional relationships and calibration experiments are needed to extract the diameters and fluxes of hypervelocity particles from space-exposed surfaces and to predict the performance of certain collisional shields. The cratering behavior of aluminum 1100 is fairly well predicted. However, crater depth is modestly deeper for our silicate impactors than the canonical value based on aluminum projectiles and aluminum 6061-T6 targets. The ballistic-limit thickness was also different. These differences attest to the great sensitivity of detailed crater geometry and penetration behavior on the physical properties of both the target and impactor. Each penetration experiment was equipped with a witness plate to monitor the nature of the debris plume emanating from the rear of the target. This plume consists of both projectile fragments and target debris. Both penetration hole and witness-plate spray patterns systematically evolve in response to projectile diameter/target thickness. The relative dimensions of the projectile and target totally dominate the experimental products documented in this report; impact velocity is an important contributor as well to the evolution of penetration holes, but is of subordinate significance for the witness-plate spray patterns.

  6. Blunt notch strength of hybrid boron/glass/aluminum fiber metal laminates

    International Nuclear Information System (INIS)

    Research highlights: → The use of boron fibers increases the modulus and yielding stress of FMLs. → High modulus and high ductility FMLs are achieved by mingling boron and glass fibers. → The notched hybrid FMLs also exhibit excellent strength retaining ability. → Finite element analysis was used to simulate the notch behavior of hybrid FMLs. - Abstract: The notch strength of high modulus hybrid fiber/metal laminates (FMLs) was investigated. The composite layers used in this material, which contain both boron fibers and S2-glass fibers, were adhesively bonded to 2024-T3 aluminum sheets and consolidated using an autoclave process. The results of tensile tests clearly showed that high modulus FMLs with a good ductility can be achieved by mingling of boron and glass fibers. The effects of notch sizes and constituents on the failure behavior were determined. The experiments showed that the notched hybrid FMLs exhibited excellent strength retaining characteristics even with the presence of large notches. Microscopy, X-ray radiography and chemical removal technique were used to examine the fracture characteristics of hybrid FMLs. A finite element analysis (FEA) model was established to analyze the notch behavior of hybrid FMLs. Experimental results of the blunt-notch strength are in good agreement with the stresses calculated by computational modeling of hybrid FMLs.

  7. Long range behaviour of HLW glasses: on the problem of the safe disposal of α-bearing wastes from reprocessing and fuel-element production

    International Nuclear Information System (INIS)

    An evaluation of questions concerning the safety aspects during perpetual storage of radioactive wastes is peformed on the basis of the concentration of α-emitters in the wastes, the relative toxicities of these α-emitters and the long-range stability of the solidified products towards the disintegration of these α-nuclides. An overview is given on all categories of α-bearing wastes from fuel reprocessing and refabrication. As an example the long-range behaviour of high-level waste glasses is discussed and investigated by an accelerated experiment with representative Cm-242 doped glass compositions

  8. 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

  9. Vitrification of sulphate bearing high level waste (HLW)

    International Nuclear Information System (INIS)

    The Indian strategy for the management of spent fuel is based on Reprocessing-Conditioning- Recycle (RCR) option. Reprocessing of spent fuel by the PUREX process leads to the generation of high-level radioactive liquid waste. Strategy for the management of high-level waste in India involves: a) Immobilization of HLW in borosilicate matrices b) Interim storage of vitrified HLW for a period of about 50 years c) Ultimate disposal of vitrified HLW in deep geological repository Borosilicate matrices have found wide acceptance for immobilization of high level wastes. Suitable glass compositions within the borosilicate family have been formulated and characterized for sulphate bearing high-level radioactive waste. Presence of sulphate in HLW, generated earlier, is on account of ferrous sulphamate as a reducing agent, added during partitioning stage of reprocessing. Solubility of sulphur in the form of sodium sulphate is very less (<1% wt) in normally deployed borosilicate melts for vitrification of HLW. The soluble alkali sulphate gets phase separated in the glass melt and its presence is not desirable since this phase is enriched with radio Cs and has high solubility in water. In addition, volatility of sulphates during glass formation is another area of concern. Attempts to address this problem were made and alternative glass forming systems based on lead and barium borosilicate systems were studied for immobilization of this sulphate bearing waste. (author)

  10. Low temperature sintering and performance of aluminum nitride/borosilicate glass

    Institute of Scientific and Technical Information of China (English)

    Hong-sheng ZHAO; Lei CHEN; Nian-zi GAO; Kai-hong ZHANG; Zi-qiang LI

    2009-01-01

    Aluminum nitride (AlN)/borosilicate glass composites were prepared by the tape casting process and hot-press sin-tered at 950 ℃ with AlN and SiO2-B2O3-ZnO-Al2O3-Li2O glass as starting materials. We characterized and analyzed the variation of the microstructure, bulk density, porosity, dielectric constant, thermal conductivity and thermal expansion coefficient (TEC) of the ceramic samples as a function of AlN content. Results show that AlN and SiO2-B2O3-ZnO-Al2O3-Li2O glass can be sintered at 950 ℃, and ZnAl2O4 and Zn2SiO4 phase precipitated to form glass-ceramic. The performance of the ceramic samples was de-termined by the composition and bulk density of the composites. Lower AlN content was found redounding to liquid phase sin-tering, and higher bulk density of composites can be accordingly obtained. With the increase of porosity, corresponding decreases were located in the dielectric constant, thermal conductivity and TEC of the ceramic samples. When the mass fraction of AlN was 40%, the ceramic samples possessed a low dielectric constant (4.5~5.0), high thermal conductivity (11.6 W/(m·K)) and a proper TEC (3.0×10K-1, which matched that of silicon). The excellent performance makes this kind of low temperature co-fired ce-ramic a promising candidate for application in the micro-electronics packaging industry.

  11. 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.

  12. HLW Disposal System Development

    International Nuclear Information System (INIS)

    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/cm3. 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 of HLW management

  13. 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

  14. Industrial scale-plant for HLW partitioning in Russia

    International Nuclear Information System (INIS)

    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 m3 HLW and 235 MCi of radionuclides was included in glass. However only 1100 m3 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

  15. The basic corrosion mechanisms of HLW glasses

    International Nuclear Information System (INIS)

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

  16. Preparation, glass forming ability, crystallization and deformation of (zirconium, hafnium)-copper-nickel-aluminum-titanium-based bulk metallic glasses

    Science.gov (United States)

    Gu, Xiaofeng

    Multicomponent Zr-based bulk metallic glasses are the most promising metallic glass forming systems. They exhibit great glass forming ability and fascinating mechanical properties, and thus are considered as potential structural materials. One potential application is that they could be replacements of the depleted uranium for making kinetic energy armor-piercing projectiles, but the density of existing Zr-based alloys is too low for this application. Based on the chemical and crystallographic similarities between Zr and Hf, we have developed two series of bulk metallic glasses with compositions of (HfxZr1-x) 52.5Cu17.9Ni14.6Al10Ti5 and (HfxZr1-x) 57Cu20Ni8Al10Ti5 ( x = 0--1) by gradually replacing Zr by Hf. Remarkably increased density and improved mechanical properties have been achieved in these alloys. In these glasses, Hf and Zr play an interchangeable role in determining the short range order. Although the glass forming ability decreases continuously with Hf addition, most of these alloys remain bulk glass-forming. Recently, nanocomposites produced from bulk metallic glasses have attracted wide attention due to improved mechanical properties. However, their crystalline microstructure (the grain size and the crystalline volume fraction) has to be optimized. We have investigated crystallization of (Zr, Hf)-based bulk metallic glasses, including the composition dependence of crystallization paths and crystallization mechanisms. Our results indicate that the formation of high number density nanocomposites from bulk metallic glasses can be attributed to easy nucleation and slowing-down growth processes, while the multistage crystallization behavior makes it more convenient to control the microstructure evolution. Metallic glasses are known to exhibit unique plastic deformation behavior. At low temperature and high stress, plastic flow is localized in narrow shear bands. Macroscopic investigations of shear bands (e.g., chemical etching) suggest that the internal

  17. Thermal inactivation of Listeria innocua in salmon (Oncorhynchus keta) caviar using conventional glass and novel aluminum thermal-death-time tubes.

    Science.gov (United States)

    Al-Holy, M; Quinde, Z; Guan, D; Tang, J; Rasco, B

    2004-02-01

    Differences in the come-up times and thermal inactivation parameters of Listeria innocua in salmon (Oncorhynchus keta) caviar containing 2.5% salt using conventional thermal-death-time (TDT) glass tubes and a novel aluminum tube were tested and compared. Generally, the come-up times and decimal reduction times (D-values) were shorter and the change in temperature required to change the D-value (z-value) was longer in the aluminum than in the glass tubes. The D-values at 60, 63, and 65 degrees C for the aluminum TDT tubes were 2.97, 0.77, and 0.40 min, respectively, and for the glass TDT tubes, these values were 3.55, 0.84, and 0.41 min. The z-values were 5.7 degrees C in the aluminum and 5.3 degrees C in the glass. Because of the shorter come-up time, the aluminum TDT tubes may provide a more precise measurement of microbial thermal inactivation than the glass TDT tubes, particularly for viscous materials, solid foods, and foods containing particulate matter. PMID:14968974

  18. 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.

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

    International Nuclear Information System (INIS)

    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

  20. 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.

  1. Feasibility study for removal of sulphate from HLW prior to its immobilisation

    International Nuclear Information System (INIS)

    Sulphate removal from HLW can improve the waste loading significantly during vitrification using barium borosilicate glass matrix. This paper describes the experimental works carried out to see the feasibility of sulphate removal from HLW by precipitation method using barium nitrate. Various parameters like optimization of stoichiometric amount of barium, extent of sulphate removal, partitioning of radionuclide and feasibility of sludge fixation in cement matrix etc were also studied. (author)

  2. Advances in the Glass Formulations for the Hanford Tank Waste Treatment and Immobilization Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Vienna, John D.; Kim, Dong Sang

    2015-01-14

    The Department of Energy-Office of River Protection (DOE-ORP) is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to treat radioactive waste currently stored in underground tanks at the Hanford site in Washington. The WTP that is being designed and constructed by a team led by Bechtel National, Inc. (BNI) will separate the tank waste into High Level Waste (HLW) and Low Activity Waste (LAW) fractions with the majority of the mass (~90%) directed to LAW and most of the activity (>95%) directed to HLW. The pretreatment process, envisioned in the baseline, involves the dissolution of aluminum-bearing solids so as to allow the aluminum salts to be processed through the cesium ion exchange and report to the LAW Facility. There is an oxidative leaching process to affect a similar outcome for chromium-bearing wastes. Both of these unit operations were advanced to accommodate shortcomings in glass formulation for HLW inventories. A by-product of this are a series of technical challenges placed upon materials selected for the processing vessels. The advances in glass formulation play a role in revisiting the flow sheet for the WTP and hence, the unit operations that were being imposed by minimal waste loading requirements set forth in the contract for the design and construction of the plant. Another significant consideration to the most recent revision of the glass models are the impacts on resolution of technical questions associated with current efforts for design completion.

  3. HLW Tank Space Management, Final Report

    International Nuclear Information System (INIS)

    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

  4. HLW Tank Space Management, Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Miller, M.S.; Abell, G.; Garrett, R.; d' Entremont, P.; Fowler, J.R.; Mahoney, M.; Poe, L.

    1999-09-20

    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.

  5. 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.

  6. Synthesis of aluminum nitride nanoparticles by a facile urea glass route and influence of urea/metal molar ratio

    International Nuclear Information System (INIS)

    Attention toward nanosized aluminum nitride (AlN) was rapidly increasing due to its physical and chemical characteristics. In this work, nanocrystalline AlN particles were prepared via a simple urea glass route. The effect of the urea/metal molar ratio on the crystal structure and morphology of nanocrystalline AlN particles was studied using X-ray powder diffraction (XRD), scanning electron microscope (SEM) and transmission electron microscope (TEM). The results revealed that the morphology and the crystal structure of AlN nanoparticles could be controlled by adjusting the urea/metal ratio. Furthermore, a mixture of Al2O3 and h-AlN was detected at the urea/metal molar ratio of 4 due to the inadequate urea content. With increasing the molar ratio, the pure h-AlN was obtained. In addition, the nucleation and growth mechanisms of AlN nanocrystalline were proposed.

  7. Modeling of Spinel Settling in Waste Glass Melter

    Energy Technology Data Exchange (ETDEWEB)

    Hrma, Pavel; Schill, Petr; Nemec, Lubomir; Klouzek, Jaroslav, Mika, Martin; Brada, Jiri Glass Service, Ltd., Vsetin, Czech Republic

    2000-06-01

    Our objective is to determine the fraction and size of spinel crystals in molten HLW glass that are compatible with low-risk melter operation. To this end, we are investigating spinel behavior in HLW glass and obtaining data to be used in a mathematical model for spinel settling in a HLW glass melter. We will modify the current glass-furnace model to incorporate spinel concentration distribution and to predict the rate of spinel settling. Also, we will determine the nucleation agents that control the number density and size of spinel crystals in HLW glass.

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

    International Nuclear Information System (INIS)

    Al-free glasses of general composition 0.340SiO2:0.300ZnO:(0.250-a-b)CaO:aSrO:bMgO:0.050Na2O:0.060P2O5 (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 tuned to

  9. 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

  10. Mass Attenuation Coefficients and Effective Atomic Numbers of Thermoluminescent Aluminum Oxide Based Glasses

    International Nuclear Information System (INIS)

    The photon mass attenuation coefficient of a newly prepared 15Al2O3-35P2O5- xCaO-(50-x)Na2CO3 glass system (symbolized as APCN), where x=5, 10, 15, 20, 25, 30, 35, 40 all in mol%, have been calculated at photon energies of 0.662 MeV (137Cs source) and 1.25 MeV (60Co source). In addition, the photon mass attenuation coefficient of 15Al2O3-35P2O5-25CaO-25Na2CO3 glass system (symbolized as APCN25-25), all in mol%, doped with different concentrations of SiO2 have been calculated. The WinXCOM software program on the basis of mixture rule was utilized in calculations. The total atomic (σt) and electronic (σe) cross sections, effective atomic number (Zeff) and electron density (Nel) were calculated. The results showed that the total mass attenuation coefficient showed an extremely dependence on incoherent scattering processes where it varies with Na2CO3 contents in the APCN composition while changing the concentrations of SiO2 in APCN25-25 glass showed slight changes in the values. Otherwise, the mass attenuation coefficient (µm) had higher values at 0.662 MeV than those of 1.25 MeV in both APCN and APCN25-25 glass systems. The values of Zeff showed a decrease with increasing Na2CO3 contents in the APCN composition. The should highly be considered in dealing with such prepared APCN glass system as a gamma ray detector, specially as thermoluminescence dosimeter.

  11. FINAL REPORT. SETTLING OF SPINEL IN A HIGH-LEVEL WASTE GLASS MELTER

    Science.gov (United States)

    The research on spinel settling in the high-level waste (HLW) glass melter was conducted to assist HLW retrieval, glass formulation, feed preparation, and melter design and operation for minimum-risk and minimum-cost vitrification of HLW at Hanford and other DOE sites. The main r...

  12. Microstructure of Aluminum/Glass Joint Bonded by Ultrasonic Wire Welding

    Science.gov (United States)

    Iwamoto, Chihiro

    2013-11-01

    An Al/glass joint created by using ultrasonic welding was analyzed by means of multiscale observation techniques. A cross-sectional analysis of the microstructure revealed that a directly joined interface without reaction phases formed at the periphery of a round joined region. The size of Al grains markedly decreased after ultrasonic welding and some subgrains were observed along the interface. The finer Al grains observed around the periphery of the joined interface showed active plastic flow that promoted welding.

  13. MODELING OF SPINEL SETTLING IN WASTE GLASS MELTER

    Science.gov (United States)

    The topic of this multi-institutional bi-national research is the formation and settling of spinel, the most common crystalline phase that precipitates in molten high-level waste HLW) glass. For the majority of HLW streams, spinel formation in the HLW melter limits the waste fra...

  14. Effect of ZnO and CaO on Alkali Borosilicate Glass Waste-form Immobilizing Simulated Mixed HLW%ZnO 和 CaO对模拟高放废液硅酸盐玻璃固化体性能的影响研究

    Institute of Scientific and Technical Information of China (English)

    张华; N.C.Hyatt; J.R.Stevens; R.Hand

    2015-01-01

    针对有些高放废液含有较多Fe、Cr、Ni过渡金属元素,在玻璃固化工艺过程中易于形成晶体,导致熔融玻璃体的黏度增加、化学稳定性变差以及工艺过程中易出现出料口堵塞等问题,研究了废物包容量为15%和20%、添加ZnO (5.6%)和CaO (1.75%)的配方对形成的4种玻璃固化体的物理性能(密度、硬度、断裂韧性)、化学性能(产品一致性测试和蒸汽腐蚀测试)和结构(X射线衍射析晶分析、拉曼光谱分析)的影响。研究分析显示,提高废物包容量至20%以及添加ZnO和CaO均可促进硼硅酸盐玻璃固化体网络结构的稳定性和化学稳定性,并增强玻璃体的密度,提高硬度;但玻璃固化体的高温黏度升高,断裂韧性下降。%Since the transit metals ,such as Fe ,Cr and Ni ,contained in some kinds of mixed HLW ,can likely to form crystal ,increase the melt viscosity ,destroy the chemi‐cal durability and block the discharge port .T he results obtained from investigating four glass waste‐forms ,including the alkali borosilicate glass matrix and alkali borosilicate glass matrix doped with 5.6% ZnO and 1.75% CaO in base matrixes ,immobilizing the simulated mixed HLW with 15% and 20% waste loadings aiming to determinate the effect of ZnO on the alkali borosilicate glass chemical durability with waste loading increasing ,were presented in this paper .Glass samples were characterized with XRD and Raman spectroscopy .The chemical durability was investigated using the standard protocols PCT and VHT .The XRD analysis results show that spinel crystal appears and grows in glass samples at the waste loading in 20% without ZnO addition and waste loading in 15% and 20% added ZnO .T he Raman spectroscopy analysis results indicate that ZnO and CaO can enhance the glass network connective ,and the chemical durability test results display that the addition of ZnO and CaO can improve the short term

  15. 12 Flasktransport of vitrified High Level Waste (HLW)

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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 m2 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. 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.

  18. Replacement of glass in the Nakhla meteorite by berthierine: Implications for understanding the origins of aluminum-rich phyllosilicates on Mars

    Science.gov (United States)

    Lee, Martin R.; Chatzitheodoridis, Elias

    2016-07-01

    A scanning and transmission electron microscope study of aluminosilicate glasses within melt inclusions from the Martian meteorite Nakhla shows that they have been replaced by berthierine, an aluminum-iron serpentine mineral. This alteration reaction was mediated by liquid water that gained access to the glasses along fractures within enclosing augite and olivine grains. Water/rock ratios were low, and the aqueous solutions were circumneutral and reducing. They introduced magnesium and iron that were sourced from the dissolution of olivine, and exported alkalis. Berthierine was identified using X-ray microanalysis and electron diffraction. It is restricted in its occurrence to parts of the melt inclusions that were formerly glass, thus showing that under the ambient physico-chemical conditions, the mobility of aluminum and silicon were low. This discovery of serpentine adds to the suite of postmagmatic hydrous silicates in Nakhla that include saponite and opal-A. Such a variety of secondary silicates indicates that during aqueous alteration compositionally distinct microenvironments developed on sub-millimeter length scales. The scarcity of berthierine in Nakhla is consistent with results from orbital remote sensing of the Martian crust showing very low abundances of aluminum-rich phyllosilicates.

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

    International Nuclear Information System (INIS)

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

  20. 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)

    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.

  1. 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.

  2. Glasses

    DEFF Research Database (Denmark)

    Dyre, Jeppe

    2004-01-01

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

  3. 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.

  4. HLW system plan - revision 2

    International Nuclear Information System (INIS)

    The projected ability of the Tank Farm to support DWPF startup and continued operation has diminished somewhat since revision 1 of this Plan. The 13 month delay in DWPF startup, which actually helps the Tank Farm condition in the near term, was more than offset by the 9 month delay in ITP startup, the delay in the Evaporator startups and the reduction to Waste Removal funding. This Plan does, however, describe a viable operating strategy for the success of the HLW System and Mission, albeit with less contingency and operating flexibility than in the past. HLWM has focused resources from within the division on five near term programs: The three evaporator restarts, DWPF melter heatup and completion of the ITP outage. The 1H Evaporator was restarted 12/28/93 after a 9 month shutdown for an extensive Conduct of Operations upgrade. The 2F and 2H Evaporators are scheduled to restart 3/94 and 4/94, respectively. The RHLWE startup remains 11/17/97

  5. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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. (papers)

  9. Effect of aluminum closed-cell foam filling on the quasi-static axial crush performance of glass fiber reinforced polyester composite and aluminum/composite hybrid tubes

    OpenAIRE

    Güden, Mustafa; YÜKSEL, Sinan; Taşdemirci, Alper; Tanoğlu, Metin

    2007-01-01

    The effect of Al closed-cell foam filling on the quasi-static crushing behavior of an E-glass woven fabric polyester composite tube and thin-walled Al/polyester composite hybrid tube was experimentally investigated. For comparison, empty Al, empty composite and empty hybrid tubes were also tested. Empty composite and empty hybrid tubes crushed predominantly in progressive crushing mode, without applying any triggering mechanism. Foam filling was found to be ineffective in increasing the crush...

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

    International Nuclear Information System (INIS)

    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

  11. 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.

  12. 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

  13. 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)

    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 m2) 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/m2/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 WVDP

  14. Formation and Compression Behavior of Two-Phase Bulk Metallic Glasses with a Minor Addition of Aluminum

    Institute of Scientific and Technical Information of China (English)

    ZONG Hai-Tao; MA Ming-Zhen; ZHANG Xin-Yu; QI Li; LI Gong; JING Qin; LIU Ri-Ping

    2011-01-01

    A remarkable enhancement in room-temperature compressive deformability is realized by the minor-addition of 1.5 at. % Al in ZrTi-based bulk metallic glass.Two amorphous phases are observed by transmission electron microscopy in the Al-containing alloys and this explains the improvement of compression deformability. The studies suggest that phase separation might occur in glass forming alloys with a negative enthalpy of mixing.

  15. Sintered glass

    International Nuclear Information System (INIS)

    The authors present an overview of various attempts to sinter high-level matrix. This paper focuses on the development of the porous glass matrix process (PGM process) for the fixation of HLW in a high-silica glass matrix by sintering. In the PGM process a borosilicate-type base glass is phase-separated by heat treatment. The low-silica phase is leached in HC1 and washed out. The remaining fine powder is soaked with HLW, dried under vacuum by gradual heating up to 1123 K, ground or ball-milled and finally sintered at 1473 K to a dense solid, consisting of a matrix of 96% SiO2 and 4% B2O3 in which the waste elements are trapped. The main advantage of the process is the high chemical durability of the final product. The rather complicated process technology, also leading to the generation of secondary waste (e.g. washing solution), is considered a disadvantage

  16. Update on spent fuel and HLW management in Spain

    International Nuclear Information System (INIS)

    Full text: According to strategy adopted in the fifth General Radioactive Waste Management Plan (GRWMP) of 1999, ENRESA, the agency responsible for radioactive waste management in Spain, has implemented, together with the NPP operators, different spent fuel storage technologies to manage the overall inventory (around 7,000 tU) expected to be generated in the country. In the nineties, all the spent fuel NPP pools were re-racked as an initial stage to provide additional At-Reactor capacity. The next steps were, and will be directed, towards using dry spent fuel storage technologies. In this latter respect, the major milestones achieved and plans are the following ones: - For the first plant requiring additional out of pool capacity, the TRILLO NPP (KWU design), a dual-purpose cask known as the DPT cask, was designed and licensed in Spain. The two first units have already been loaded in 2002 and a campaign of 6 more loads is scheduled to take place in 2003; - A massive Cask Storage Building has been constructed in the site to host the 80 casks that will be needed during the expected plant's operating lifetime. The startup licence was issued in May 2002; - The JOSE CABRERA (ZORITA) NPP will stop its operation by April 2006, three years ahead of its original schedule. As a part of the decommissioning activities, the spent fuel stored in the pool will be evacuated to a dry storage system. ENRESA is currently studying the possible alternatives to manage the total spent fuel inventory (almost 400 fuel assemblies). An Away-From-Reactor Spent Fuel and HLW Centralised Storage Installation is foreseen to be available by 2010. It will collect all the spent fuel and some medium and high active wastes generated in the country, together with glasses and other HLW that will come back to Spain from previous reprocessing contracts. Finally, different studies and demonstration projects are being developed related to the deep geological disposal. (author)

  17. 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.

  18. 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.

  19. Strategic management of HLW repository projects

    International Nuclear Information System (INIS)

    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

  20. Nanodispersion formation as a universal subcritical property of a substance on its path from homogeneous to microheterogeneous state. Glass transitions

    International Nuclear Information System (INIS)

    Glass formation transitions interpreted as second-kind phase transitions were considered as critical phenomena as applied to the problem of high-level waste (HLW) vitrification. Certain regularities of HLW vitrification were explained from the viewpoint mentioned. Specifically, increase in dissolving ability of the vitrified systems during glass transitions and reduction to zero of diffusion coefficients are predicted

  1. 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.

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

    International Nuclear Information System (INIS)

    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.

  3. 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.

  4. Influence of Y, Gd and Sm on the glass forming ability and thermal crystallization of aluminum based alloy

    International Nuclear Information System (INIS)

    Al-based amorphous alloys represent an important family of metals and a great scientific activity has been devoted to determine the main features of both glass forming ability (GFA) and crystallization behavior in order to have a comprehensive framework aimed at potential technological applications. Nowadays, it is well known that the best Al-based amorphous alloys are formed in ternary systems such as Al- RE-TM, where RE is a rare earth and TM a transition metal. This paper presents results of research in Al85Ni10RE5 alloys (RE = Y, Gd and Sm). Amorphous ribbons were processed by melt-spinning under the same conditions and subsequently characterized by x-ray diffraction (XRD) and differential scanning calorimetry (DSC). Results show appreciable micro structural differences as function of the rare earth, thus crystal is obtained for Y, nano-glassy for Gd and, fully amorphous structure for Sm. (author)

  5. Cytotoxicity of glass ionomer cement on human exfoliated deciduous teeth stem cells correlates with released fluoride, strontium and aluminum ion concentrations

    Directory of Open Access Journals (Sweden)

    Kanjevac Tatjana V.

    2015-01-01

    Full Text Available Stem cells from human exfoliated deciduous teeth (SHED can be used as a cell-based therapy in regenerative medicine and in immunomodulation. Pulp from human deciduous teeth can be stored as a source of SHED. Glass ionomer cements (GICs are commonly used in restorative dentistry and in cavity lining. GICs have lower biocompatibility and are cytotoxic for dental pulp cells. In this study, seven commonly used GICs were tested for their cytotoxic effects on SHED, for their potential to arrest mitosis in cells and induce chromosome aberrations, and were compared with the effects of composite. Fuji II, Fuji VIII, Fuji IX, Fuji plus and Vitrebond had significantly higher cytotoxic effects on SHED than composite. Only SHEDs that have been treated with Fuji I, Fuji IX, Fuji plus and composite recovered the potential for proliferation, but no chromosome aberrations were found after treatment with GICs. The cytotoxic effects of GICs on SHEDs were in strong correlation with combined concentrations of released fluoride, aluminum and strontium ions. Fuji I exhibited the lowest activity towards SHEDs; it did not interrupt mitosis and did not induce chromosome aberrations, and was accompanied by the lowest levels of released F, Al and Sr ions. Projekat Ministarstva nauke Republike Srbije, br. ON175069, br. ON175071 i br. ON175103

  6. Development and adoption of low sodium glass frit for vitrification of high level radioactive liquid waste at Tarapur

    International Nuclear Information System (INIS)

    High level Liquid Waste (HLW) is generated during the reprocessing of spent nuclear fuel which is used to recover uranium and plutonium. More than 99% of the fission product activity generated during the burning of nuclear fuel in the reactor is present in HLW. For the efficient management of HLW by vitrification, sodium borosilicate glass has been adopted worldwide. Sodium oxide acts as modifier in glass matrix and variation in its concentration may vary the properties of the glass and hence the melter parameters. The HLW presently used for vitrification has higher concentration of sodium. As the composition of the base glass is fixed the concentration of Na in the HLW is one of the limiting factors for the waste loading for the vitrification process. Present article gives a brief account of the formulation of a base glass frit with lower sodium content and the feedback after implementing in the vitrification plant. (author)

  7. 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.

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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. HIALI, the hazard index based on ALI, due to radiotoxicity of Group R can be lower than HIALI due to standard mill tailing (smt) or uranium ore after about 300 years. (author)

  10. Substrate biasing effect on the physical properties of reactive RF-magnetron-sputtered aluminum oxide dielectric films on ITO glasses.

    Science.gov (United States)

    Liang, Ling Yan; Cao, Hong Tao; Liu, Quan; Jiang, Ke Min; Liu, Zhi Min; Zhuge, Fei; Deng, Fu Ling

    2014-02-26

    High dielectric constant (high-k) Al2O3 thin films were prepared on ITO glasses by reactive RF-magnetron sputtering at room temperature. The effect of substrate bias on the subband structural, morphological, electrode/Al2O3 interfacial and electrical properties of the Al2O3 films is systematically investigated. An optical method based on spectroscopic ellipsometry measurement and modeling is adopted to probe the subband electronic structure, which facilitates us to vividly understand the band-tail and deep-level (4.8-5.0 eV above the valence band maximum) trap states. Well-selected substrate biases can suppress both the trap states due to promoted migration of sputtered particles, which optimizes the leakage current density, breakdown strength, and quadratic voltage coefficient of capacitance. Moreover, high porosity in the unbiased Al2O3 film is considered to induce the absorption of atmospheric moisture and the consequent occurrence of electrolysis reactions at electrode/Al2O3 interface, as a result ruining the electrical properties. PMID:24490685

  11. Microstructure and mechanical properties of aluminum alloy matrix composites reinforced with Fe-based metallic glass particles

    International Nuclear Information System (INIS)

    Highlights: • Al-2024/FMG composites have been prepared by powder metallurgy method. • Mechanical milling resulted in significant grain refinement of the Al matrix. • The high strength is attributed to the refined microstructure and FMG particles. - Abstract: Fe-based metallic glass (FMG) particles reinforced Al-2024 matrix composites were fabricated by using the powder metallurgy method successfully. Mechanical alloying result in nanostructured Al-2024 matrix with a grain size of about 30 nm together with a good distribution of the FMG particles in the Al matrix. The consolidation of the composites was performed at a temperature in the super-cooled liquid region of the FMG particles, where the FMG particles act as a soft liquid-like binder, resulting in composites with low or zero porosity. The microstructure and mechanical properties of the composites were investigated by X-ray diffraction (XRD), scanning electron microscopy (SEM), transmission electron microscopy (TEM) and compression test. The yield and fracture strength of the composites are 403 MPa and 660 MPa, respectively, while retaining a considerable fracture deformation of about 12%. The strengthening mechanism is associated with the grain refinement of the matrix and uniform distribution of the FMG particles

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

    International Nuclear Information System (INIS)

    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)

  13. 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)

  14. HLW Long-term Management Technology Development

    International Nuclear Information System (INIS)

    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

  15. 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)

    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

  16. 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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  19. Experimental Plan for the Cold Demonstration (Scoping Tests) of Glass Removal Methods from a DWPF Melter

    International Nuclear Information System (INIS)

    SRS and WVDP currently do not have the capability to size reduce, decontaminate, classify, and dispose of large, failed, highly contaminated equipment. Tanks Focus Area Task 777 was developed to address this problem. The first activity for Task 777 is to develop and demonstrate techniques suitable for removing the solid HLW glass from HLW melters. This experimental plan describes the work that will be performed for this glass removal demonstration

  20. The Preparation and Characterization of INTEC Phase 2b Composition Variation Study Glasses

    Energy Technology Data Exchange (ETDEWEB)

    B. A. Staples; B. A. Scholes; L. L. Torres; C. A. Musick; B. R. Boyle (INEEL); D. K. Peeler (SRTC); J. D. Vienna (PNNL)

    2000-02-01

    The second phase of the composition variation study (CVS) for the development of glass compositions to immobilize Idaho Nuclear Technology and Engineering Center (INTEC) high level wastes (HLW) is complete. This phase of the CVS addressed waste composition of high activity waste fractions (HAW) from the initial separations flowsheet. Updated estimates if INTEC calcined HLW compositions and of high activity waste fractions proposed to be separated from dissolved calcine were used as the waste component for this CVS phase. These wastes are of particular interest because high aluminum, calcium, zirconium, fluorine, potassium, and low iron and sodium content places them outside the vitrification experience in the Department of Energy complex. Because of the presence of calcium and fluorine, two major zirconia calcine components not addressed in Phase I, a series of scooping tests, designated Phase 2a, were performed. The results of these tests provided information on the effects of calcium and fluoride solubility and their impacts on product properties and composition boundary information for Phase 2b. Details and results of Phase 2a are reported separately. Through application of statistical techniques and the results of Phase 2a, a test matrix was defined for Phase 2b of the CVS. From this matrix, formulations were systematically selected for preparation and characterization with respect to visual and optical homogeneity, viscosity as a function of melt temperature, liquidus temperature (TL), and leaching properties based on response to the product consistency test. The results of preparing and characterizing the Phase 2b glasses are presented in this document. Based on the results, several formulations investigated have suitable properties for further development. A full analysis of the composition-product characteristic relationship of glasses being developed for immobilizing INTEC wastes will be performed at the completion of composition-property relationship

  1. 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.

  2. Suitability of pressed-fired Alumina Zirconia Silica (AZS) as backup refractory in joule heated ceramic melter for vitrification of HLW at AVS, Tarapur

    International Nuclear Information System (INIS)

    Vitrification of high level liquid waste (HLW) by single step joule heated ceramic melter (JHCM) based technology developed indigenously and operated successfully for vitrification of HLW generated during reprocessing of spent fuel. In a ceramic melter glass is contained in a ceramic lined furnace. The glass-contact refractory is designed with indigenous Alumina-Zirconia-Silica (AZS) refractory. Bubble alumina is used as backup refractory to reduce molten glass migration. Backup refractory, as the name suggests is the second line of thermally and chemically stable material which supports the primary glass-contact refractory. Acidic vapour and/or molten vitreous mass can seep through the joints of blocks of primary refractory. As a result, acidic vapour and/or molten vitreous mass can come in contact with backup refractory. Bubble alumina has not shown adequate durability in nitric acid/molten vitreous mass because of carbonaceous material used as binder. Therefore rearrangement of suitable backup refractory up to the glass pool level in contact with primary glass-contact refractory is desired to increase the path-length of the molten mass before it reaches castable. In addition, backup refractory should have good durability with respect to HNO3 and molten glass. Here, corrosive environment is not that aggressive in terms of temperature of vitreous mass (800-850 deg C) and concentration of nitric acid (3

  3. GLASS SELECTION STRATEGY: DEVELOPMENT OF US AND KRI TEST MATRICIES

    International Nuclear Information System (INIS)

    High-level radioactive wastes are stored as liquids in underground storage tanks at the Department of Energy's (DOE) Savannah River Site (SRS) and Hanford Reservation. These wastes are to be prepared for permanent disposition in a geologic repository by vitrification with glass forming additives (e.g., frit), creating a waste form with long-term durability. Wastes at SRS are being vitrified in the Defense Waste Processing Facility (DWPF). Vitrification of the wastes stored at Hanford is planned for the Waste Treatment and Immobilization Plant (WTP) when completed. Some of the wastes at SRS, and particularly those at Hanford, contain high concentrations of aluminum, chromium and sulfate. These elements make it more difficult to produce a waste glass with a high waste loading (WL) without crystallization occurring in the glass (either within the melter or upon cooling of the glass), potentially exceeding the solubility limit of critical components, having negative impacts on durability, and/or resulting in the formation of a sulfate salt layer on the molten glass surface. Although the overall scope of the task is focused on all three critical, chemical components, the current work will primarily address the potential for crystallization (e.g., nepheline and/or spinel) in high level waste (HLW) glasses. Recent work at the Savannah River National Laboratory (SRNL) and by other groups has shown that nepheline (NaAlSiO4), which is likely to crystallize in high-alumina glasses, has a detrimental effect on the durability of the glass. The objective of this task is to develop glass formulations for specific SRS and Hanford waste streams to avoid nepheline formation while meeting waste loading and waste throughput expectations, as well as satisfying critical process and product performance related constraints. Secondary objectives of this task are to assess the sulfate solubility limit for the DWPF composition and spinel settling for the WTP composition. SRNL has partnered

  4. High level radioactive waste (HLW) disposal a global challenge

    CERN Document Server

    PUSCH, R; NAKANO, M

    2011-01-01

    High Level Radioactive Waste (HLW) Disposal, A Global Challenge presents the most recent information on proposed methods of disposal for the most dangerous radioactive waste and for assessing their function from short- and long-term perspectives. It discusses new aspects of the disposal of such waste, especially HLW.The book is unique in the literature in making it clear that, due to tectonics and long-term changes in rock structure, rock can serve only as a ""mechanical support to the chemical apparatus"" and that effective containment of hazardous elements can only be managed by properly des

  5. 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)

    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 m2) 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.

  6. 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.

  7. Preparation, characterization and standard molar enthalpy of formation of BaO containing sodium borosilicate glasses and its comparison with international standard glass

    International Nuclear Information System (INIS)

    High level radioactive liquid waste (HLW) generated during reprocessing of spent nuclear fuel is immobilized in sodium borosilicate (NBS) glasses. Addition of BaO in NBS glass helps to improve the solubility of ThO2 in glass matrix. The knowledge of thermodynamic stability of glasses used for immobilization of HLW is important in predicting their long term stability. Several BaO substituted NBS glass samples were prepared by melt-quench technique and characterized by XRD, DTA, MAS-NMR. The standard molar enthalpy of formation of BaO substituted NBS glasses and the International Standard Glass (ISG) were determined. This work is done with an understanding that even though the above glass matrices are metastable in nature and meaningful measurement of equilibrium thermodynamic data is difficult; the information on relative thermodynamic stability data of NBS glasses with varying compositions prepared exactly in similar fashion will be helpful in deciding the most stable matrix for nuclear waste disposal

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

    International Nuclear Information System (INIS)

    '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.'

  9. FLOWSHEET FOR ALUMINUM REMOVAL FROM SLUDGE BATCH 6

    Energy Technology Data Exchange (ETDEWEB)

    Pike, J; Jeffrey Gillam, J

    2008-12-17

    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

  10. Synthesis and thermophysical property measurements on various types of glasses for nuclear waste immobilization

    International Nuclear Information System (INIS)

    Borosilicate glasses (BSG) are worldwide known host matrices for immobilization of radioactive High Level Waste (HLW). Different types of borosilicate glasses were prepared by changing the modifier concentrations and compositions to know the efficacy of the resulting glass in terms of glass formation, durability towards various waste elements, stability at higher temperatures, mobility of ionic species etc. towards nuclear applications. In this study BSG, Aluminium borosilicate glass (AlBSG), Barium borosilicate glass (BaBSG) and Lead borosilicate glasses (PbBSG) were prepared and characterised to confirm the glass formations. Percentage linear thermal expansion and glass transition temperatures were measured by dilatometric techniques

  11. MAGNOX:BUTEX URANIUM BEARING GLASSES PHYSICAL AND CHEMICAL ANALYSIS DATA PACKAGE

    Energy Technology Data Exchange (ETDEWEB)

    Peeler, D.; Imrich, K.; Click, D.

    2011-03-08

    applied) of 1050 C. Any glass which has a T{sub L} predicted value > 1050 C would be classified as unacceptable and the SME product would not be transferred to the melter. As another example, consider durability (as defined by the PCT test) and its related constraints to determine acceptability. If the glass composition yields predicted normalized release values that exceed those associated with the Environmental Assessment (EA) glass (with uncertainties applied) then the glass is deemed unacceptable. The issue of acceptability plays a critical role in high level waste processing but without knowing the pre-defined constraints for the UK HLW system, assessments of acceptability of the glasses to be characterized in this study can not be made. The results of this study will be compared to DWPF constraints to provide a benchmark for determining acceptability. The objective of this task is to provide Sellafield Ltd. with the technical data to evaluate the impacts of various Magnox:Butex blend ratios and WLs on key glass properties of interest. The uranium bearing glasses span a compositional region of interest to Sellafield Ltd. and were physically characterized for key processing and product performance properties as defined in the WFO [WFO (2010)]. One of the specific technical issues (as defined in the WFO) is the potential impact of increasing aluminum concentrations on key properties (in particular viscosity).

  12. Glasses for nuclear waste immobilization

    International Nuclear Information System (INIS)

    Vitrification of nuclear wastes is attractive because of its flexibility, the large number of elements which can be incorporated in the glass, its high corrosion durability and the reduced volume of the resulting waste form. Vitrification is a mature technology and has been used for high level nuclear waste (HLW) immobilisation for more than 40 years in France, Germany and Belgium, Russia, UK, Japan and the USA. Vitrification involves melting of waste materials with glass-forming additives so that the final vitreous product incorporates the waste contaminants in its macro- and micro-structure. Hazardous waste constituents are immobilised either by direct incorporation into the glass structure or by encapsulation when the final glassy material can be in form of a glass composite material (GCM). Both borosilicate and phosphate glasses are currently used to immobilise nuclear wastes, moreover in addition to relatively homogeneous glasses novel GCM are used to immobilise problematic waste streams. The spectrum of wastes which are currently vitrified increases from HLW to low and intermediate wastes (LILW) such as legacy wastes in Hanford, USA and nuclear power plant operational wastes in Russia and Korea. (authors)

  13. Glass, Ceramics, and Composites

    International Nuclear Information System (INIS)

    Many studies of plutonium in glass and ceramics have taken place in the thirty years covered by this book. These studies have led to a substantial understanding, arising from fundamental research of actinides in solids and research and development in three technical fields: immobilization of the high level wastes (HLW) from commercial nuclear power plants and processing of nuclear weapons materials, environmental restoration in the nuclear weapons complex and, most recently, the immobilization of weapons-grade plutonium as a result of disarmament activities

  14. Systems approach to developing and implementing an HLW licensing process

    International Nuclear Information System (INIS)

    Developing and licensing a repository for the disposal of high-level nuclear waste (HLW) is an important national goal. Congress reaffirmed this goal by the passage of the Nuclear Waste Policy Amendments Act (NWPAA) in December 1987. In addition to designating Yucca Mountain in Nevada as the sole site to undergo characterization, it authorized a monitored retrievable storage facility (MRS) and created a negotiator, a technical review board, and other entities whose efforts are directed to provide greater certainty in the nation's ability to manage its HLW. Responsibility for the formal licensing process for the life of the repository is assigned by law to the US Nuclear Regulatory Commission (NRC). In October of 1987, the NRC established the Center for Nuclear Waste Regulatory Analysis as a federally funded research and development center to assist in fulfilling its licensing mission with regard to HLW. The broad objective of the center is to assist the NRC in identifying and recommending solutions for technical, regulatory, and institutional uncertainties to ensure timely and credible completion of the licensing process. To meet this objective, the center is currently developing a systems approach to implement and streamline the HLW licensing process. This paper describes the basic systems approach and the design process and provides a status on the initial efforts of the center to develop this system

  15. Disposal concepts for HLW and TRU waste in France

    International Nuclear Information System (INIS)

    In France, the policy for long term radioactive waste management calls for isolation of high- level vitrified waste (HLW) and transuranic waste (TRU) in deep continental geological formations. Four potential sites are at present being investigated with the objective of selecting a candidate site by the end of 1990. After several years of in-depth investigation through an underground site validation laboratory (1991-1995) approval procedure (1995-1996) and construction work (1997-2000) start up of the repository is scheduled in 2001 for TRU waste and 2010 for HLW. This paper reports several disposal concepts are envisaged for TRU and HLW. In the crystalline formations (granite, schist) the vault concept is the preferred one for TRU waste, although the borehole concept seems well adapted to HLW. In the sedimentary formations (clay, salt), the borehole concept is suitable for any type of waste and the silo concept in salt is considered as an alternative concept for TRU waste. Final choice of disposal concepts will be made for each geological formation once geotechnical and hydrogeological information is available from the surface drilling program and that safety assessments on the candidate scenarios are carried out

  16. Development of glass dissolution database

    International Nuclear Information System (INIS)

    The dissolution rate of radionuclide from radioactive waste (HLW) glass is one of the important parameters for the safety assessment of the geological disposal system. A lot of information on the dissolution behavior of the glass are opened to the public and provide useful information to understand the essential characteristic of the glass under various conditions. We developed the database with collection of information on the dissolution behavior of the glass (Glass, Experimental Type, Solution, Experimental Conditions, Experimental Results, Reference) from Scientific Basis for Nuclear Waste Management published by Materials Research Society (MRS). This system contains 846 data now. This database was made by Windows Me/Access 2002 to search of data, reference document, and graphs. (author)

  17. 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 ...

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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 (SiO2 :B2O3 :Na2O : MnO : TiO2) 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)

  20. Mechanical properties of nuclear waste glasses

    International Nuclear Information System (INIS)

    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.

  1. Transport of HLW and spent fuel in Japan

    International Nuclear Information System (INIS)

    In Japan, transport of HLW started in spring, 1995 when vitrified waste was returned to Japan from Europe. Domestic transport of spent fuel has been carried out since 1978 from domestic nuclear power plant to the reprocessing plant in Tokai-mura, Ibaraki Prefecture, amounting to some 720 tons by the end of 1994. A commercial reprocessing plant is under construction in Rokkasho-mura, Aomori Prefecture and the transport thereto of high burn-up spent fuel is scheduled to start in 1997. A 150-ton wharf crane and a dedicated transport vehicle were completed in 1994 for unloading and overland transport of HLW casks, and specifically designed casks and an exclusive-use vessel are being prepared for domestic transport of high burn-up spent fuel

  2. Conceptual design requirements for Korean Reference HLW disposal System

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui Joo; Choi, Jong Won; Hahn, Pil Son; Lee, Jong Youl; Kim, Kyung Soo; Kim, Sung Ki; Cho, Dong Keun; Lee, Yang

    2005-05-15

    This report outlined the requirements for the conceptual design of KRS(Korean Reference HLW disposal System). The site for the disposal of high-level radioactive wastes has not yet been selected in Korea. Since the KRS should be designed under these circumstances, the necessary requirements which should be determined are studied in the report. The amounts of spent fuels from the nuclear power plants in the long-term national power development plan are projected. With this estimation the disposal rates of CANDU and PWR spent fuels are analyzed and determined. The national and international regulations regarding the disposal of HLW are summarized. The functions of the underground facilities are defined. The representative geological conditions are determined since no site is yet decided in Korea.

  3. 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

  4. Proposed review of current regulatory safety criteria for the HLW

    International Nuclear Information System (INIS)

    In line with a general international trend to move for many high risk activities to a more risk-informed management of the underlying hazards involving all stakeholders, the issue of how to provide a high degree of transparency in order to ensure commensurability and consistency in the approaches is high on the research agenda. This paper considers the case of assessing the risks from a specific 'risk activity', i.e. the long-term storage of high level radioactive waste (HLW) in a deep geological formation. Further, in order to provide consistent and systematic insights in the similarities and main differences among the existing different technical approaches on how to assess HLW risks, this paper proposes to: - develop a generic template that allows to capture, map and communicate in a consistent way the existing large variety of current approaches, - and to perform on the basis of this template a systematic review of current HLW risk assessment approaches as recommended or prescribed in various countries. If successful, the results of the proposed work would not say whether the risks originating from specific nuclear waste repositories are acceptable or not, but contribute: - to better communication of these risks among all stakeholders; and - to their more consistent mapping and cross comparison. (authors)

  5. Microbial activity in a deep underground HLW disposal cell

    International Nuclear Information System (INIS)

    The research program conducted by IRSN on microbial activity in the context of deep underground radioactive waste disposal is focused on the potential impact of bacteria on the corrosion of carbon steel materials involved in the design of high level waste (HLW) disposal cells developed in France by Andra. This program combines two different approaches: firstly, the evaluation of biodiversity of argillite, from undisturbed and disturbed (gallery wall, excavation damaged zone, presence of metallic platelets..) samples and secondly, the development of a conceptual model for development of microbial activity in a HLW disposal cell based on mass and energy balances. At this stage, it is worth noting that the effects of high radiation have not been taken into account. The characterization of biodiversity of the Toarcian argillite of Tournemire (experimental tunnel operated to develop IRSN skills in earth sciences) has shown that even if the argillite probably acts as a natural selective substratum for bacterial colonization, exogenous microorganisms may develop within disturbed areas. Indeed, the observed bacterial diversity tends to depend on the different oxygen and humidity conditions, and also probably on space availability. This characterization also highlighted the presence of a sulphate-reducing and iron-reducing bacterium capable to resist to high temperatures in a sample collected by scratching a steel platelet. Besides, the preliminary conceptual model of bacterial development has shown that iron-reducing and sulphate-reducing bacteria may be able to grow in the environment of HLW disposal cell. (orig.)

  6. R and D on HLW partitioning at Khlopin Radium Institute

    International Nuclear Information System (INIS)

    Results of more than thirty-years investigations on high level radioactive wastes (HLW) partitioning at Khlopin Radium Institute (KRI), St-Petersburg are observed and discussed. The objective of research and development is to assess HLW partitioning technical feasibility and advantages over the 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 phosphine oxides, carbamoylphosphineoxides, phosphorus - containing calixarenes, crown ethers, dialkylphosphoric acids, diamides of heterocyclic acids were studied at KRI. Many extraction systems on the base of chlorinated cobalt dicarbollide, including UNEX-extractant and it's modifications, also observed. Diamides of 2,6 pyridinedicarboxylic acid and diamides of 2,2' ,6,6' dipyridyl dicarboxylic acid demonstrates very promising properties for minor actinide-lanthanide separation. Comparison of different solvents and possible ways for implementation of new flowsheets in radiochemical technology is also discussed. (author)

  7. Approach to operational safety in the Japanese HLW repository project

    International Nuclear Information System (INIS)

    The safety of a repository for vitrified high-level radioactive waste (HLW) has two aspects: operational safety of activities during the period extending from construction to closure of the repository and long-term safety following closure of the repository. In Japan, 40,000 waste packages are scheduled to be disposed of in a deep repository over an operational period of approximately 100 years. As the implementer, the Nuclear Waste Management Organization of Japan (NUMO) is responsible for maintaining operational safety during this period. Such operational safety is envisaged as being a key concern on the part of the general public. NUMO adopts a structured approach to developing robust and flexible repository concepts (RC) in a transparent manner, considering the stepwise progress of the HLW repository project in Japan. One of the design factors for development of RC is operational safety. NUMO plans to manage a hierarchical structure of design factors relating to operational safety using a requirements management system. This paper presents the approach to achieving NUMO's goal of operational safety, from construction to closure of the repository, taking into consideration the special features of the HLW disposal project in Japan. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Vince Maio

    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.

  9. Aluminum Nanoholes for Optical Biosensing

    Directory of Open Access Journals (Sweden)

    Carlos Angulo Barrios

    2015-07-01

    Full Text Available Sub-wavelength diameter holes in thin metal layers can exhibit remarkable optical features that make them highly suitable for (biosensing applications. Either as efficient light scattering centers for surface plasmon excitation or metal-clad optical waveguides, they are able to form strongly localized optical fields that can effectively interact with biomolecules and/or nanoparticles on the nanoscale. As the metal of choice, aluminum exhibits good optical and electrical properties, is easy to manufacture and process and, unlike gold and silver, its low cost makes it very promising for commercial applications. However, aluminum has been scarcely used for biosensing purposes due to corrosion and pitting issues. In this short review, we show our recent achievements on aluminum nanohole platforms for (biosensing. These include a method to circumvent aluminum degradation—which has been successfully applied to the demonstration of aluminum nanohole array (NHA immunosensors based on both, glass and polycarbonate compact discs supports—the use of aluminum nanoholes operating as optical waveguides for synthesizing submicron-sized molecularly imprinted polymers by local photopolymerization, and a technique for fabricating transferable aluminum NHAs onto flexible pressure-sensitive adhesive tapes, which could facilitate the development of a wearable technology based on aluminum NHAs.

  10. Aluminum Nanoholes for Optical Biosensing

    Science.gov (United States)

    Barrios, Carlos Angulo; Canalejas-Tejero, Víctor; Herranz, Sonia; Urraca, Javier; Moreno-Bondi, María Cruz; Avella-Oliver, Miquel; Maquieira, Ángel; Puchades, Rosa

    2015-01-01

    Sub-wavelength diameter holes in thin metal layers can exhibit remarkable optical features that make them highly suitable for (bio)sensing applications. Either as efficient light scattering centers for surface plasmon excitation or metal-clad optical waveguides, they are able to form strongly localized optical fields that can effectively interact with biomolecules and/or nanoparticles on the nanoscale. As the metal of choice, aluminum exhibits good optical and electrical properties, is easy to manufacture and process and, unlike gold and silver, its low cost makes it very promising for commercial applications. However, aluminum has been scarcely used for biosensing purposes due to corrosion and pitting issues. In this short review, we show our recent achievements on aluminum nanohole platforms for (bio)sensing. These include a method to circumvent aluminum degradation—which has been successfully applied to the demonstration of aluminum nanohole array (NHA) immunosensors based on both, glass and polycarbonate compact discs supports—the use of aluminum nanoholes operating as optical waveguides for synthesizing submicron-sized molecularly imprinted polymers by local photopolymerization, and a technique for fabricating transferable aluminum NHAs onto flexible pressure-sensitive adhesive tapes, which could facilitate the development of a wearable technology based on aluminum NHAs. PMID:26184330

  11. Rhenium volatilization in waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Kai; Pierce, David A. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Hrma, Pavel, E-mail: pavel.hrma@pnnl.gov [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Schweiger, Michael J. [Pacific Northwest National Laboratory, Richland, WA 99352 (United States); Kruger, Albert A. [U.S. Department of Energy, Office of River Protection, Richland, WA 99352 (United States)

    2015-09-15

    Highlights: • Re did not volatilize from a HLW feed until 1000 °C. • Re began to volatilize from LAW feeds at ∼600 °C. • The vigorous foaming and generation of gases from salts enhanced Re evaporation in LAW feeds. • The HLW glass with less foaming and salts is a promising medium for Tc immobilization. - Abstract: We investigated volatilization of rhenium (Re), sulfur, cesium, and iodine during the course of conversion of high-level waste melter feed to glass and compared the results for Re volatilization with those in low-activity waste borosilicate glasses. Whereas Re did not volatilize from high-level waste feed heated at 5 K min{sup −1} until 1000 °C, it began to volatilize from low-activity waste borosilicate glass feeds at ∼600 °C, a temperature ∼200 °C below the onset temperature of evaporation from pure KReO{sub 4}. Below 800 °C, perrhenate evaporation in low-activity waste melter feeds was enhanced by vigorous foaming and generation of gases from molten salts as they reacted with the glass-forming constituents. At high temperatures, when the glass-forming phase was consolidated, perrhenates were transported to the top surface of glass melt in bubbles, typically together with sulfates and halides. Based on the results of this study (to be considered preliminary at this stage), the high-level waste glass with less foaming and salts appears a promising medium for technetium immobilization.

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

    International Nuclear Information System (INIS)

    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

  13. 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.

  14. LIQUIDUS TEMPERATURE OF HIGH-LEVEL WASTE BOROSILICATE GLASSES WITH SPINEL PRIMARY PHASE

    Science.gov (United States)

    Liquidus temperatures (TL) were measured for high-level waste (HLW) borosilicate glasses covering a Savannah River composition region. The primary crystallization phase for most glasses was spinel, a solid solution of trevorite (NiFe2O4) with other oxides (FeO, MnO, and Cr2O3). T...

  15. MILLIMETER-WAVE MEASUREMENTS OF HIGH LEVEL AND LOW ACTIVITY GLASS MELTS

    Science.gov (United States)

    The objectives of the proposed multi-organizational research are to develop new real-time sensors for characterizing glass melts in high level waste (HLW) and low activity waste (LAW) melters, and to understand the scientific basis and bridge the gap between glass melt model data...

  16. A relation between high loading of HLW and space reduction of disposal

    International Nuclear Information System (INIS)

    For an effective use of geological disposal place, namely, the reduction of disposal space, the extension of disposal life, it is important to reduce the amounts of disposed package of high level radioactive waste (HLW). Considered the characteristic of heat generation of HLW during the geological disposal, the effect of space reduction was evaluated by the technical application of a separation of nuclides in HLW and a solidification with high loading of waste. (author)

  17. Kinetics of Conversion of High-Level Waste to Glass

    Czech Academy of Sciences Publication Activity Database

    Izák, Pavel; Hrma, P.; Schweiger, M. J.

    Chapter 19. American Chemical Society, 2000 - (Gary, E.; Heineman, W.), s. 314-328 ISBN 0-8412-3718-2. - (ACS Symposium Series. 778) Institutional research plan: CEZ:AV0Z4072921 Keywords : conversation * glass control * simulated HLW Subject RIV: CI - Industrial Chemistry, Chemical Engineering

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

    International Nuclear Information System (INIS)

    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)

  19. A comparison of the behavior of vitrified HLW in repositories in salt, clay and granite. I: Experimental

    International Nuclear Information System (INIS)

    Laboratories from 11 countries participated in a round robin comparative test exercise using the European Commission's Repository Systems Simulation Test. The compulsory core part of this exercise involved testing an inactive simulant of the French high-level waste glass (SON68) in containers in which the conditions were similar to those that might occur after disposal of vitrified waste in a future repository. There were three options for the test, representing repositories in clay, granite and salt. Fifteen laboratories, mostly from within the European Community, but including laboratories from USA, Canada and Japan took part in the round robin. The HLW (high-level radioactive waste) borosilicate glass SON68 (R7T7), to be returned to and eventually disposed of in various geological repositories by Cogema's reprocessing customers, was tested in a stainless steel autoclave developed and provided by CEA, Marcoule. The glass was prepared and delivered by CEA; together with crushed granite, sand, smectite and groundwater for the granite option. SCK-Mol provided the clay, smectite, sand and the recipe for claywater, HMI-Berlin the recipe for the rock salt composition. The analytical method was the only variable to be chosen by the participants. After termination of the experiments (duration 14-364d, temperature 90C), the concentrations of the glass constituents Si, B, Mo, Li, in solution had to be determined. The data base was used to evaluate the test procedure and to identify analytical problems

  20. Glass-ceramic waste forms for immobilizing plutonium

    International Nuclear Information System (INIS)

    Results are reported on several new glass and glass-ceramic waste formulations for plutonium disposition. The approach proposed involves employing existing calcined high level waste (HLW) present at the Idaho Chemical Processing Plant (ICPP) and an additive to: (1) aid in the formation of a durable waste form and (2) decrease the attractiveness level of the plutonium from a proliferation viewpoint. The plutonium, PuO2, loadings employed were 15 wt% (glass) and 17 wt% (glass-ceramic). Results in the form of x-ray diffraction patterns, microstructure and durability tests are presented on cerium surrogate and plutonium loaded waste forms using simulated calcined HLW and demonstrate that durable phases, zirconia and zirconolite, contain essentially all the plutonium

  1. Glass-ceramic waste forms for immobilizing plutonium

    Energy Technology Data Exchange (ETDEWEB)

    O`Holleran, T.P.; Johnson, S.G.; Frank, S.M.; Meyer, M.K.; Noy, M.; Wood, E.L. [Argonne National Lab.-West, Idaho Falls, ID (United States); Knecht, D.A.; Vinjamuri, K.; Staples, B.A. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States)

    1997-12-31

    Results are reported on several new glass and glass-ceramic waste formulations for plutonium disposition. The approach proposed involves employing existing calcined high level waste (HLW) present at the Idaho Chemical Processing Plant (ICPP) and an additive to: (1) aid in the formation of a durable waste form and (2) decrease the attractiveness level of the plutonium from a proliferation viewpoint. The plutonium, PuO{sub 2}, loadings employed were 15 wt% (glass) and 17 wt% (glass-ceramic). Results in the form of x-ray diffraction patterns, microstructure and durability tests are presented on cerium surrogate and plutonium loaded waste forms using simulated calcined HLW and demonstrate that durable phases, zirconia and zirconolite, contain essentially all the plutonium.

  2. 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)

    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/m2/d and 0.4 MT/m2/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/m2/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

  3. 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

  4. 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.; Kim, Dong-Sang; Skorski, Daniel C.; Matyas, Josef

    2013-07-31

    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.

  5. Moving HLW-EBS concepts into the 21st century

    International Nuclear Information System (INIS)

    Most national high-level waste (HLW) disposal programs actually reflect, or are based on, concepts which were developed during the '70s or early '80s. Although suitable for demonstration of concept feasibility, designs of the engineered barrier system (EBS) do not take into account the tremendous developments in system understanding and materials technology over the last two decades, the practicality (and cost) of their quality assurance and implementation on an industrial scale and the transparency of the demonstration of the safety case. In many ways, due to the increased significance of popular acceptance over the last decade, the last point may be of particular relevance This paper reviews the work already carried out on second generation concepts and extends this to identify the key attributes of an ideal design for the specific case of disposal of vitrified HLW from reprocessing in a 'wet' host rock (either crystalline or sedimentary). Based on the concept developed, key R and D requirements are identified. Copyright (2001) Material Research Society

  6. Local containment and evidence needed in HLW disposal

    International Nuclear Information System (INIS)

    Groundwater- and migration-oriented approaches to HLW disposal suffer from uncertainty, whereas much solid evidence indicates that long-term containment of natural model substances and sensitive indicator minerals is the rule in common, groundwater-saturated silicate rocks. The conditions for persisting containment are outlined. Confirmation from investigations at study sites and actual candidate sites is needed. Elaborate models of the main flow system in this context appear unimportant, because at most sites more than 99.5% of the total ground-water flow would bypass the nearfield of the waste. Only minor, local fluxes can interact with the nearfield barriers. These fluxes and associated transport can be estimated from in-situ measurements. Therefore, licensing and final approval of HLW disposal need not be based on large-scale models of nuclide transport. Nearfield quality control, time-scaled demonstration and monitoring of local containment, and evaluation of potential future changes in the site-specific conditions, are required. (author) 2 figs., 2 tabs., 46 refs

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

    International Nuclear Information System (INIS)

    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

  8. 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.

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

    International Nuclear Information System (INIS)

    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

  10. Immobilization of hazardous and radioactive waste into glass structures

    International Nuclear Information System (INIS)

    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

  11. The precision of product consistency tests conducted with a glass-bonded ceramic waste form

    International Nuclear Information System (INIS)

    The product consistency test (PCT) that is used for qualification of borosilicate high-level radioactive waste (HLW) glasses for disposal can be used for the same purpose in the qualification of the glass-bonded sodalite ceramic waste form (CWF). The CWF was developed to immobilize radioactive salt wastes generated during the electrometallurgical treatment of spent sodium-bonded nuclear fuels. An interlaboratory study was conducted to measure the precision of PCTs conducted with the CWF for comparison with the precision of PCTs conducted with HLW glasses. The six independent sets of triplicate PCT results generated in the study were used to calculate the intralaboratory and interlaboratory consistency based on the concentrations of Al, B, Na, and Si in the test solutions. The results indicate that PCTs can be conducted as precisely with the CWF as with HLW glasses. For example, the values of the reproducibility standard deviation for Al, B, Na, and Si were 1.36, 0.347, 3.40, and 2.97 mg/l for PCT with CWF. These values are within the range of values measured for borosilicate glasses, including reference HLW glasses

  12. Construction of Basic Evaluation Criteria for Candidate HLW Repository Sites

    International Nuclear Information System (INIS)

    The final objectives of the research are to study basically site selection criteria and methods for high level radioactive wastes disposal. In order to accomplish the final objectives, this study proposes basic concept of site evaluation and selection for HLW disposal and develops a geologic database system for management of geologic data which will be used for site selection. This study performed the following contents such as i) development of a technical guide for geologic evaluation system for candidate sites selection; ii) construction of geological evaluation factors and selection methods for candidate sites; iii) development of geologic information system for evaluation and selection of candidate sites. The results are expected to be applied for basic research results for setting up further research foundation for site selection, development of systematic site evaluation process considered with geologic conditions of candidate sites, and utilization of the research result for public acceptance

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

    International Nuclear Information System (INIS)

    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)

  14. Arc spraying solderable tabs to glass

    Science.gov (United States)

    Lindmayer, J.

    1981-01-01

    Tabs suitable for electrical or mechanical connections in solar cells and integrated circuits are made by spraying technique. Solder wets copper, copper bonds to aluminum, and aluminum adheres to glass. Arc spraying is automated and integrated with encapsulation, eliminating hand tabbing, improving reliability, and reducing cost.

  15. 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

  16. Summary Of FY11 Sulfate Retention Studies For Defense Waste Processing Facility Glass

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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 R2 approximately 0.97

  18. Liquidus Temperature Data for DWPF Glass

    Energy Technology Data Exchange (ETDEWEB)

    GF Piepel; JD Vienna; JV Crum; M Mika; P Hrma

    1999-05-21

    This report provides new liquidus temperature (TL) versus composition data that can be used to reduce uncertainty in TL calculation for DWPF glass. According to the test plan and test matrix design PNNL has measured TL for 53 glasses within and just outside of the current DWPF processing composition window. The TL database generated under this task will directly support developing and enhancing the current TL 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 lifecycle tank cleanup costs by decreasing process time and the volume of waste glass produced.

  19. CLOSURE OF HLW TANKS FORMULATION FOR A COOLING COIL GROUT

    Energy Technology Data Exchange (ETDEWEB)

    Harbour, J; Vickie Williams, V; Erich Hansen, E

    2008-05-23

    The Tank Closure and Technology Development Groups are developing a strategy for closing the High Level Waste (HLW) tanks at the Savannah River Site (SRS). Two Type IV tanks, 17 and 20 in the F-Area Tank Farm, have been successfully filled with grout. Type IV tanks at SRS do not contain cooling coils; on the other hand, the majority of the tanks (Type I, II, III and IIIA) do contain cooling coils. The current concept for closing tanks equipped with cooling coils is to pump grout into the cooling coils to prevent pathways for infiltrating water after tank closure. This task addresses the use of grout to fill intact cooling coils present in most of the remaining HLW tanks on Site. The overall task was divided into two phases. Phase 1 focused on the development of a grout formulation (mix design) suitable for filling the HLW tank cooling coils. Phase 2 will be a large-scale demonstration of the filling of simulated cooling coils under field conditions using the cooling coil grout mix design recommended from Phase 1. This report summarizes the results of Phase 1, the development of the cooling coil grout formulation. A grout formulation is recommended for the full scale testing at Clemson Environmental Technology Laboratory (CETL) that is composed by mass of 90% Masterflow (MF) 816 (a commercially available cable grout) and 10% blast furnace slag, with a water to cementitious material (MF 816 + slag) ratio of 0.33. This formulation produces a grout that meets the fresh and cured grout requirements detailed in the Task Technical Plan (2). The grout showed excellent workability under continuous mixing with minimal change in rheology. An alternative formulation using 90% MF 1341 and 10% blast furnace slag with a water to cementitious material ratio of 0.29 is also acceptable and generates less heat per gram than the MF 816 plus slag mix. However this MF 1341 mix has a higher plastic viscosity than the MF 816 mix due to the presence of sand in the MF 1341 cable grout and a

  20. 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.

  1. Extraction studies on the partitioning of actinides from HLW

    International Nuclear Information System (INIS)

    TRUEX process uses Octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) for the partitioning of actinides from acidic waste. CMPO is one of the most effective organophosphorus compounds. CMPO has drawbacks like third phase formation. A two-step process is developed using TBP and CMPO as extractants. The first step of the proposed process is a 'uranium depletion step' in which uranium in HLW is removed using 30% TBP in n-dodecane. Neptunium and plutonium, extracted in TBP, are recovered using a mixture of hydrogen peroxide (0.25 M) and ascorbic acid (0.05 M) in 2.0 M nitric acid medium. Neptunium and plutonium are reduced to Np(V) and Pu(III). Americium and curium as well as traces of uranium, neptunium and plutonium are partitioned in the second step. The separation of neptunium and plutonium from large quantities of uranium from loaded TBP will simplify their further separation. Use of citrate containing buffer solution for the recovery of actinides extracted in CMPO-TBP phase eliminates the problem of reflux of americium. This reduces the generation of secondary wastes. The process has been standardised based on the data generated in the batch and counter-current studies. Solvent extraction studies have also been carried out using a mixture of di-(2-ethylhexyl)phosphoric acid (HDEHP) and CMPO in n-paraffin. It is seen that the actinides can be extracted even from solutions of HLW at high acid concentration of 3 M using the mixed extractant. Plutonium is also stripped along with the trivalent actinides. This mixed solvent may find useful applications in the partitioning of actinides from waste solutions. Supported liquid membrane (SLM) technique for partitioning of actinides from high level waste of PUREX origin uses solution of CMPO in n-dodecane with polytetrafluoroethylene support and a mixture of citric acid, formic acid and hydrazine hydrate as a receiving phase. Studies indicated good transport of actinides across the membrane from nitric acid

  2. German approaches to closing the nuclear fuel cycle and final disposal of HLW

    International Nuclear Information System (INIS)

    The paper describes the present situation of the German approach to closing the nuclear fuel cycle and the disposal concept for high level waste (HLW), either waste from reprocessing or spent fuel. The first part of the paper deals with the spent fuel generation from a 22 GWe scenario, comparing especially the anticipated plutonium inventories after different fuel burnup and its amounts in HLW for disposal with and without recycling. The second part comprises the geological disposal concept of HLW, either as spent fuel or vitrified high level waste from reprocessing, in a salt formation in Gorleben. The third part includes an approach to safety assessment of the HLW repository for the post-operational period by applying a multibarrier concept that covers the appraisal of the geochemical and migration behaviour of long-lived radionuclides in given aquifer systems. (orig.)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  5. 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

  6. Formulation of Japanese consensus-building model for HLW geological disposal site determination. 1. Introduction

    International Nuclear Information System (INIS)

    To establish the sustainable community in Japan, formation of Japanese consensus-building model for HLW geological disposal site determination is one of key issues. In our project, we have reviewed the past history for HLW geological disposal site determination and propose next-generation Japanese consensus-building model which is based on the discussion not only with government and specialists but also with citizens. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. Biosphere modelling for a HLW repository scenario and parameter variations

    International Nuclear Information System (INIS)

    In Switzerland high-level radioactive wastes have been considered for disposal in deep-lying crystalline formations. The individual doses to man resulting from radionuclides entering the biosphere via groundwater transport are calculated. The main recipient area modelled, which constitutes the base case, is a broad gravel terrace sited along the south bank of the river Rhine. An alternative recipient region, a small valley with a well, is also modelled. A number of parameter variations are performed in order to ascertain their impact on the doses. Finally two scenario changes are modelled somewhat simplistically, these consider different prevailing climates, namely tundra and a warmer climate than present. In the base case negligibly low doses to man in the long term, resulting from the existence of a HLW repository have been calculated. Cs-135 results in the largest dose (8.4x10-7 mrem/y at 6.1x106 y) while Np-237 gives the largest dose from the actinides (3.6x10-8 mrem/y). The response of the model to parameter variations cannot be easily predicted due to non-linear coupling of many of the parameters. However, the calculated doses were negligibly low in all cases as were those resulting from the two scenario variations. (author)

  10. 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.

  11. Public Perspectives in the Japanese HLW Disposal Program

    International Nuclear Information System (INIS)

    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

  12. Chemical modeling of cementitious grout materials alteration in HLW repositories

    International Nuclear Information System (INIS)

    This paper reports on an investigation initiated into the nature of the chemical alteration of cementitious grout in HLW repository seals, and the implications for long-term seal performance. The equilibrium chemical reaction of two simplified portland cement-based grout models with natural Canadian Shield groundwater compositions was modeled with the computer codes PHREEQE and EQ3NR/EQ6. Increases in porosity and permeability of the grout resulting from dissolution of grout phases and precipitation of secondary phases were estimated. Two bounding hydrologic scenarios were evaluated, one approximating a high gradient, high flow regime, the other a low-gradient, sluggish flow regime. Seal longevity depends in part upon the amount of groundwater coming into intimate contact with, and dissolving, the grout per unit time. Results of the analyses indicate that, given the assumptions and simplifications inherent in the models, acceptable seal performance (i.e., acceptable increases in hydraulic conductivity of the seals) may be expected for at least thousands of years in the worst cases analyzed, and possibly much longer

  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, 5/6/04

    International Nuclear Information System (INIS)

    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. 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

  15. 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

  16. Acoustic emission characteristics of single edge notched glass fiber/metal laminates

    International Nuclear Information System (INIS)

    Fracture behaviors of single-edge-notched monolithic aluminum sheets and glass fiber/aluminum laminates under tensile loadings have been investigated using acoustic emission(AE) monitoring. AE signals from monolithic aluminum could be classified into two different types. For glass fiber/aluminum laminates, AE signals with high amplitude and long duration were additionally confirmed on FFT frequency analysis, which corresponded to macrocrack propagation and/or delamination. On the basis of the above AE analysis and fracture observation, characteristic features of fracture processes of single-edge-notched glass fiber/aluminum laminates were elucidated according to different fiber ply orientations.

  17. Multi-barrier, copper-base containers for HLW disposal

    International Nuclear Information System (INIS)

    Copper and aluminum bronze have been shown to exhibit high degrees of kinetic stability in anticipated repository environments, including high gamma fields. The nature of their thermodynamic and kinetic stabilities are discussed. It is proposed that a robust, composite waste container comprised of a copper mantle surrounding an inner shell of aluminum bronze would make the best use of the metals' corrosion- and creep-related properties. Several designs and closure techniques are suggested. A bimetallic, centrifugally cast cylinder with a diameter and wall thickness appropriate to a high-level waste burial container has been produced. The advantages of the bimetallic casting are discussed, including suggestions for future work. Creation of an engineered analog is suggested as an additional redundant safeguard in the proposed repository

  18. Corrosion resistance of metal materials for HLW canister

    International Nuclear Information System (INIS)

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

  19. 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.

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

    International Nuclear Information System (INIS)

    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 (SiO2: 30.5 wt%, B2O3: 20.0 wt%, Na2O: 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

  1. IR Laser Plasma Interaction with Glass

    OpenAIRE

    Rabia Qindeel; Noriah Bidin; Yaacob M.  Daud

    2007-01-01

    The interaction of laser plasma with respect to glass surface is reported in this paper. A Q-switched Nd:YAG laser was used as ablation source. Glass material is utilized as target specimen. Aluminum plate is used as a rotating substrate. The dynamic expansion of the plasma was visualized by using CCD video camera and permanently recorded via image processing system. The exposed glass material was examined under photomicroscope and scanning electron microscope (SEM). The optical radiation fro...

  2. Actinide partitioning and recovery of valuables from HLW

    International Nuclear Information System (INIS)

    The Indian nuclear power programme is sustained by adoption of a closed fuel cycle where in the fissile and fertile materials are recycled by reprocessing of spent fuel. The reprocessing step leads to the generation of high level waste which is presently vitrified using borosilicate matrices. With the nuclear power profile on the brink of an exponential increase, it becomes imperative to consider and adopt cross-cut technologies that would not only lead to a substantial reduction in repository capacity both in terms of volumes and thermal loads but also lead to a reduction in radiotoxicity of the waste forms. Partitioning of high level waste (HLW) is the first step towards achieving the above objectives. Developmental efforts in the last decade have placed partitioning of high level waste in the realms of practical application. This paper will present a compilation of various R and D efforts on development of processes and technologies under consideration for partitioning of high level waste in the Indian context. While numerous laboratory trials are being pursued, some of them which have matured for plant scale demonstration are related to partitioning of actinides from acidic high level waste and recovery of cesium and strontium from high level waste. A structured R and D framework has been worked out to develop deployable processes and technologies for their demonstration on engineering scale. One of the most defining step in this work is selection of potentially successful extraction system based on the systematic study on the extraction properties and their optimization for full scale studies. (author)

  3. Application of Archimedes Filter for Reduction of Hanford HLW

    International Nuclear Information System (INIS)

    Archimedes Technology Group, Inc., is developing a plasma mass separator called the Archimedes Filter that separates waste oxide mixtures ion by ion into two mass groups: light and heavy. For the first time, it is feasible to separate large amounts of material atom by atom in a single pass device. Although vacuum ion based electromagnetic separations have been around for many decades, they have traditionally depended on ion beam manipulation. Neutral plasma devices, on the other hand, are much easier, less costly, and permit several orders of magnitude greater throughput. The Filter has many potential applications in areas where separation of species is otherwise difficult or expensive. In particular, radioactive waste sludges at Hanford have been a particularly difficult issue for pretreatment and immobilization. Over 75% of Hanford HLW oxide mass (excluding water, carbon, and nitrogen) has mass less than 59 g/mol. On the other hand, 99.9% of radionuclide activity has mass greater than 89 g/mol. Therefore, Filter mass separation tuned to this cutoff would have a dramatic effect on the amount of IHLW produced--in fact IHLW would be reduced by a factor of at least four. The Archimedes Filter is a brand new tool for the separations specialist's toolbox. In this paper, we show results that describe the extent to which the Filter separates ionized material. Such results provide estimates for the potential advantages of Filter tunability, both in cutoff mass (electric and magnetic fields) and in degree of ionization (plasma power). Archimedes is now engaged in design and fabrication of its Demonstration Filter separator and intends on performing a full-scale treatment of Hanford high-level waste surrogates. The status of the Demo project will be described

  4. Characterization of uranium, plutonium, neptunium, and americium in HLW supernate for LLW certification

    International Nuclear Information System (INIS)

    The 1S Manual requires that High Level Waste (HLW) implement a waste certification program prior to sending waste packages to the E-Area vaults. To support the waste certification plan, the HLW supernate inventory of uranium, plutonium, neptunium and americium have been characterized. This characterization is based on the chemical, isotopic and radiological properties of these elements in HLW supernate. This report uses process knowledge, solubility data, isotopic inventory data and sample data to determine if any isotopes of the aforementioned elements will exceed the minimum reportable quantity (MRQ) for waste packages contaminated with HLW supernate. If the MRQ can be exceeded for a particular nuclide, then a method for estimating the waste package content is provided. Waste packages contaminated from HLW supernate do not contain sufficient U-233, U-234, U-235, U-236, U-238, Pu-239, Pu-240, Pu-241, Pu-242 or Am-241 to warrant separate reporting on the shipping manifest. Calculations show that, on average, more than 100 gallons of supernate is required to exceed the PAC (package acceptance criteria) for each of these nuclides. Thus it is highly unlikely that the PAC would be exceeded for these nuclides and unlikely that the MRQ would be exceeded. These nuclides should be manifested as zero for waste packages contaminated with HLW supernate. The only actinide isotopes that may exceed the MRQ are Np-237 and Pu-238. The recommended method to calculate the amount of these two isotopes in waste packages contaminated with HLW supernate is to ratio them to the measured Cs-137 activity

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

    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)

    2016-06-01

    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.

  6. Thermal Analysis Of Waste Glass Melter Feeds

    International Nuclear Information System (INIS)

    Melter feeds for high-level nuclear waste (HLW) typically contain a large number of constituents that evolve gas on heating, Multiple gas-evolving reactions are both successive and simultaneous, and include the release of chemically bonded water, reactions of nitrates with organics, and reactions of molten salts with solid silica. Consequently, when a sample of a HLW feed is subjected to thermogravimetric analysis (TGA), the rate of change of the sample mass reveals multiple overlapping peaks. In this study, a melter feed, formulated for a simulated high-alumina HLW to be vitrified in the Waste Treatment and Immobilization Plant, currently under construction at the Hanford Site in Washington State, USA, was subjected to TGA. In addition, a modified melter feed was prepared as an all-nitrate version of the baseline feed to test the effect of sucrose addition on the gas-evolving reactions. Activation energies for major reactions were determined using the Kissinger method. The ultimate aim of TGA studies is to obtain a kinetic model of the gas-evolving reactions for use in mathematical modeling of the cold cap as an element of the overall model of the waste-glass melter. In this study, we focused on computing the kinetic parameters of individual reactions without identifying their actual chemistry, The rough provisional model presented is based on the first-order kinetics.

  7. Modelling spent fuel and HLW behaviour in repository conditions

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

  8. ROLE OF MANGANESE REDUCTION/OXIDATION (REDOX) ON FOAMING AND MELT RATE IN HIGH LEVEL WASTE (HLW) MELTERS (U)

    International Nuclear Information System (INIS)

    High-level nuclear waste is being immobilized at the Savannah River Site (SRS) by vitrification into borosilicate glass at the Defense Waste Processing Facility (DWPF). Control of the Reduction/Oxidation (REDOX) equilibrium in the DWPF melter is critical for processing high level liquid wastes. Foaming, cold cap roll-overs, and off-gas surges all have an impact on pouring and melt rate during processing of high-level waste (HLW) glass. All of these phenomena can impact waste throughput and attainment in Joule heated melters such as the DWPF. These phenomena are caused by gas-glass disequilibrium when components in the melter feeds convert to glass and liberate gases such as H2O vapor (steam), CO2, O2, H2, NOx, and/or N2. During the feed-to-glass conversion in the DWPF melter, multiple types of reactions occur in the cold cap and in the melt pool that release gaseous products. The various gaseous products can cause foaming at the melt pool surface. Foaming should be avoided as much as possible because an insulative layer of foam on the melt surface retards heat transfer to the cold cap and results in low melt rates. Uncontrolled foaming can also result in a blockage of critical melter or melter off-gas components. Foaming can also increase the potential for melter pressure surges, which would then make it difficult to maintain a constant pressure differential between the DWPF melter and the pour spout. Pressure surges can cause erratic pour streams and possible pluggage of the bellows as well. For these reasons, the DWPF uses a REDOX strategy and controls the melt REDOX between 0.09 (le) Fe2+/(summation)Fe (le) 0.33. Controlling the DWPF melter at an equilibrium of Fe+2/(summation)Fe (le) 0.33 prevents metallic and sulfide rich species from forming nodules that can accumulate on the floor of the melter. Control of foaming, due to deoxygenation of manganic species, is achieved by converting oxidized MnO2 or Mn2O3 species to MnO during melter preprocessing. At the

  9. ROLE OF MANGANESE REDUCTION/OXIDATION (REDOX) ON FOAMING AND MELT RATE IN HIGH LEVEL WASTE (HLW) MELTERS (U)

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C; Michael Stone, M

    2007-03-30

    High-level nuclear waste is being immobilized at the Savannah River Site (SRS) by vitrification into borosilicate glass at the Defense Waste Processing Facility (DWPF). Control of the Reduction/Oxidation (REDOX) equilibrium in the DWPF melter is critical for processing high level liquid wastes. Foaming, cold cap roll-overs, and off-gas surges all have an impact on pouring and melt rate during processing of high-level waste (HLW) glass. All of these phenomena can impact waste throughput and attainment in Joule heated melters such as the DWPF. These phenomena are caused by gas-glass disequilibrium when components in the melter feeds convert to glass and liberate gases such as H{sub 2}O vapor (steam), CO{sub 2}, O{sub 2}, H{sub 2}, NO{sub x}, and/or N{sub 2}. During the feed-to-glass conversion in the DWPF melter, multiple types of reactions occur in the cold cap and in the melt pool that release gaseous products. The various gaseous products can cause foaming at the melt pool surface. Foaming should be avoided as much as possible because an insulative layer of foam on the melt surface retards heat transfer to the cold cap and results in low melt rates. Uncontrolled foaming can also result in a blockage of critical melter or melter off-gas components. Foaming can also increase the potential for melter pressure surges, which would then make it difficult to maintain a constant pressure differential between the DWPF melter and the pour spout. Pressure surges can cause erratic pour streams and possible pluggage of the bellows as well. For these reasons, the DWPF uses a REDOX strategy and controls the melt REDOX between 0.09 {le} Fe{sup 2+}/{summation}Fe {le} 0.33. Controlling the DWPF melter at an equilibrium of Fe{sup +2}/{summation}Fe {le} 0.33 prevents metallic and sulfide rich species from forming nodules that can accumulate on the floor of the melter. Control of foaming, due to deoxygenation of manganic species, is achieved by converting oxidized MnO{sub 2} or Mn

  10. Glass sealing

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

  11. A STATISTICAL REVIEW OF THE CHEMICAL COMPOSITION MEASUREMENTS AND PCT RESULTS FOR THE GLASSES FABRICATED AS PART OF THE US TEST MATRIX

    International Nuclear Information System (INIS)

    The Savannah River National Laboratory (SRNL) is part of a consortium that is looking to improve the retention of aluminum, chromium, and sulfate in high level radioactive waste (HLW) glass. Such glass has been produced by the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) in South Carolina since it began operating in 1996 and is planned to be produced by the River Protection Project-Waste Treatment Plant (WTP) at the Hanford Site in Washington. The consortium conducting this study, which is designated as Task No.6 by the Department of Energy (DOE) Environmental Management (EM) program sponsoring this effort, is made up of personnel from SRNL, the Pacific Northwest National Laboratory (PNNL), and the V.G. Khlopin Radium Institute (KRI). Coordinated glass experimental work will be performed by each member of the consortium. The glasses that are being studied were selected to further the understanding of composition-property relationships within the glass regions of interest to both DWPF and WTP. Forty-five (45) glasses, making up the US test matrix, were batched and fabricated to support the study. The chemical compositions of these glasses were measured by SRNL's Process Science Analytical Laboratory (PSAL) under the auspices of an analytical plan. In addition, two heat treatments (quenched and centerline canister cooled, ccc) of each glass were subjected to the 7-day Product Consistency Test (PCT) to assess their durabilities. More specifically, the Method A of the PCT (ASTM C-1285-2002) was used for these tests. Measurements of the resulting leachate solutions were conducted by PSAL under the auspices of three analytical plans. A statistical review of the PSAL measurements of the chemical compositions and of the PCT results for the glasses making up the US test matrix is provided in this memorandum. Target, measured, and measured bias-corrected compositional views were determined for these glasses. The durability results for the US

  12. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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 directly. The

  15. 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.

  16. A State-of-the Art on the Siting Cases of Nuclear Advanced Nations for a HLW Repository

    International Nuclear Information System (INIS)

    We analyzed the siting cases of nuclear advanced nations (e.g. Finland, Sweden, U.S., Japan, France, Germany, and Switzerland) for a HLW repository. The analysis includes the siting standard, procedure, and the factors to be considered of each nation for a HLW repository. We are expecting that the results can be used to develop new research topics

  17. Comparison of Micro-Leakage from Resin-Modified Glass Ionomer Restorations in Cavities Prepared by Er:YAG (Erbium-Doped Yttrium Aluminum Garnet) Laser and Conventional Method in Primary Teeth

    OpenAIRE

    Bahrololoomi, Zahra; Razavi, Forooghosadat; Soleymani, Ali Asghar

    2014-01-01

    Introduction: In recent years, significant developments have been taking place in caries removal and cavity preparation using laser in dentistry. As laser use is considered for cavity preparation, it is necessary to determine the quality of restoration margins. Glass ionomer cements have great applications for conservative restoration in the pediatric field. The purpose of this in vitro study was to compare resin-modified glass ionomer restorations micro-leakage in cavities prepared by Er:YAG...

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

    International Nuclear Information System (INIS)

    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

  19. Spark plasma sintering of aluminum matrix composites

    Science.gov (United States)

    Yadav, Vineet

    2011-12-01

    Aluminum matrix composites make a distinct category of advanced engineering materials having superior properties over conventional aluminum alloys. Aluminum matrix composites exhibit high hardness, yield strength, and excellent wear and corrosion resistance. Due to these attractive properties, aluminum matrix composites materials have many structural applications in the automotive and the aerospace industries. In this thesis, efforts are made to process high strength aluminum matrix composites which can be useful in the applications of light weight and strong materials. Spark Plasma Sintering (SPS) is a relatively novel process where powder mixture is consolidated under the simultaneous influence of uniaxial pressure and pulsed direct current. In this work, SPS was used to process aluminum matrix composites having three different reinforcements: multi-wall carbon nanotubes (MWCNTs), silicon carbide (SiC), and iron-based metallic glass (MG). In Al-CNT composites, significant improvement in micro-hardness, nano-hardness, and compressive yield strength was observed. The Al-CNT composites further exhibited improved wear resistance and lower friction coefficient due to strengthening and self-lubricating effects of CNTs. In Al-SiC and Al-MG composites, microstructure, densification, and tribological behaviors were also studied. Reinforcing MG and SiC also resulted in increase in micro-hardness and wear resistance.

  20. Glass microsphere lubrication

    Science.gov (United States)

    Geiger, Michelle; Goode, Henry; Ohanlon, Sean; Pieloch, Stuart; Sorrells, Cindy; Willette, Chris

    1991-01-01

    The harsh lunar environment eliminated the consideration of most lubricants used on earth. Considering that the majority of the surface of the moon consists of sand, the elements that make up this mixture were analyzed. According to previous space missions, a large portion of the moon's surface is made up of fine grained crystalline rock, about 0.02 to 0.05 mm in size. These fine grained particles can be divided into four groups: lunar rock fragments, glasses, agglutinates (rock particles, crystals, or glasses), and fragments of meteorite material (rare). Analysis of the soil obtained from the missions has given chemical compositions of its materials. It is about 53 to 63 percent oxygen, 16 to 22 percent silicon, 10 to 16 percent sulfur, 5 to 9 percent aluminum, and has lesser amounts of magnesium, carbon, and sodium. To be self-supporting, the lubricant must utilize one or more of the above elements. Considering that the element must be easy to extract and readily manipulated, silicon or glass was the most logical choice. Being a ceramic, glass has a high strength and excellent resistance to temperature. The glass would also not contaminate the environment as it comes directly from it. If sand entered a bearing lubricated with grease, the lubricant would eventually fail and the shaft would bind, causing damage to the system. In a bearing lubricated with a solid glass lubricant, sand would be ground up and have little effect on the system. The next issue was what shape to form the glass in. Solid glass spheres was the only logical choice. The strength of the glass and its endurance would be optimal in this form. To behave as an effective lubricant, the diameter of the spheres would have to be very small, on the order of hundreds of microns or less. This would allow smaller clearances between the bearing and the shaft, and less material would be needed. The production of glass microspheres was divided into two parts, production and sorting. Production includes the

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

    International Nuclear Information System (INIS)

    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

  2. Construction of Basic Evaluation Criteria for Candidate HLW Repository Sites (2)

    International Nuclear Information System (INIS)

    The final objectives of the research are to study basically site selection criteria and methods for high level radioactive wastes disposal. In order to accomplish the final objectives, this study proposes basic concept of site evaluation and selection for HLW disposal and develops a geologic database system for management of geologic data which will be used for site selection

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

    International Nuclear Information System (INIS)

    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/m3 , 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 m3 and the weight was 21.7 Mg. The dry density of the entire buffer was 1.62 Mg/m3 , 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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  6. 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.

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

    International Nuclear Information System (INIS)

    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

  8. The Aluminum Smelting Process

    OpenAIRE

    Kvande, Halvor

    2014-01-01

    This introduction to the industrial primary aluminum production process presents a short description of the electrolytic reduction technology, the history of aluminum, and the importance of this metal and its production process to modern society. Aluminum's special qualities have enabled advances in technologies coupled with energy and cost savings. Aircraft capabilities have been greatly enhanced, and increases in size and capacity are made possible by advances in aluminum technology. The me...

  9. CORALUS - An integrated underground test on the corrosion of high-level radioactive waste glass

    International Nuclear Information System (INIS)

    As part of the studies on the long-term performance of the French R7T7 HLW (high-level radioactive waste) glass in a deep underground repository that is backfilled with clay, we have performed during the last 10 years the CORALUS project (CORrosion of alpha-Active gLass in Underground Storage conditions). The CORALUS project combines underground and surface laboratory integrated tests on the alteration of SON 68 reference glass that simulates the R7T7 HLW glass. Knowledge of the behaviour of this type of waste in disposal conditions contributes to formulate radionuclide source terms, which are mathematical descriptions of the release of radionuclides from the waste as a function of time.The objectives of the CORALUS project are: to compare the results from integrated in situ glass corrosion tests, performed under realistic disposal conditions, with results from surface laboratory experiments and modelling predictions, and this for two temperatures (30 degrees Celsius and 90 degrees Celsiusq) and three backfill materials; to study, under realistic disposal conditions, (1) the combined effect of high temperature and gamma irradiation and (2) the effect of the specific alpha activity of the glass, on the glass corrosion. The results will help to assess the validity of the insights in the SON 68 glass corrosion, as these were mostly obtained from less integrated surface laboratory experiments performed under less representative conditions

  10. EVALUATION OF LOW TEMPERATURE ALUMINUM DISSOLUTION IN TANK 51

    Energy Technology Data Exchange (ETDEWEB)

    Pike, J

    2008-09-04

    Liquid Waste Organization (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, which became the baseline aluminum dissolution process. LWO initiated a project to modify a waste tank to meet these requirements. Subsequent to an alternative evaluation, LWO management identified an opportunity to perform aluminum dissolution on sludge destined for Sludge Batch 5, but within a limited window that would not allow time for any modifications for tank heating. A variation of the baseline process, dubbed Low Temperature Aluminum Dissolution (LTAD), was developed based on the constraint of available energy input in Tank 51 and the window of opportunity, but was not constrained to a minimum extent of dissolution, i.e. dissolve as much aluminum as possible within the time available. This process was intended to operate between 55 and 70 C, but for a significantly longer time than the baseline process. LTAD proceeded in parallel with the baseline project. The preliminary evaluation at the completion of LTAD focused on the material balance and extent of the aluminum dissolved. The range of values of extent of dissolution, 56% to 64%, resulted from the variation in liquid phase sample data available at the time. Additional solid phase data is available from a sample taken after LTAD to refine this range. This report provides additional detailed evaluation of the LTAD process based on analytical and field data and includes: a summary of the process chronology; a determination of an acceptable blending strategy for the aluminum-laden supernate stored in Tank 11; an update to the determination of aluminum dissolved using more complete sample results; a determination of the effect of LTAD on uranium, plutonium, and other metals; a determination of the rate of heat

  11. Corrosion behavior of simulated high-level waste glass in the presence of calcium ion or metallic iron

    International Nuclear Information System (INIS)

    Static leach tests were conducted for simulated high-level waste (HLW) glass in CaCl2/Ca(OH)2 solutions to investigate the corrosion behavior of HLW glass under calcium-rich environments induced by cement based materials in geological repositories. Another series of leach tests were conducted in deionized water in the presence of iron to investigate the effects of iron over-pack on the glass corrosion. In CaCl2/Ca(OH)2 solutions, corrosion of the glass was inhibited during the test period compared to that in deionized water at the pH range of 6 - 11, while higher corrosion rate was observed in the initial stage of the test in Ca(OH)2 solution at the initial pH of 12. However, the corrosion rate dropped due to a formation of calcium silicates that covered the surface of the glass. Under the condition that iron exists in the vicinity of the glass, glass corrosion was enhanced compared to that without iron throughout the testing period. In addition, an alteration layer including iron and silicon was observed at the interface between the glass surface and the iron after the leach tests, and thermodynamic calculation showed that formation of an iron silicate was favored under the chemical compositions of the leachate during the period. The enhancement of the glass corrosion was assumed to be accompanied with transformation of silica, a glass network former, into iron silicates. (author)

  12. 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.

  13. 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.

  14. Graphene-aluminum nanocomposites

    International Nuclear Information System (INIS)

    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.

  15. Graphene-aluminum nanocomposites

    Energy Technology Data Exchange (ETDEWEB)

    Bartolucci, Stephen F., E-mail: stephen.bartolucci@us.army.mil [U.S. Army Benet Laboratories, Armaments Research Development and Engineering Center, Watervliet, NY 12189-4000 (United States); Paras, Joseph [U.S. Army Benet Laboratories, Armaments Research Development and Engineering Center, Watervliet, NY 12189-4000 (United States); Rafiee, Mohammad A. [Department of Mechanical Engineering and Materials Science, Rice University, Houston, TX 77005 (United States); Rafiee, Javad [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy, New York 12180 (United States); Lee, Sabrina; Kapoor, Deepak [U.S. Army Benet Laboratories, Armaments Research Development and Engineering Center, Watervliet, NY 12189-4000 (United States); Koratkar, Nikhil, E-mail: koratn@rpi.edu [Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy, New York 12180 (United States)

    2011-10-15

    Highlights: {yields} We investigated the mechanical properties of aluminum and aluminum nanocomposites. {yields} Graphene composite had lower strength and hardness compared to nanotube reinforcement. {yields} Processing causes aluminum carbide formation at graphene defects. {yields} 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.

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

    International Nuclear Information System (INIS)

    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 FeCl3 species. Results indicate that the forward reaction rate increases with an increase in the FeCl3 concentration. The addition of FeCl3 causes the drop in the pH due to hydrolysis of Fe3+ 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)

  17. Decontamination glass

    International Nuclear Information System (INIS)

    Glass for the decontamination of the furnace for vitrification of radioactive wastes contains 50 to 60 wt.% of waste glass, 15 to 30 wt.% of calcium oxide, 1 to 6 wt.% sodium oxide, 1 to 5 wt.% phosphorus pentoxide and 5 to 20 wt.% boron oxide. The melting furnace is flushed with the glass such that it melts in the furnace for at least 60 mins and is then poured out of the furnace. After the furnace has cooled down the settled glass spontaneously cracks and peels off the walls leaving a clean surface. The glass may be used not only for decontamination of the furnace but also for decontamination of melting crucibles and other devices contaminated with radioactive glass. (J.B.)

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

    International Nuclear Information System (INIS)

    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 (99Tc 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

  19. A safety assessment test for geological disposal of HLW in China

    International Nuclear Information System (INIS)

    A post-closure safety assessment test of HLW repository has been conducted under the concept design of engineered barrier system (EBS), and geosphere characteristics in Beishan. Nuclide migration of the high-level waste (HLW) repository in the near- and far-field as well as a transport through a biosphere has been modeled by utilizing GoldSim. The results show that the maximum annual effective dose is 3.95 E-02 μSv which is mainly contributed by 79Se, 126Sn and 135 Cs under reference scenario. The maximum annual effective dose is 1.22 μSv, 2.22 μSv and 5.42 μSv under other scenarios, namely no solubility limitations, negligible transport resistance in the buffer and no retention in the rock, respectively. (authors)

  20. Effect of microstructure changes on the mobility of radionuclides in simulated HLW ceramics

    International Nuclear Information System (INIS)

    Ceramic matrices for immobilization of HLW (e.g. perovskite, zirconolite, brannerite, zircon mineral based ceramics), prepared at ANSTO and C.I.A.E., respectively were characterized at the NRI Rez from the viewpoint of their microstructure and transport property changes caused by leaching. Scanning electron microscopy (SEM) and diffusion structural analysis (DSA) techniques were used. The thermal behavior of 'as leached' and 'as prepared' samples were compared. The DSA was used for the evaluation of atomic transport properties of the ceramic matrices. The effect of leaching on the thermal stability of the ceramics microstructure was characterized. The behaviour of the ceramic HLW matrices in simulated repository conditions was predicted by using the results of the mathematical modeling. (author)

  1. A construction of PA tool for promoting public understanding of HLW and geological disposal (III)

    International Nuclear Information System (INIS)

    'High-level Radioactive Waste (HLW) and Geological Disposal' is a subject, which comes to have more and more importance in public acceptance (PA) activities. In PA activities various methods or tools have been tried, however, apparently not with an appreciable effect yet. In '97 study therefore, based on the survey and analysis of the status of educational methods in PA activities, a more effective tool for those activities was devised and its concept was designed. Moreover, a prototype (data system and sample date) was made up for verifying effectiveness of the concept, and functions to be realized were clarified. In this study, PA tool for promoting Public Understanding of HLW and Geological Disposal was constructed, based on the '98 study. (author)

  2. Laboratory measurement of a long term behavior in HLW near-field by centrifugal model test

    International Nuclear Information System (INIS)

    The objective of this paper is to evaluate the long term behavior of HLW near-field by the centrifugal model test. The model specimen consists of rock mass, bentonite buffer and model waste. The specimen was enclosed with the pressure vessel and centrifugal model tests were conducted at 30 G of centrifugal force field with confining pressures of 2 to 10 MPa and injecting water. As a result, we observed the strain of rock mass, swelling of bentonite and displacement of overpack, and these values did not converged more than 100 equivalent years. In addition, the measured values showed the confining pressure dependency. It is suggested that the stability of HLW in the bentonite buffer takes more than 100 years at least. (author)

  3. Recycle Glass in Foam Glass Production

    OpenAIRE

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

    2014-01-01

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

  4. Recycle Glass in Foam Glass Production

    DEFF Research Database (Denmark)

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

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

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

    International Nuclear Information System (INIS)

    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.)

  6. Experimental investigations on the thermal conductivity characteristics of Beishan granitic rocks for China's HLW disposal

    Science.gov (United States)

    Zhao, X. G.; Wang, J.; Chen, F.; Li, P. F.; Ma, L. K.; Xie, J. L.; Liu, Y. M.

    2016-06-01

    Crystalline rocks are potential host rock types for the construction of high-level radioactive waste (HLW) repositories. A better understanding of thermal conductivity of rocks is essential to safe evaluation and engineering optimization of a HLW disposal system in the rock at depth. In the present study, experimental investigations on the thermal conductivity characteristics of 47 pairs of granitic rock specimens were conducted using the Transient Plane Source (TPS) method. The specimens were collected from borehole cores in the Beishan area, which is being considered as the most potential candidate area for China's HLW repository. To evaluate geological nature of the rocks, mineralogical compositions of the rocks were identified, and porosity of the specimens was measured. The thermal conductivities of the specimens under dry and water-saturated conditions were determined, and the effect of water saturation on the thermal conductivity was investigated. In addition, the influence of temperature and axial compression stress on the thermal conductivity of dry specimens was studied. The results revealed that the thermal conductivity of tested rocks was dependent on water saturation, temperature and compression stress. Based on the obtained data, some models considering porosity were established for describing the thermal conductivity characteristics of the tested rocks. Furthermore, when the rocks have a similar porosity, the quartz content dominates the thermal conductivity, and there exists an obvious increase of the thermal conductivity with increasing quartz content. The test results constitute the first systematic measurements on the Beishan granitic rocks and can further be used for the development of thermal models for predicting thermal response near the underground excavations for HLW disposal.

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

    International Nuclear Information System (INIS)

    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

  8. Nonwoven glass fiber mat reinforces polyurethane adhesive

    Science.gov (United States)

    Roseland, L. M.

    1967-01-01

    Nonwoven glass fiber mat reinforces the adhesive properties of a polyurethane adhesive that fastens hardware to exterior surfaces of aluminum tanks. The mat is embedded in the uncured adhesive. It ensures good control of the bond line and increases the peel strength.

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

    International Nuclear Information System (INIS)

    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-products. 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

  10. Are We Serious in the US about the Disposal of HLW? - 13561

    International Nuclear Information System (INIS)

    Since all efforts to date to dispose of HLW in the US have been unsuccessful, the following specific actions need to be taken if we are serious about such disposal: - The requirement in the EPA environmental radiation protection standards to predict the behavior of these unwanted residuals for one million years is meaningless. The Standards must be revisited. - Characterize two sites. There are myriad ways a site can be found to be unacceptable. Additionally, the existing HLW inventory requires a second repository. - Congress should specify incentives to states under consideration for a site. Perhaps 5% of total cost would be appropriate. - An independent technical review group should be established in such states to evaluate a proposed repository similar to the New Mexico Environmental Evaluation Group (EEG) for the WIPP Project because the state's interests are not necessarily the same as DOE's. - Acceptance or rejection of a proposed site should be based on technical issues, not social ones. Professionals in this field should present papers identifying the merits of HLW disposal in their own state. The scarcity of such research suggests Not In My Back Yard (NIMBY) syndrome. - Medical diagnostic ionizing radiation exposure to the US public is now 8,000 times greater than radiation exposure from nuclear energy. People accept this believing the benefits outweigh any risks. A major effort needs to focus on both benefits as well as risks of radioactive waste disposal. - DOE needs to announce preferences of host rock formations, incentives for states, and potential consequences should we fail to act. (author)

  11. Production of aluminum orthophosphate and basic aluminum polyphosphate under hydrothermal conditions

    Science.gov (United States)

    Adkhamov, A. A.; Iaroslavskii, I. M.; Popolitov, V. I.; Umarov, B. S.; Iliaev, A. B.

    Berlinite (AlPO4) crystals, which are used in piezoelectronic devices, have been produced by hydrothermal synthesis using the methods proposed by Stanley (1954) and Kolb and Laudise (1978). Also, the possibility of AlPO4 crystallization from metastable aluminophosphate glass has been investigated. It is found that berlinite can be crystallized by slowly raising the temperature in the retrograde solubility region; the crystal growth temperature can be reduced by using metastable aluminophosphate glass. Basic aluminum polyphosphate crystals, which decompose with the formation of Al(PO3)3, have been produced and investigated.

  12. Development of Crystal-Tolerant High-Level Waste Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Matyas, Josef; Vienna, John D.; Schaible, Micah J.; Rodriguez, Carmen P.; Crum, Jarrod V.; Arrigoni, Alyssa L.; Tate, Rachel M.

    2010-12-17

    Twenty five glasses were formulated. They were batched from HLW AZ-101 simulant or raw chemicals and melted and tested with a series of tests to elucidate the effect of spinel-forming components (Ni, Fe, Cr, Mn, and Zn), Al, and noble metals (Rh2O3 and RuO2) on the accumulation rate of spinel crystals in the glass discharge riser of the high-level waste (HLW) melter. In addition, the processing properties of glasses, such as the viscosity and TL, were measured as a function of temperature and composition. Furthermore, the settling of spinel crystals in transparent low-viscosity fluids was studied at room temperature to access the shape factor and hindered settling coefficient of spinel crystals in the Stokes equation. The experimental results suggest that Ni is the most troublesome component of all the studied spinel-forming components producing settling layers of up to 10.5 mm in just 20 days in Ni-rich glasses if noble metals or a higher concentration of Fe was not introduced in the glass. The layer of this thickness can potentially plug the bottom of the riser, preventing glass from being discharged from the melter. The noble metals, Fe, and Al were the components that significantly slowed down or stopped the accumulation of spinel at the bottom. Particles of Rh2O3 and RuO2, hematite and nepheline, acted as nucleation sites significantly increasing the number of crystals and therefore decreasing the average crystal size. The settling rate of ≤10-μm crystal size around the settling velocity of crystals was too low to produce thick layers. The experimental data for the thickness of settled layers in the glasses prepared from AZ-101 simulant were used to build a linear empirical model that can predict crystal accumulation in the riser of the melter as a function of concentration of spinel-forming components in glass. The developed model predicts the thicknesses of accumulated layers quite well, R2 = 0.985, and can be become an efficient tool for the formulation

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

    International Nuclear Information System (INIS)

    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

  14. High level radioactive waste glass production and product description

    International Nuclear Information System (INIS)

    This report examines borosilicate glass as a means of immobilizing high-level radioactive wastes. Borosilicate glass will encapsulate most of the defense and some of the commercial HLW in the US. The resulting waste forms must meet the requirements of the WA-SRD and the WAPS, which include a short term PCT durability test. The waste form producer must report the composition(s) of the borosilicate waste glass(es) produced but can choose the composition(s) to meet site-specific requirements. Although the waste form composition is the primary determinant of durability, the redox state of the glass; the existence, content, and composition of crystals; and the presence of glass-in-glass phase separation can affect durability. The waste glass should be formulated to avoid phase separation regions. The ultimate result of this effort will be a waste form which is much more stable and potentially less mobile than the liquid high level radioactive waste is currently

  15. Increasing High-Level Waste Loading In Glass Without Changing The Baseline Melter Technology

    Energy Technology Data Exchange (ETDEWEB)

    Hrma, Pavel R.; Alton, Jesse; Plaisted, Trevor J.; Klouzek, Jaroslav; Matyas, Josef; Mika, Martin; Schill, Petr; Trochta, Miroslav; Nemec, Lubomir

    2001-02-25

    The main factors that determine the cost of high-level waste (HLW) vitrification are the waste loading (which determines the volume of glass) and the melting rate. Product quality should be the only factor determining the waste loading while melter design should provide a rapid melting technology. In reality, the current HLW melters are slow in glass-production rate and are subjected to operational risks that require waste loading to be kept far below its intrinsic level. One of the constraints that decrease waste loading is the liquidus-temperature limit. close inspection reveals that this constraint is probably too severe, even for the current technology. The purpose of the liquidus-temperature constraint is to prevent solids from settling on the melter bottom. It appears that some limited settling would niether interfere with melter operation nor shorten its lifetime and that the rate of settling can be greatly reduced if only small crystals are allowed to form.

  16. 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.

  17. Glass Glimpsed

    DEFF Research Database (Denmark)

    Lock, Charles

    2015-01-01

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

  18. Cosmos & Glass

    DEFF Research Database (Denmark)

    Beim, Anne

    1996-01-01

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

  19. Research on HLW and ILW Interim Storage Complementary to Deep Geological Disposal

    International Nuclear Information System (INIS)

    According to the Act (28 June 2006), interim storage is a research route for the sustainable management of HLW and intermediate level long lived radioactive waste (ILW) in France, along with partitioning/transmutation and reversible disposal in a deep geological formation. Interim storage is intended to play a complementary role to the geological reversible repository. ANDRA is responsible for defining and coordinating the research on both interim storage and reversible disposal of HLW/ILW. A long storage duration as considered before 2006 (up to 300 years) is no more an objective for the research. The paper details the role of interim storage complementary to the reversible repository, with respect to the origin and characteristics of various HLW and ILW to be considered. In particular, it will make it possible for HLW to benefit of thermal decay. ANDRA cooperates with the operators of existing storage facilities at production sites. Working groups have been created. The objectives are the following: - to learn from the experience gained at designing, building and operating the facilities, - to check the capability of existing and projected facilities to meet with the requirements of the National plan for the management of the radioactive matters and waste. In 2008, French waste producers updated the National inventory of radioactive waste and matters. According to the Act, they have incorporated an inventory of their storage facilities. ANDRA has made a review of these inventories to provide the Government with an evaluation of needs for new facilities. ANDRA has also investigated new design options of storage facilities on both production and disposal sites. On the repository site, interim storage above ground and in shallow geological formation have been studied. The waste packages are stored as produced, eventually secured in handling cask, or over-packed for disposal in steel (HLW) or concrete containers (ILW). A service life as long as one hundred years may

  20. Spin glasses

    CERN Document Server

    Bovier, Anton

    2007-01-01

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

  1. Plutonium immobilization plant using glass in existing facilities at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    DiSabatino, A., LLNL

    1998-06-01

    The Plutonium Immobilization Plant (PIP) accepts plutonium (Pu) from pit conversion and from non-pit sources and, through a glass immobilization process, converts the plutonium into an immobilized form that can be disposed of in a high level waste (HLW) repository. The objective is to make an immobilized form, suitable for geologic disposal, in which the plutonium is as inherently unattractive and inaccessible as the plutonium in spent fuel from commercial reactors.

  2. Embedded adhesive connection for laminated glass plates

    DEFF Research Database (Denmark)

    Hansen, Jens Zangenberg; Poulsen, S.H.; Bagger, A.; Stang, Henrik; Olesen, John Forbes

    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 is...... ensured using a structural adhesive. At first, the elastic and viscoelastic material properties of the adhesive are identified where the influence of load-rate and failure properties are also examined. Through an inverse analysis using the finite element method, the experimental observations are...... replicated to identify a material model of the adhesive. The material model consists of an elastic and linear viscoelastic formulation suitable for a numerical implementation of the material. Based on two relevant load cases, out-of-plane bending and in-plane shear, the connection performance is investigated...

  3. The incorporation of P, S, Cr, F, Cl, I, Mn, Ti, U, and Bi into simulated nuclear waste glasses: Literature study

    International Nuclear Information System (INIS)

    Waste currently stored on the Hanford Reservation in underground tanks will be into High Level Waste (HLW) and Low Level Waste (LLW). The HLW melter will high-level and transuranic wastes to a vitrified form for disposal in a geological repository. The LLW melter will vitrify the low-level waste which is mainly a sodium solution. Characterization of the tank wastes is still in progress, and the pretreatment processes are still under development Apart from tank-to-tank variations, the feed delivered to the HLW melter will be subject to process control variability which consists of blending and pretreating the waste. The challenge is then to develop glass formulation models which can produce durable and processable glass compositions for all potential vitrification feed compositions and processing conditions. The work under HLW glass formulation is to study and model glass and melt pro functions of glass composition and temperature. The properties of interest include viscosity, electrical conductivity, liquidus temperature, crystallization, immiscibility durability. It is these properties that determine the glass processability and ac waste glass. Apart from composition, some properties, such as viscosity are affected by temperature. The processing temperature may vary from 1050 degrees C to 1550 degrees C dependent upon the melter type. The glass will also experience a temperature profile upon cooling. The purpose of this letter report is to assess the expected vitrification feed compositions for critical components with the greatest potential impact on waste loading for double shell tank (DST) and single shell tank (SST) wastes. The basis for critical component selection is identified along with the planned approach for evaluation. The proposed experimental work is a crucial part of model development and verification

  4. The incorporation of P, S, Cr, F, Cl, I, Mn, Ti, U, and Bi into simulated nuclear waste glasses: Literature study

    Energy Technology Data Exchange (ETDEWEB)

    Langowski, M.H.

    1996-02-01

    Waste currently stored on the Hanford Reservation in underground tanks will be into High Level Waste (HLW) and Low Level Waste (LLW). The HLW melter will high-level and transuranic wastes to a vitrified form for disposal in a geological repository. The LLW melter will vitrify the low-level waste which is mainly a sodium solution. Characterization of the tank wastes is still in progress, and the pretreatment processes are still under development Apart from tank-to-tank variations, the feed delivered to the HLW melter will be subject to process control variability which consists of blending and pretreating the waste. The challenge is then to develop glass formulation models which can produce durable and processable glass compositions for all potential vitrification feed compositions and processing conditions. The work under HLW glass formulation is to study and model glass and melt pro functions of glass composition and temperature. The properties of interest include viscosity, electrical conductivity, liquidus temperature, crystallization, immiscibility durability. It is these properties that determine the glass processability and ac waste glass. Apart from composition, some properties, such as viscosity are affected by temperature. The processing temperature may vary from 1050{degrees}C to 1550{degrees}C dependent upon the melter type. The glass will also experience a temperature profile upon cooling. The purpose of this letter report is to assess the expected vitrification feed compositions for critical components with the greatest potential impact on waste loading for double shell tank (DST) and single shell tank (SST) wastes. The basis for critical component selection is identified along with the planned approach for evaluation. The proposed experimental work is a crucial part of model development and verification.

  5. Chemical decomposition of high-level nuclear waste storage/disposal glasses under irradiation. 1998 annual progress report

    International Nuclear Information System (INIS)

    'The objective of this project is to employ the technique of electron spin resonance (ESR), in conjunction with other experimental methods, to study radiation-induced decomposition of vitreous compositions proposed for immobilization/disposal of high-level nuclear wastes (HLW) or excess weapons plutonium. ESR is capable of identifying, even at the parts-per-million level, displaced atoms, ruptured bonds, and free radicals created by radiation in such glassy forms. For example, one of the scientific goals is to determine whether ESR-detectable superoxide (O2-) and ozonide (O3-) ions are precursors of radiation-induced oxygen gas bubbles reported by other investigators. The fundamental understandings obtained in this study will enable reliable predictions of the long-term effects of and decays of the immobilized radionuclides on HLW glasses. This report represents the results of an 18-month effort performed under a 3-year research award. Four categories of materials were studied: (1) several actual and proposed HLW glass compositions fabricated at Savannah River Technology Center (SRTC), samples of which had been irradiated to a dose of 30 MGy (1 Gy = 100 rad) to simulate decay effects, (2) several high-iron phosphate glasses fabricated at the University of Missouri-Rolla (UMR), (3) one other model HLW glass and several simulated natural glasses which had been implanted with 160-keV He+ ions to simulate-decay damage, and (4) an actual geological glass damaged by decays of trace amounts of contained 238U and 232Th over a period of 65 Myears. Among the category-1 materials were two samples of Defense Waste Processing Facility (DWPF) borosilicate glasses modeling compositions currently being used to vitrify HLW at SRTC. The ESR spectra recorded for the unirradiated DWPF-glass simulants were attributable to Fe 3+ ions. The 30-MGy irradiation was found to change the Fe3+ concentration of these glasses by a statistically insignificant factor (0.987 261 0.050) and not to

  6. Chemical decomposition of high-level nuclear waste storage/disposal glasses under irradiation. 1998 annual progress report

    Energy Technology Data Exchange (ETDEWEB)

    Griscom, D.L.; Merzbacher, C.I.

    1998-06-01

    'The objective of this project is to employ the technique of electron spin resonance (ESR), in conjunction with other experimental methods, to study radiation-induced decomposition of vitreous compositions proposed for immobilization/disposal of high-level nuclear wastes (HLW) or excess weapons plutonium. ESR is capable of identifying, even at the parts-per-million level, displaced atoms, ruptured bonds, and free radicals created by radiation in such glassy forms. For example, one of the scientific goals is to determine whether ESR-detectable superoxide (O{sub 2}{sup -}) and ozonide (O{sub 3}{sup -}) ions are precursors of radiation-induced oxygen gas bubbles reported by other investigators. The fundamental understandings obtained in this study will enable reliable predictions of the long-term effects of and decays of the immobilized radionuclides on HLW glasses. This report represents the results of an 18-month effort performed under a 3-year research award. Four categories of materials were studied: (1) several actual and proposed HLW glass compositions fabricated at Savannah River Technology Center (SRTC), samples of which had been irradiated to a dose of 30 MGy (1 Gy = 100 rad) to simulate decay effects, (2) several high-iron phosphate glasses fabricated at the University of Missouri-Rolla (UMR), (3) one other model HLW glass and several simulated natural glasses which had been implanted with 160-keV He{sup +} ions to simulate-decay damage, and (4) an actual geological glass damaged by decays of trace amounts of contained {sup 238}U and {sup 232}Th over a period of 65 Myears. Among the category-1 materials were two samples of Defense Waste Processing Facility (DWPF) borosilicate glasses modeling compositions currently being used to vitrify HLW at SRTC. The ESR spectra recorded for the unirradiated DWPF-glass simulants were attributable to Fe 3{sup +} ions. The 30-MGy irradiation was found to change the Fe{sup 3+} concentration of these glasses by a

  7. Is the Aluminum Hypothesis Dead?

    OpenAIRE

    Lidsky, Theodore I.

    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 w...

  8. Photonic glasses

    CERN Document Server

    Gan, Fuxi

    2006-01-01

    This book introduces the fundamental mechanism of photonic glasses - the linear and nonlinear optical effects in glass under intense light irradiation: phot-induced absorption, refraction, polarization, frequency, coherence and monochromaticity changes. Emphasis is placed on new developments in the structure, spectroscopy and physics of new glassy materials for photonics applications, such as optical communication, optical data storage, new lasers and new photonic components and devices. The book presents the research results of the authors in new glasses for photonics over the last decade. Sa

  9. Anodizing Aluminum with Frills.

    Science.gov (United States)

    Doeltz, Anne E.; And Others

    1983-01-01

    "Anodizing Aluminum" (previously reported in this journal) describes a vivid/relevant laboratory experience for general chemistry students explaining the anodizing of aluminum in sulfuric acid and constrasting it to electroplating. Additions to this procedure and the experiment in which they are used are discussed. Reactions involved are also…

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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. The aluminum smelting process.

    Science.gov (United States)

    Kvande, Halvor

    2014-05-01

    This introduction to the industrial primary aluminum production process presents a short description of the electrolytic reduction technology, the history of aluminum, and the importance of this metal and its production process to modern society. Aluminum's special qualities have enabled advances in technologies coupled with energy and cost savings. Aircraft capabilities have been greatly enhanced, and increases in size and capacity are made possible by advances in aluminum technology. The metal's flexibility for shaping and extruding has led to architectural advances in energy-saving building construction. The high strength-to-weight ratio has meant a substantial reduction in energy consumption for trucks and other vehicles. The aluminum industry is therefore a pivotal one for ecological sustainability and strategic for technological development. PMID:24806722

  13. Iron Phosphate Glass Development and Demonstration (AJHM and CCIM)

    International Nuclear Information System (INIS)

    Iron phosphate glasses retain high concentrations of some waste components that are difficult to dissolve into borosilicate melts - Sulfates, phosphates, heavy metals (Cr, Bi, Mo) and halides (F, Cl). This translates into significant increases in waste loading (WL) or reduced canister counts for specific waste streams. Programmatic objectives are: (1) Develop phosphate glass systems at high waste loadings for Hanford LAW and other DOE HLWs (Hanford and INL); (2) Develop and demonstrate a process flowsheet for implementation of phosphate based glasses in melters (JHM or AJHM) similar to those currently employed at Hanford - Determine if an alternative melter (e.g., CCIM) could be used; (3) Determine potential benefits and costs for implementation of iron phosphate glasses into WTP; and (4) Begin the effort to qualify LAW-based iron phosphate glasses for disposal on site at Hanford. FY10 activities primarily focused on developing iron phosphate glasses for Hanford LAW immobilization. Iron phosphate glasses have been shown to retain high concentrations of some waste components that are difficult to dissolve into borosilicate melts - Sulfates, phosphates, heavy metals (Cr, Bi, Mo) and halides (F, Cl). FeP glasses may offer significant increases in waste loading (reduced canister counts) for specific waste streams. EM-31 FeP Team has developed an integrated program to assess the impact of implementing the FeP glass system on Hanford LAW/HLW vitrification facilities: (a) Glass formulation and optimization → high WLs for various waste streams, Hanford AZ-102 LAW primary focus of FY10; (b) Systems analysis → glass volumes or mass; (c) WTP flowsheet impacts → storage and transport issues; and (d) Long-term performance → support PA.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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 9living 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

  16. CLOSURE OF HLW TANKS PHASE 2 FULL SCALE COOLING COILS GROUT FILL DEMONSTATIONS

    International Nuclear Information System (INIS)

    This report documents the Savannah River National Laboratory (SRNL) support for the Tank Closure and Technology Development (TCTD) group's strategy for closing high level radioactive waste (HLW) tanks at the Savannah River Site (SRS). Specifically, this task addresses the ability to successfully fill intact cooling coils, presently within the HLW tanks, with grout that satisfies the fresh and cured grout requirements [1] under simulated field conditions. The overall task was divided into two phases. The first phase was the development of a grout formulation that satisfies the processing requirements for filling the HLW tank cooling coils [5]. The second phase of the task, which is documented in this report, was the filling of full scale cooling coils under simulated field conditions using the grout formulation developed in the first phase. SRS Type I tank cooling coil assembly design drawings and pressure drop calculations were provided by the Liquid Waste (LW) customer to be used as the basis for configuring the test assemblies. The current concept for closing tanks equipped with internal cooling coils is to pump grout into the coils to inhibit pathways for infiltrating water. Access to the cooling coil assemblies is through the existing supply/return manifold headers located on top of the Type I tanks. The objectives for the second phase of the testing, as stated in the Task Technical and Quality Assurance plan (TTQAP) [2], were to: (1) Perform a demonstration test to assess cooling coil grout performance in simulated field conditions, and (2) Measure relevant properties of samples prepared under simulated field conditions. SRNL led the actual work of designing, fabricating and filling two full-scale cooling coil assemblies which were performed at Clemson Engineering Technologies Laboratory (CETL) using the South Carolina University Research and Education Foundation (SCUREF) program. A statement of work (SOW) was issued to CETL [6] to perform this work

  17. A decision-making students project for site selection of HLW repository

    International Nuclear Information System (INIS)

    With the escalating costs, constant delays and adverse public opinion, the YMP is not only under fire but the DOE and the other parties involved have failed to convince the public that they know what they are doing and that theirs is the best way. For example, while the whole world is evaluating alternative HLW disposal sites, why is the United States characterizing just one site? A student class at Penn State decided that they may have a better concept of what should be done

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

    International Nuclear Information System (INIS)

    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

  19. Synthesis of SrTiO3 for immobilization of simulated HLW by SHS

    Institute of Scientific and Technical Information of China (English)

    2005-01-01

    Strontium titanate synroc samples were synthesized by self-propagating high-temperature synthesis (SHS). Sr directly took part in the synthesis process. As a result, the loading content issue is basically resolved. The products were characterized by density, microhardness X-ray diffraction, and scanning electron microscopy (SEM/EDS). The leaching rate was measured by the method of PCT (product consistency test). The results indicate that the Sr2+-SrTiO3 compound is of high density, low leach rate and high stability and the synthesis process is feasible in technology and economy. It can be concluded that the strontium titanate synroc is a perfect material to immobilize HLW.

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Aurah, Mirwaise Y.; Roberts, Mark A.

    2013-12-12

    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.

  2. Effect of natural and synthetic iron corrosion products on silicate glass alteration processes

    Science.gov (United States)

    Dillmann, Philippe; Gin, Stéphane; Neff, Delphine; Gentaz, Lucile; Rebiscoul, Diane

    2016-01-01

    Glass long term alteration in the context of high-level radioactive waste (HLW) storage is influenced by near-field materials and environmental context. As previous studies have shown, the extent of glass alteration is strongly related to the presence of iron in the system, mainly provided by the steel overpack around surrounding the HLW glass package. A key to understanding what will happen to the glass-borne elements in the geological disposal lies in the relationship between the iron-bearing phases and the glass alteration products formed. In this study, we focus on the influence of the formation conditions (synthetized or in-situ) and the age of different iron corrosion products on SON68 glass alteration. Corrosion products obtained from archaeological iron artifacts are considered here to be true analogues of the corrosion products in a waste disposal system due to the similarities in formation conditions and physical properties. These representative corrosion products (RCP) are used in the experiment along with synthetized iron anoxic corrosion products and pristine metallic iron. The model-cracks of SON68 glass were altered in cell reactors, with one of the different iron-sources inserted in the crack each time. The study was successful in reproducing most of the processes observed in the long term archaeological system. Between the different systems, alteration variations were noted both in nature and intensity, confirming the influence of the iron-source on glass alteration. Results seem to point to a lesser effect of long term iron corrosion products (RCP) on the glass alteration than that of the more recent products (SCP), both in terms of general glass alteration and of iron transport.

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

    Energy Technology Data Exchange (ETDEWEB)

    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, {sup 99}Tc, {sup 237}Np, {sup 239}Pu, and {sup 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{sup 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-08

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

  5. Burning characteristics of individual aluminum/aluminum oxide particles

    OpenAIRE

    Ruttenberg, Eric C.

    1996-01-01

    Approved for public release; distribution is unlimited An experimental investigation was conducted in which the burning characteristics of individual aluminum/aluminum oxide particles were measured using a windowed combustion bomb at atmospheric pressure and under gravity-fall conditions. A scanning electron microscope (SEM) was used to measure the size distribution of the initial aluminum particles and the aluminum oxide residue. Analysis of the residue indicated that the mass of aluminum...

  6. Purifying Aluminum by Vacuum Distillation

    Science.gov (United States)

    Du Fresne, E. R.

    1985-01-01

    Proposed method for purifying aluminum employs one-step vacuum distillation. Raw material for process impure aluminum produced in electrolysis of aluminum ore. Impure metal melted in vacuum. Since aluminum has much higher vapor pressure than other constituents, boils off and condenses on nearby cold surfaces in proportions much greater than those of other constituents.

  7. Modeling of Spinel Settling in Waste Glass Melter

    Energy Technology Data Exchange (ETDEWEB)

    Hrma, Pavel R.; Nemec, Lubomir; Schill, Petr

    1999-06-01

    Each 1% increase of waste loading (W), defined as the high-level waste (HLW) mass fraction in glass, can save the U.S. Department of Energy (DOE) over a half billion U.S. dollars for vitrification and disposal. For a majority of Hanford and Savannah River waste streams, W is limited by spinel precipitation and settling in waste glass melters. Therefore, a fundamental understanding of spinel behavior is crucial for economy and the low-risk operation of HLW vitrification. The goal of this research is to develop a basic understanding of the dynamics of spinel formation and motion in velocity, temperature, and redox fields that are characteristic for the glass-melting process. This goal is being achieved by directly studying spinel formation and settling in molten glass and by developing a mathematical tool for predicting the spinel behavior and accumulation rate in the melter. The main potential benefit of this study is achieving a lower waste-glass volume, which translates into a shorter cleanup time, a smaller processing facility, a smaller repository space, and, hence, a reduced investment of time and money to reach acceptable technical risks. Additional benefits include (1) more accurately assessing sensible limits for problem constituents (such as chromium) in the melter feed, (2) reducing the blending requirements, and (3) comparing cost and risk with other options (pretreatment, blending or diluting the waste) to determine the best path forward. The results of this study will allow alternate melter designs and operating conditions to be evaluated. The study will also address the option of removing the settled sludge from the melter.

  8. Separation of actinides from sulphate-bearing HLW solutions using CMPO-extraction and extraction chromatographic studies

    International Nuclear Information System (INIS)

    Partitioning of actinides from sulphate-bearing synthetic and actual Purex HLW solutions by solvent extraction and extraction chromatography has been carried out using CMPO. A pre-treatment of the HLW solution with 30% TBP was essential to deplete the uranium content. The gross alpha in the raffinate was about 0.4% while using solvent extraction and actual sulphate bearing HLW (SBHLW). The extraction chromatographic technique has been applied with synthetic SBHLW solution. The species extracted into the CMPO phase from sulphate bearing solutions was only the nitrato complex, such as Am(NO3)3.3CMPO, UO2(NO3)2.2CMPO and Pu(NO3)4.2CMPO. (author). 3 refs., 1 tab

  9. A Study on the Development of the FEP and Scenario for the HLW Disposal in Korea

    International Nuclear Information System (INIS)

    The impacts influenced on the performance and safety of a repository are classified as units of Features, Events, and Processes (FEP), for the total system performance assessment (TSPA) related to the permanent disposal of HLW. The importance is evaluated in consideration of the frequency, consequence, regulation, suitability of a specific site, etc. and then these are grouped as a similar FEP. A scenario describing the migration of radionuclide from the repository to the biosphere is derived from understanding the interaction among these groups. KAERI has developed the KAERI FEP lists by review and collation of the foreign studies. The KAERI FEP list has been reviewed by several Korean experts. The five major scenarios describing possible future evolutions of the geological disposal system have been developed by RES and PID methods. Also the CYPRUS which is a KAERI integrated database management system for the total system performance assessment (TSPA) related to the permanent disposal of HLW has been developed and the results of the FEP and scenario development have been uploaded in this system.

  10. Strategies for technical confidence building in the NUMO HLW disposal programme

    International Nuclear Information System (INIS)

    Building confidence in the safety of geological disposal is a challenge faced by all waste management programmes. It is recognized that confidence building measures are needed for interacting with the technical community and also with the general public. Although both aspects are critical and strongly overlap, this paper addresses primarily the strategies for enhancing technical confidence. This is an absolute pre-requisite for engaging a wider public. Unless and until the scientific and technical community actually engaged in implementing repositories has great confidence in their feasibility and safety - and can successfully and convincingly communicate this in the wider scientific circles - then it is premature to take the debate into the public arena. The Japanese HLW disposal organization, NUMO was created only relatively recently (2000). It instantly faced the challenge of initiating one of the largest HLW disposal programmes in the world - and it has chosen to tackle the key challenge of identifying potential repository sites by using a completely transparent volunteering approach. This paper describes the measures that NUMO has taken to establish its scientific credibility at home and abroad. (author)

  11. Some illustrative examples from parametric studies for a reference HLW repository using ACGEO

    International Nuclear Information System (INIS)

    Spent fuels are to be directly disposed of and the only type of HLW in Korea. According to the basic repository design, which is similar to Swedish KBS-3 repository concept, spent fuel is encapsulated in corrosion resistant canisters, and then emplaced into the deposition holes surrounded by high density bentonite clay in tunnels at a depth of ∼500 m in a plutonic rock (Fig. 1). The objective of this paper is to show how we could show reliability of radiological safety assessment through probabilistic calculation with parameters having uncertainties associated with safety of HLW system. To demonstrate a typical way, possibly among many others, some probabilistic calculation results for nuclide release and transport in the near- and farfield of the repository when such selected key parameters as buffer thickness, nuclide travel length in the fractured rock medium, and initial number of defective canisters are varying are presented. ACGEO, a case file developed by utilizing AMBER, a compartment modeling software package was used for calculation

  12. Developing an advanced safety concept for an HLW repository in salt rock

    International Nuclear Information System (INIS)

    Preliminary Total Performance Assessments for HLW repositories in salt rock were performed within the framework of German R and D projects at the end of the 1980 and in the first half of the 1990's. In the meantime, several remarkable developments improved the basis for developing an advanced safety case. In view of these changes, DBE TECHNOLOGY GmbH, BGR and GRS are developing and testing an advanced safety concept within a joint R and D project in order to identify the major needs for further R and D work. Whilst former safety assessment approaches for HLW repositories in salt rock used a conservative, even hypothetical, release scenario in order to show the compliance with dose constraints, the recent safety concept focuses on proving the safe enclosure without radionuclide release if the repository evolution remains undisturbed. Release scenarios are considered in the case of disturbed evolution only. This approach is considered to be more appropriate for a salt rock repository and takes advantage of its specific properties. In this context, special attention is given to the proof of integrity (sufficient tightness) of the multi-barrier system, comprised mainly of the host rock as the main geological barrier and the engineered geotechnical barriers. The approach also focuses on appropriate scenario analyses that allow the identification and assessment of remaining radionuclide release scenarios which cannot be ruled out. (A.L.B.)

  13. Effects of nitrate on nuclide solubility for co-location disposal of TRU waste and HLW

    International Nuclear Information System (INIS)

    Part of TRU waste includes a large amount of nitrate salt, the effects of which have to be evaluated in a safety assessment of co-location disposal with high level radioactive waste (HLW). High concentrations of nitrate ions from TRU waste might affect the solubility of different radionuclides in the HLW. In the current study, the effects of nitrate salt on radionuclide solubility were investigated experimentally. Solubility experiments of important and redox sensitive radionuclides, Tc(IV), Np(IV) and Se(0), were performed using various concentrations of sodium nitrate (NaNO3) and of Np(V) in NaNO3 solutions to investigate complex formation with NO3- ions. Solubility experiments of Pd(II), Sn(IV) and Nb(V) using ammonium chloride (NH4Cl) solution were also undertaken to investigate complex formation with NH3/NH4+ ions. No significant solubility enhancement was observed for Np and Se. Tc solubility in ≥0.1 mol/dm3 NaNO3 solution increased due to oxidation by nitrate ions. An increase of Np(V) solubility was expected by the chemical equilibrium model calculation with JNC-TDB, however, solubility enhancement by complex formation of Np(V) with nitrate ions was not observed. Solubility enhancement by complex formation of Sn and Nb were also not observed, only Pd solubility was increased by complex formation with NH3/NH4+ ions. (author)

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

    International Nuclear Information System (INIS)

    The Power Reactor and Nuclear Fuel Development Corporation (PNC) has conducted research and development on geologic disposal of high level radioactive waste (HLW) as a leading organization in accordance with the overall management program defined by the Atomic Energy Commission (AEC) of Japan. Fundamental features of the R and D carried out by PNC at this stage are: Site generic research and development, establishment of scientific and technical bases of geologic isolation of HLW in Japan, about 15 years program from 1989 with documentation of progress reports, and approach from near-field to far-field. PNC summarized the results and submitted the first progress report (H3 Report) in September 1992 on three subjects: geologic environment studies, disposal technology development and performance assessment studies. Overview of this comprehensive report and results of reviews and evaluation for the H3 Report by AEC of Japan are described. Overall procedures and schedule for the geologic disposal are described. The main target of the future research is a detailed analysis of the near-field performance. (T.H.)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  17. DEVELOPMENT OF CASTING TECHNOLOGIES DURING FORMATION OF PROPERTIES OF ALUMINUM-BASED MATERIALS WITH CARBON OF DIFFERENT STRUCTURAL CONDITION

    OpenAIRE

    A. T. Volochko

    2015-01-01

    The paper gives an assessment of existing casting methods used for manufacturing products from aluminum materials with carbon filling compounds. It presents results of comparative studies of properties of aluminum materials in which microcrystalline graphite, fullerene black, nanotubes and an amorphous phase of glass carbon have been used as filling compounds.

  18. DEVELOPMENT OF CASTING TECHNOLOGIES DURING FORMATION OF PROPERTIES OF ALUMINUM-BASED MATERIALS WITH CARBON OF DIFFERENT STRUCTURAL CONDITION

    Directory of Open Access Journals (Sweden)

    A. T. Volochko

    2015-11-01

    Full Text Available The paper gives an assessment of existing casting methods used for manufacturing products from aluminum materials with carbon filling compounds. It presents results of comparative studies of properties of aluminum materials in which microcrystalline graphite, fullerene black, nanotubes and an amorphous phase of glass carbon have been used as filling compounds.

  19. Corrosion Inhibitors for Aluminum.

    Science.gov (United States)

    Muller, Bodo

    1995-01-01

    Describes a simple and reliable test method used to investigate the corrosion-inhibiting effects of various chelating agents on aluminum pigments in aqueous alkaline media. The experiments that are presented require no complicated or expensive electronic equipment. (DDR)

  20. Advances in aluminum anodizing

    Science.gov (United States)

    Dale, K. H.

    1969-01-01

    White anodize is applied to aluminum alloy surfaces by specific surface preparation, anodizing, pigmentation, and sealing techniques. The development techniques resulted in alloys, which are used in space vehicles, with good reflectance values and excellent corrosive resistance.

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

    International Nuclear Information System (INIS)

    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

  2. Effects of clay on the dissolution behaviour of nuclear waste glass

    International Nuclear Information System (INIS)

    In the concept for the geological disposal of vitrified nuclear HLW (high-level waste), the biosphere is protected from the waste by multiple barriers, each with their specific safety functions. A first barrier is the waste matrix itself, which should be resistant against leaching, resulting in a slow radionuclide release into the surrounding host rock. As a second barrier, the previous reference disposal design for HLW glass in Belgium foresaw a bentonite buffer. The third barrier would be the Boom Clay host rock. Therefore, our laboratory has performed many experiments with glass in contact with bentonite or Boom Clay, making from the interaction of waste glass and clay an important field of expertise. The objectives of the studies on the stability of waste glass in conditions representative of the relevant disposal concept, are to create a database for the evaluation of the radiological long-term safety, and to validate certain simplified hypotheses in these evaluations, which allow the modelling of the glass dissolution behaviour

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

    International Nuclear Information System (INIS)

    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.

  4. 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.

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

    OpenAIRE

    Basel A. Khader; Declan J. Curran; Sean Peel; Mark R. Towler

    2016-01-01

    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 applicat...

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

    International Nuclear Information System (INIS)

    Molybdenum oxide immobilization (MoO3, 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 MoO3 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 SiO2-B2O3- Na2O-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 (95Mo, 29Si, 11B, 23Na 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 (CaMoO4, Na2MoO4) that depend on the evolution of MoO3, CaO and B2O3 contents were correlated with changes of sodium and calcium cations proportions in the environment of molybdate

  7. Fluorescent lighting with aluminum nitride phosphors

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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. 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-theshelf 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

  10. Fabrication of aluminum foam from aluminum scrap Hamza

    OpenAIRE

    O. A. Osman1 ,; Mining and Petroleum Engineering, Faculty of Engineering- Qena, Al_Azhar University, Egypt

    2015-01-01

    In this study the optimum parameters affecting the preparation of aluminum foam from recycled aluminum were studied, these parameters are: temperature, CaCO3 to aluminum scrap wt. ratio as foaming agent, Al2O3 to aluminum scrap wt. ratio as thickening agent, and stirring time. The results show that, the optimum parameters are the temperature ranged from 800 to 850oC, CaCO3 to aluminum scrap wt. ratio was 5%, Al2O3 to aluminum scrap wt. ratio was 3% and stirring time was 45 second ...

  11. Immobilization of simulated high-level waste in sintered glasses

    International Nuclear Information System (INIS)

    The vitrification of simulated HLW from the reprocessing of PHWR fuel was studied on a laboratory scale by the alternative method of powder glass sintering. The HLW was added in a proportion of 10 wt% to aluminoborosilicate glass powder. The powder mixture was treated using the procedure of cold pressing and subsequent sintering at about 1000 K. The microstructures of the sintered samples were analyzed by optical microscopy and SEM. Leaching tests were done in deionized water using the MCC-1 method (Ref. [1], D.M. Strachan, R.P. Turcotte and B.O. Barnes, Nucl. Technol. 56 (1982) 306) at 343 K between 7 and 300 days. The leaching rates were of the order of 10-2 g m-2 d-1. Devitrification studies were conducted by heat-treating the samples at 950, 1000, 1050 and 1100 K for 3, 30, 300 and 3000 h. TTT type curves were obtained, showing that the maximum devitrification rate occurs at 1000 K. The samples present 53% of crystallinity after 3000 h at this temperature. Leaching rate measurements on devitrified samples did not show a significant change of this property in relation with the values for the as-sintered samples. ((orig.))

  12. Vitrification of high chrome oxide nuclear waste in iron phosphate glasses

    International Nuclear Information System (INIS)

    A simulated high level waste (HLW) containing 4 mass% chrome oxide, whose overall composition is representative of the high chrome oxide wastes at Hanford WA USA, was easily vitrified in a phosphate glass at temperatures ranging from 1150 deg. C, for waste loadings of 55 mass%, to 1250 deg. C for waste loadings of 75 mass%. Even at these high waste loadings, these wasteforms had an excellent chemical durability. The best chemical durability was achieved when the O/(Si + P) atomic ratio was between 3.5 and 3.8. These wasteforms were also resistant to crystallization although trace amounts of crystalline Cr2O3 were present in wasteforms containing more than 70 mass% HLW. It is concluded that up to 45 mass% of the total HLW at Hanford, especially that containing as high as 4.5 mass% chrome oxide, could be directly vitrified into an iron phosphate glass, that meets all of the current chemical durability requirements by simply adding 25-35 mass% P2O5 to the waste and melting the mixture at 1150-1250 deg. C for a few (<6) hours

  13. Test of lead glass shower counters

    International Nuclear Information System (INIS)

    Lead glass counters made of wedge shaped blocks of SF6 were tested with positrons at SLAC. The beam energy ranged from 2 to 17.5 GeV. Energy dependence and beam position dependence of pulse height and energy resolution were studied with lead glass blocks of various lengths. The effect of a BK-7 light guide on pulse height was clearly observed. Degradation of the energy resolution due to aluminum absorbers of various lengths was investigated. A mesh type photomultiplier was also tested

  14. Report - Melter Testing of New High Bismuth HLW Formulations VSL-13R2770-1

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-13

    The primary objective of the work described was to test two glasses formulated for a high bismuth waste stream on the DM100 melter system. Testing was designed to determine processing characteristics and production rates, assess the tendency for foaming, and confirm glass properties. The glass compositions tested were previously developed to maintain high waste loadings and processing rates while suppressing the foaming observed in previous tests

  15. 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.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-12-29

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

  18. Transmission Properties of a New Glass Ceramic and Doped with Co2+ as Saturable Absorber for 1.54 μm Er Glass Short Pulse Laser

    Institute of Scientific and Technical Information of China (English)

    YU Chunlei; CHEN Li; FENG Suya; HE Dongbing; WANG Meng; HU Lili

    2012-01-01

    The preparation and characteristics of a new transparent glass ceramic were described.Crystal phase particles with nanometer size were successfully precipitated in glass matrix,which was confirmed to be one of indium aluminum zinc oxide compounds (InxAlyZnzO).The presence of aluminum (Al) and indium (In) impurities in the zinc oxides (ZnO) crystal lattice leads to some changes of the carrier concentration in the material and then promote the sharply changes of transmission spectra in IR range wavelength.And subsequently,the IR cut-off edge blue shifted from 5.5 μm in base glass to 3 μm in transparent glass ceramic sample.Furthermore,passive Q switched 1.54 μm Er glass laser pulses with pulse energy of 10 mJ and pulse width of 800 ns were successfully obtained by using the cobalt doped transparent glass ceramic as a saturable absorber.

  19. Oriented aluminum nanocrystals in a one-step process

    International Nuclear Information System (INIS)

    Aluminum coatings were deposited on glass substrates by chemical vapor deposition using N-methylpiperidine (nmp) stabilized dichloroalane [Cl2AlH·2nmp] as aluminum precursor. With regard to temperature, the experimental conditions were varied between 75 °C and 125 °C for the precursor and between 250 °C and 450 °C for the substrate. Depending on these parameters, highly textured layers could be deposited. The substrates have been consistently covered by a layer of idiomorphic, mostly distorted octahedra of aluminum single crystals. The morphologies of the structures and the degree of orientation of the crystals were investigated by a scanning electron microscopy and X-ray diffraction measurements. The high order of [111] orientation was found to decrease with increasing precursor and substrate temperature. We propose a mechanism for the generation of the octahedral structures based on the formation of mesocrystals. On heating, the dichloroalane (stabilized with nmp) loses the nmp ligands together with hydrogen and chlorine. The amine (nmp) seems to trigger the formation of aluminum crystals depending on the temperature and thus influences the texture of the Al-layer and the formation of well-formed octahedron-like structures. - Highlights: • An original chemical gas phase synthesis of aluminum single crystals is shown. • N-methylpiperidine triggers a preferred crystal growth in [111] direction. • A decomposition mechanism of the dichloroalane adduct is discussed. • Agglomeration to larger aluminum meso-crystals is observed at a higher gas flow

  20. Extended Development Work to Validate a HLW Calcine Waste Form via INL's Cold Crucible Induction Melter

    International Nuclear Information System (INIS)

    To accomplish calcine treatment objectives, the Idaho Clean-up Project contractor, CWI, has chosen to immobilize the calcine in a glass-ceramic via the use of a Hot-Isostatic-Press (HIP); a treatment selection formally documented in a 2010 Record of Decision (ROD). Even though the HIP process may prove suitable for the calcine as specified in the ROD and validated in a number of past value engineering sessions, DOE is evaluating back-up treatment methods for the calcine as a result of the technical, schedule, and cost risk associated with the HIPing process. Consequently DOE HQ has requested DOE ID to make INL's bench-scale cold-crucible induction melter (CCIM) available for investigating its viability as a process alternate to calcine treatment. The waste form is the key component of immobilization of radioactive waste. Providing a solid, stable, and durable material that can be easily be stored is the rationale for immobilization of radioactive waste material in glass, ceramic, or glass-ceramics. Ceramic waste forms offer an alternative to traditional borosilicate glass waste forms. Ceramics can usually accommodate higher waste loadings than borosilicate glass, leading to smaller intermediate and long-term storage facilities. Many ceramic phases are known to possess superior chemical durability as compared to borosilicate glass. However, ceramics are generally multiphase systems containing many minor phase that make characterization and prediction of performance within a repository challenging. Additionally, the technologies employed in ceramic manufacture are typically more complex and expensive. Thus, many have proposed using glass-ceramics as compromise between in the more inexpensive, easier to characterize glass waste forms and the more durable ceramic waste forms. Glass-ceramics have several advantages over traditional borosilicate glasses as a waste form. Borosilicate glasses can inadvertently devitrify, leading to a less durable product that could crack

  1. The effect of inversion of matrix and inclusions composition in liquation phospho-silicate glasses.

    Science.gov (United States)

    Sitarz, M

    2011-08-15

    Silico-phosphate glasses of XCaPO(4)-SiO(2) and XCaPO(4)-AlPO(4)-SiO(2) (X=Na(+) and/or K(+)) system have been the subject of the presented investigations. Glasses belonging to those systems are characterized by a liquation phenomenon-spherical amorphous inclusions dispersed in an amorphous matrix. Thorough EDX investigations have shown that introduction of aluminum ions into the structure of phospho-silicate glasses results in inversion of matrix and inclusions composition, when XCaPO(4) exceeds 25-35% mol. Such a substantial influence of aluminum ions on phospho-silicate glasses texture as well as matrix and inclusions composition (inversion) must be a result of structural changes. (27)Al MAS NMR research stated that aluminum ions in structures of XCaPO(4)-AlPO(4)-SiO(2) phospho-silicate glasses always acts as a glass-forming ion-i.e. aluminum always occupies fourfold coordinated sites. (23)Na and (31)P MAS NMR research has shown that the inversion of matrix and inclusions composition, brought about by introduction of aluminum ions into the structure of phospho-silicate glasses, is an outcome of a change in phosphorous and alkaline ions coordination. PMID:20864392

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

    International Nuclear Information System (INIS)

    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

  3. Radiation Characteristics of Glass Containing Gas Bubbles

    OpenAIRE

    Pilon, Laurent; Viskanta, Raymond

    2003-01-01

    In many materials processing and manufacturing situations such as steel, aluminum, ceramics and glass, gas bubbles can form in liquid and solid phases. The presence of such bubbles affects the thermophysical properties and radiation characteristics of the two-phase system and hence the transport phenomena. This paper presents a general formulation of the radiation characteristics of semitransparent media containing large gas bubbles (bubble radius is much larger than the wavelength of radiati...

  4. Chemical Decomposition of High-Level Nuclear Waste Storage/Disposal Glasses Under Irradiation

    International Nuclear Information System (INIS)

    The objective of this project is to employ the technique of electron spin resonance (ESR), in conjunction with other experimental methods, to study radiation-induced decomposition of vitreous compositions proposed for immobilization/disposal of high-level nuclear wastes (HLW) or excess weapons plutonium. ESR is capable of identifying, even at the parts-per-million level, displaced atoms, ruptured bonds, and free radicals created by radiation in such glassy forms. For example, one of the scientific goals is to search for ESR-detectable superoxide (O2 -) and ozonide (O3 -) ions, which could be precursors of radiation-induced oxygen gas bubbles reported by other investigators via the disproportionation reaction, 2O2 - : O2 2- + O2. The fundamental understandings obtained in this study will enable reliable predictions of the long-term effects of a and B decays of the immobilized radionuclides on the chemical integrity of HLW glasses

  5. Chemical Decomposition of High-Level Nuclear Waste Storage/Disposal Glasses Under Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Griscom, David L.

    1999-06-01

    The objective of this project is to employ the technique of electron spin resonance (ESR), in conjunction with other experimental methods, to study radiation-induced decomposition of vitreous compositions proposed for immobilization/disposal of high-level nuclear wastes (HLW) or excess weapons plutonium. ESR is capable of identifying, even at the parts-per-million level, displaced atoms, ruptured bonds, and free radicals created by radiation in such glassy forms. For example, one of the scientific goals is to search for ESR-detectable superoxide (O2 -) and ozonide (O3 -) ions, which could be precursors of radiation-induced oxygen gas bubbles reported by other investigators via the disproportionation reaction, 2O2 - : O2 2- + O2. The fundamental understandings obtained in this study will enable reliable predictions of the long-term effects of a and B decays of the immobilized radionuclides on the chemical integrity of HLW glasses.

  6. Fabrication of aluminum foam from aluminum scrap Hamza

    Directory of Open Access Journals (Sweden)

    O. A. Osman1 ,

    2015-02-01

    Full Text Available In this study the optimum parameters affecting the preparation of aluminum foam from recycled aluminum were studied, these parameters are: temperature, CaCO3 to aluminum scrap wt. ratio as foaming agent, Al2O3 to aluminum scrap wt. ratio as thickening agent, and stirring time. The results show that, the optimum parameters are the temperature ranged from 800 to 850oC, CaCO3 to aluminum scrap wt. ratio was 5%, Al2O3 to aluminum scrap wt. ratio was 3% and stirring time was 45 second with stirring speed 1200 rpm. The produced foam apparent densities ranged from 0.40-0.60 g/cm3. The microstructure of aluminum foam was examined by using SEM, EDX and XRD, the results show that, the aluminum pores were uniformly distributed along the all matrices and the cell walls covered by thin oxide film.

  7. ALUMINUM RECLAMATION BY ACIDIC EXTRACTION OF ALUMINUM-ANODIZING SLUDGES

    Science.gov (United States)

    Extraction of aluminum-anodizing sludges with sulfuric acid was examined to determine the potential for production of commercial-strength solutions of aluminum sulfate, that is liquid alum. The research established kinetic and stoichiometric relationships and evaluates product qu...

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

    International Nuclear Information System (INIS)

    Highlights: ► Bi in high level nuclear waste glasses was of interest due to melt foaming issues. ► Bi was also found associated with phosphate in some samples. ► X-ray absorption spectroscopy found similar Bi bonding within all glasses studied. ► The glasses contain Bi3+O3 environments with average Bi–O distances near 2.13 Å. ► No Bi-phosphate glass domains nor any link between Bi and melt foaming were observed. - Abstract: X-ray absorption spectra (XAS) were collected and analyzed to characterize bismuth (Bi) environments in borosilicate glass formulations developed for the immobilization of high level nuclear wastes (HLW), from the bismuth phosphate process. Therefore, the structural role of Bi in these glasses is of interest; in addition in the present study, more particular interest in Bi originated from unusual foaming that was observed during melt cooling, where it was initially suspected that Bi3+ reduction to Bi0 may generate oxygen that caused the foaming. Observations from scanning electron microscopy of some HLW glass samples indicated a Bi-phosphate association. Bi LIII XAS of 13 Bi-containing waste glass formulations of various compositions were measured that exhibited varying degrees of melt foaming. The Bi XAS are similar for all glasses investigated, and indicate Bi3+O3 nearest-neighbor environments with Bi–O distances near 2.13 Å. This environment is similar to the most localized Bi coordination characteristics in the crystalline Bi-silicates, eulytite (Bi4Si3O12) and bismutoferrite (BiFe2Si2O8OH). However, the Bi-environments in the glasses are distinctly different from the Bi-site in crystalline BiPO4; therefore, XAS indicates no evidence of Bi-phosphate domains in the glasses measured. No XAS evidence was observed in any of the glasses investigated for Bi clustering, such as metallic Bi, or Bi–O–Bi bonding. Since the local Bi environments look similar for all glasses investigated, Bi XAS data and analyses show no association

  9. High level waste characterization in support of low level waste certification. I. HLW supernate radionuclide characterization

    Energy Technology Data Exchange (ETDEWEB)

    Jamison, M.E.; d`Entremont, P.D.; Clemmons, J.S.; Bess, C.E.; Brown, D.F.

    1994-07-08

    High Level Waste Programs has radioactive waste storage, treatment and processing facilities that are located in the F and H Areas at the Savannah River Site. These facilities include the Effluent Treatment Facility (ETF), F and H Area Tank Farms, Extended Sludge Processing (ESP), and In-Tank Precipitation (ITP). Job wastes are generated from operation, maintenance, and construction activities inside radiological areas. These items may have been contaminated with radioactive supernate, salt, and sludge material. Most of these wastes will be disposed of in the E-area Vaults. Therefore, an isotopic and hazardous characterization must be performed. The characterization of HLW supernate radionuclides is discussed in Chapter I. The characterization for salt and sludge phases, which can also contaminate LLW, will be included in other Chapters.

  10. Savannah River Site (SRS) high level waste (HLW) structural integrity program

    International Nuclear Information System (INIS)

    The Savannah River Site has fifty-one underground tanks for radioactive waste storage and processing with doubly-contained piping systems for waste transfer. The SRS High Level Waste structural Integrity Program provides a process for evaluation and documenting material aging issues for structures, systems and components (SSC) in these facilities to maintain their confinement function. SRS has been monitoring waste, waste storage tanks, testing transfer lines and controlling waste chemistry for many years. A successful structural integrity (SI) program requires the following: detailed understanding of applicable degradation mechanisms; controlled chemistries and additions, as necessary; regular chemistry sampling and monitoring; structural capacity considerations; and a combination of on-line and periodic inspection and testing programs to provide early detection of generic degradation and verify effectiveness of the management of degradation under aging conditions identified by the SI Program. The application of these elements in the HLW SI Program achieves confinement in the facilities throughout desired service life

  11. Assessment of dose conversion factors in a generic biosphere of a Korea HLW repository

    International Nuclear Information System (INIS)

    Radioactive species released from a waste repository migrate through engineered and natural barriers and eventually reach the biosphere. Once entered the biosphere, contaminants transport various exposure pathways and finally reach a human. In this study the full RES matrix explaining the key compartments in the biosphere and their interactions is introduced considering the characteristics of the Korean biosphere. Then the three exposure groups are identified based on the compartments of interest. The full exposure pathways and corresponding mathematical expression for mass transfer coefficients and etc are developed and applied to assess the dose conversion factors of nuclides for a specific exposure group. Dose conversion factors assessed in this study will be used for total system performance assessment of a potential Korean HLW repository

  12. KAERI Underground Research Facility (KURF) for the Demonstration of HLW Disposal Technology

    International Nuclear Information System (INIS)

    In order to dispose of high-level radioactive waste(HLW) safely in geological formations, it is necessary to assess the feasibility, safety, appropriateness, and stability of the disposal concept at an underground research site, which is constructed in the same geological formation as the host rock. In this paper, the current status of the conceptual design and the construction of a small scale URL, which is named as KURF, were described. To confirm the validity of the conceptual design of the underground facility, a geological survey including a seismic refraction survey, an electronic resistivity survey, a borehole drilling, and in situ and laboratory tests had been carried out. Based on the site characterization results, it was possible to effectively design the KURF. The construction of the KURF was started in May 2005 and the access tunnel was successfully completed in March 2006. Now the construction of the research modules is under way

  13. High level waste characterization in support of low level waste certification. I. HLW supernate radionuclide characterization

    International Nuclear Information System (INIS)

    High Level Waste Programs has radioactive waste storage, treatment and processing facilities that are located in the F and H Areas at the Savannah River Site. These facilities include the Effluent Treatment Facility (ETF), F and H Area Tank Farms, Extended Sludge Processing (ESP), and In-Tank Precipitation (ITP). Job wastes are generated from operation, maintenance, and construction activities inside radiological areas. These items may have been contaminated with radioactive supernate, salt, and sludge material. Most of these wastes will be disposed of in the E-area Vaults. Therefore, an isotopic and hazardous characterization must be performed. The characterization of HLW supernate radionuclides is discussed in Chapter I. The characterization for salt and sludge phases, which can also contaminate LLW, will be included in other Chapters

  14. A health risk based approach for HLW repository environmental protection criteria

    International Nuclear Information System (INIS)

    This paper illustrates a health risk based criteria for the potential HLW repository at Yucca Mountain. The criteria is expressed as an overall health risk goal. Compliance with the standard is proposed to be demonstrated by meeting specific design requirements and design objectives. It is proposed that different degrees of proof apply to design objectives and design requirements. The approach also proposes a combination of more conventional code-like requirements and probabilistic requirements for protection of public health and safety and the environment. The strawman criteria are similar in many respects to ICRP 46, and European repository criteria. However, we propose that the probability distribution of dose to an average individual in a critical population group be used as the measure of health risk. Implementation is illustrated through a probabilistic analysis of a future site specific biosphere

  15. Conditions for which SSI effects in HLW waste tanks are small

    International Nuclear Information System (INIS)

    The objective of this paper is to evaluate the significance of SSI effects in controlling the seismic response of high level waste (HLW) tanks. The tanks considered in the study consist of steel base-supported tanks contained within a concrete vault structure. The parameters describing the physical characteristics of the tanks are selected to span the range of tank properties found at the DOE sites. The tank/fluid system is modeled with a cantilever beam having the frequency of the tank and a mass equal to the impulsive mass of the fluid. The vault is modeled as a rigid body and attached to the free field with lumped parameter SSI models. Transfer functions are developed both including and neglecting SSI effects, and these are compared with each other to assess the significance of SSI

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

    International Nuclear Information System (INIS)

    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 1000C. 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/cm2/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)

  17. The Swiss concept for the disposal of spent fuel and vitrified HLW

    International Nuclear Information System (INIS)

    Management of spent fuel (SF), vitrified high-level waste (HLW) and long-lived intermediate-level waste (ILW) is based on the concept of deep geological disposal, namely long-term, effective isolation of the waste in suitable deep rock formations. The first project studies carried out by Nagra in this respect already lie more than 20 years in the past, when disposal in the crystalline basement and in clay was considered. The strategy developed by Nagra over the years agrees well with the concept of 'monitored long-term geological disposal' as formulated by (EKRA 2000) and contained in the new Nuclear Energy Act of 2003. This paper provides an overview of the concept for facilities and operation of a deep geological repository for SF/HLW/ILW, as prepared for the 'Entsorgungsnachweis' project, together with a geological synthesis report for the Zuercher Weinland and a report on long-term safety. The facilities and operation concept look at the feasibility of constructing a repository in the Opalinus Clay of the Zuercher Weinland. It also provides project-specific input for analysing and demonstrating the long-term safety of such a repository. The individual structural elements and facility components for which the feasibility study was conducted are brought together as a modular system to form a stand-alone reference project. They can be adapted later to meet local features and requirements. The main focus of the paper shall be on selected system elements concerning design, layout and operational aspects including operational safety. (author)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sreten Mastilovic

    2001-07-31

    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.

  20. Oxyfluoroborate host glass for upconversion application: phonon energy calculation

    Science.gov (United States)

    Abdel-Baki, Manal; El-Diasty, Fouad

    2016-04-01

    Reducing the glass phonon energy is an essential procedure to achieve high efficient radiative upconversion process. The degree of covalence of chemical bonds is responsible for the high oscillator strength of intracenter transitions in rare-earth ions. So, conversion covalent to ionic glass character is proposed as a structure-sensitive criterion that controls the phonon energy of the glasses. A series of oxyfluoro aluminum-borate host glasses used for upconversion application is prepared by the conventional melt-quenching technique. Through lithium oxide substitution by lithium fluoride, the ionic-covalent property of Li+ ion successes to regulate the band gap energies of the studied glasses. Furthermore, a new method to determine the glass phonon energy is offered.

  1. Waveguide fabrication in phosphate glasses using femtosecond laser pulses

    International Nuclear Information System (INIS)

    We report on the response of glass to focused femtosecond (fs) laser pulses during waveguide fabrication in a commercial sodium aluminum phosphate glass (Schott IOG-1). Single-pass longitudinal translation of IOG-1 glass with respect to the focused laser beam at a rate of 20 μm/s and pulse energies of 3.5 μJ results in the formation of two waveguides located on opposite sides of the laser-exposed region, which itself does not guide light. This behavior is different from that of the more widely studied silica glass system. The precise location of the waveguides in IOG-1 glass depends on the relative tilt of the fs laser beam with respect to the sample translation direction. Fluorescence imaging of the modified glass using a confocal microscope setup reveals the formation of color center defects in the exposed region but not within the waveguides

  2. Measuring Humidity in Sealed Glass Encasements

    Science.gov (United States)

    West, James W.; Burkett, Cecil G.; Levine, Joel S.

    2005-01-01

    A technique has been devised for measuring the relative humidity levels in the protective helium/water vapor atmosphere in which the Declaration of Independence, the United States Constitution, and the Bill of Rights are encased behind glass panels on display at the National Archives in Washington, DC. The technique is noninvasive: it does not involve penetrating the encasements (thereby risking contamination or damage to the priceless documents) to acquire samples of the atmosphere. The technique could also be applied to similar glass encasements used to protect and display important documents and other precious objects in museums. The basic principle of the technique is straightforward: An encasement is maintained at its normal display or operating temperature (e.g., room temperature) while a portion of its glass front panel is chilled (see Figure 1) until condensed water droplets become visible on the inside of the panel. The relative humidity of the enclosed atmosphere can then be determined as a known function of the dew point, the temperature below which the droplets condense. Notwithstanding the straightforwardness of the basic principle, careful attention to detail is necessary to enable accurate determination of the dew point. In the initial application, the affected portion of the glass panel was cooled by contact with an aluminum plate that was cooled by a thermoelectric module, the exhaust heat of which was dissipated by a heat sink cooled by a fan. A thermocouple was used to measure the interior temperature of the aluminum plate, and six other thermocouples were used to measure the temperatures at six locations on the cooled outer surface of the glass panel (see Figure 2). Thermal grease was applied to the aluminum plate and the thermocouples to ensure close thermal contact. Power was supplied to the thermoelectric module in small increments, based on previous laboratory tests. A small flashlight and a magnifying glass were used to look for water

  3. Regeneration of aluminum hydride

    Science.gov (United States)

    Graetz, Jason Allan; Reilly, James J; Wegrzyn, James E

    2012-09-18

    The present invention provides methods and materials for the formation of hydrogen storage alanes, AlH.sub.x, where x is greater than 0 and less than or equal to 6 at reduced H.sub.2 pressures and temperatures. The methods rely upon reduction of the change in free energy of the reaction between aluminum and molecular H.sub.2. The change in free energy is reduced by lowering the entropy change during the reaction by providing aluminum in a state of high entropy, and by increasing the magnitude of the change in enthalpy of the reaction or combinations thereof.

  4. Aluminum Hydroxide and Magnesium Hydroxide

    Science.gov (United States)

    Aluminum Hydroxide, Magnesium Hydroxide are antacids used together to relieve heartburn, acid indigestion, and upset stomach. They ... They combine with stomach acid and neutralize it. Aluminum Hydroxide, Magnesium Hydroxide are available without a prescription. ...

  5. Thermal Stress Behavior of Aluminum Nanofilms under Heat Cycling

    International Nuclear Information System (INIS)

    In-situ thermal stress in aluminum nanofilms with silicon oxide glass (SOG) passivation was investigated by using synchrotron radiation at the SPring-8. Aluminum films of varying thickness (10, 20, 50 nm) were deposited on thermally oxidized silicon wafers by RF magnetron sputtering. Each specimen was heated in air over two cycles between room temperature and 300 deg. C. The following results were obtained: (1) {111} planes of aluminum nanofilm crystals were oriented parallel to the substrate normal; (2) the intensity of 111 diffraction was almost independent of temperature except in the case of the 50-nm-thick film; (3) the FWHM of 111 diffraction was almost independent of temperature at any given film thickness; and (4) for all films, the thermal stress varied linearly with heating temperature, and the hysteresis between the heating and cooling steps disappeared

  6. Behaviour of sintered glasses containing simulated high level wastes under repository conditions

    International Nuclear Information System (INIS)

    Within the framework of an IAEA-CNEA Research Contract established in 1992, we have studied the vitrification of simulated high level wastes (HLW) from reprocessed PHWR fuel on a laboratory scale by the alternative method of powder glass sintering. The HLW was added in a proportion of 10 wt% to alumino-borosilicate glass powder. The powder mixture was treated using the procedure of cold pressing and subsequent sintering at about 1000K. The microstructures of the sintered samples were analyzed by optical microscopy and SEM. Water corrosion tests were done in different aqueous media. Devitrification studies were conducted by heat-treating the samples at 950, 1000, 1050 and 1100K for 3, 30, 300 and 3000 h. Values of the corrosion resistance of the studied glasses, were similar to those found in the literature for waste forms produced by melting technology. We have obtained promising results indicating that the glass sintering method could be an interesting alternative for immobilization for not only high level wastes but intermediate level wastes. (author)

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

    International Nuclear Information System (INIS)

    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

  8. 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.

  9. RECLAMATION OF ALUMINUM FINISHING SLUDGES

    Science.gov (United States)

    The research study of the reclamation of aluminum-anodizing sludges was conducted in two sequential phases focused on enhanced dewatering of aluminum-anodizing sludges to produce commercial-strength solutions of aluminum sulfate, i.e., liquid alum. The use of high-pressure (14 to...

  10. Electrically conductive anodized aluminum coatings

    Science.gov (United States)

    Alwitt, Robert S. (Inventor); Liu, Yanming (Inventor)

    2001-01-01

    A process for producing anodized aluminum with enhanced electrical conductivity, comprising anodic oxidation of aluminum alloy substrate, electrolytic deposition of a small amount of metal into the pores of the anodized aluminum, and electrolytic anodic deposition of an electrically conductive oxide, including manganese dioxide, into the pores containing the metal deposit; and the product produced by the process.

  11. Fabrication and optical properties of lead-germanate glasses and a new class of optical fibres doped with Tm3+

    OpenAIRE

    Wang, J.; Lincoln, J.R.; Brocklesby, W.S.; Deol, R.S.; MacKechnie, C.J.; Pearson, A.; Tropper, A.C.; Hanna, D.C.; Payne, D. N.

    1992-01-01

    In this article we present a study of a new class of optical fibers based on lead germanate glass. The maximum vibrational frequency of this glass is intermediate between silica and zirconium barium lanthanum aluminum fluoride glass, causing a beneficial change in nonradiative decay and therefore quantum efficiency for particular laser transitions. Fabrication of high-strength, low-loss fibers of this glass has been achieved by modification of the composition to produce optimal physical prope...

  12. 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

  13. Invisible Display in Aluminum

    DEFF Research Database (Denmark)

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

    2005-01-01

    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...

  14. Aluminum Sulfate 18 Hydrate

    Science.gov (United States)

    Young, Jay A.

    2004-01-01

    A chemical laboratory information profile (CLIP) of the chemical, aluminum sulfate 18 hydrate, is presented. The profile lists physical and harmful properties, exposure limits, reactivity risks, and symptoms of major exposure for the benefit of teachers and students using the chemical in the laboratory.

  15. Hot pressing aluminum nitride

    International Nuclear Information System (INIS)

    Experiment was performed on the hot pressing of aluminum nitride, using three kinds of powder which are: a) made by electric arc method, b) made by nitrifying aluminum metal powder, and c) made from alumina and carbon in nitrogen atmosphere. The content of oxygen of these powders was analyzed by activation analysis using high energy neutron irradiation. The density of hot pressed samples was classified into two groups. The high density group contained oxygen more than 3 wt. %, and the low density group contained about 0.5 wt %. Typical density vs. temperature curves have a bending point near 1,5500C, and the sample contains iron impurity of 0.5 wt. %. Needle crystals were found to grow near 1,5500C by VLS mechanism, and molten iron acts a main part of mechanism as a liquid phase. According to the above-mentioned curve, the iron impurity in aluminum nitride prevents densification. The iron impurity accelerates crystal growth. Advance of densification may be expected by adding iron impurity, but in real case, the densification is delayed. Densification and crystal growth are greatly accelerated by oxygen impurity. In conclusion, more efforts must be made for the purification of aluminum nitride. In the present stage, the most pure nitride powder contains about 0.1 wt. % of oxygen, as compared with good silicon carbide crystals containing only 10-5 wt. % of nitrogen. (Iwakiri, K.)

  16. Formulation of Japanese consensus-building model for HLW geological disposal site determination. 3. Development of digital contents on social consensus-buildings at educational institution

    International Nuclear Information System (INIS)

    To establish the sustainable community in Japan, formation of Japanese social consensus-building model for HLW geologic disposal site determination is one of key issues. In our project team of faculty of education, we found interesting middle school digital contents for HLW disposal site determination in London. We have been translating all of the digital contents for HLW disposal site determination into Japanese and we have been discussing on the model for next-generation Japanese consensus-building at middle school level. (author)

  17. Aluminum doping of CdTe polycrystalline films starting from the heterostructure CdTe/Al

    OpenAIRE

    Becerril, M.; O. Vigil-Galán; G. Contreras-Puente; O. Zelaya-Angel

    2011-01-01

    Aluminum doped CdTe polycrystalline films were obtained from the heterostructure CdTe/Al/Corning glass. The aluminum was deposited by thermal vacuum evaporation and the CdTe by sputtering of a CdTe target. The aluminum was introduced into the lattice of the CdTe from a thermal annealed to the CdTe/Al/Corning glas heterostructure. The electrical, structural, nd optical properties were analyzed as a function of the Al concentrations. It found that when Al is incorporated, the electrical resisti...

  18. Structure and properties of sodium aluminosilicate glasses from molecular dynamics simulations

    DEFF Research Database (Denmark)

    Xiang, Ye; Du, Jincheng; Smedskjær, Morten Mattrup;

    2013-01-01

    . In peraluminous compositions (Al/Na > 1), small amounts of five-fold coordinated aluminum ions are present while the concentration of six-fold coordinated aluminum is negligible. Oxygen triclusters are also found to be present in peraluminous compositions, and their concentration increases with...... aluminum as a function of chemical composition in these glasses. The short- and medium-range structures such as aluminum coordination, bond angle distribution around cations, Qn distribution (n bridging oxygen per network forming tetrahedron), and ring size distribution have been systematically studied. In...... addition, the mechanical properties including bulk, shear, and Young's moduli have been calculated and compared with experimental data. It is found that aluminum ions are mainly four-fold coordinated in peralkaline compositions (Al/Na < 1) and form an integral part of the rigid silicon-oxygen glass network...

  19. Modification of deposit formation in glass furnance heat recovery

    Energy Technology Data Exchange (ETDEWEB)

    Hempel, H.-U.; Novothy, R.; Staller, S.; Kraemer, J.

    1987-07-07

    This patent describes methods for removing deposits formed in the heat exchange tube of waste heat recovery boilers for glass furnaces in which sodium aluminum silicate fine particles are introduced into the waste gases before they enter the heat exchange tubes.

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

    International Nuclear Information System (INIS)

    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.

  1. Cavern disposal concepts for HLW/SF: assuring operational practicality and safety with maximum programme flexibility

    International Nuclear Information System (INIS)

    Most conventional engineered barrier system (EBS) designs for HLW/SF repositories are based on concepts developed in the 1970s and 1980s that assured feasibility with high margins of safety, in order to convince national decision makers to proceed with geological disposal despite technological uncertainties. In the interval since the advent of such 'feasibility designs', significant progress has been made in reducing technological uncertainties, which has lead to a growing awareness of other, equally important uncertainties in operational implementation and challenges regarding social acceptance in many new, emerging national repository programs. As indicated by the NUMO repository concept catalogue study (NUMO, 2004), there are advantages in reassessing how previous designs can be modified and optimised in the light of improved system understanding, allowing a robust EBS to be flexibly implemented to meet nation-specific and site-specific conditions. Full-scale emplacement demonstrations, particularly those carried out underground, have highlighted many of the practical issues to be addressed; e.g., handling of compacted bentonite in humid conditions, use of concrete for support infrastructure, remote handling of heavy radioactive packages in confined conditions, quality inspection, monitoring / ease of retrieval of emplaced packages and institutional control. The CAvern REtrievable (CARE) concept reduces or avoids such issues by emplacement of HLW or SF within multi-purpose transportation / storage / disposal casks in large ventilated caverns at a depth of several hundred metres. The facility allows the caverns to serve as inspectable stores for an extended period of time (up to a few hundred years) until a decision is made to close them. At this point the caverns are backfilled and sealed as a final repository, effectively with the same safety case components as conventional 'feasibility designs'. In terms of operational practicality an d safety, the CARE

  2. Fluoride and aluminum release from restorative materials using ion chromatography

    Directory of Open Access Journals (Sweden)

    Zeynep Okte

    2012-02-01

    Full Text Available OBJECTIVE: The aim of this study was to determine the amounts of fluoride and aluminum released from different restorative materials stored in artificial saliva and double-distilled water. Material and METHODS: Cylindrical specimens (10 x 1 mm were prepared from 4 different restorative materials (Kavitan Plus, Vitremer, Dyract Extra, and Surefil. For each material, 20 specimens were prepared, 10 of which were stored in 5 mL artificial saliva and 10 of which were stored in 5 mL of double-distilled water. Concentrations of fluoride and aluminum in the solutions were measured using ion chromatography. Measurements were taken daily for one week and then weekly for two additional weeks. Data were analyzed using two-way ANOVA and Duncan's multiple range tests (p<0.05. RESULTS: The highest amounts of both fluoride and aluminum were released by the resin-modified glass ionomer cement Vitremer in double-distilled water (p<0.05. All materials released significantly more fluoride in double-distilled water than in artificial saliva (p<0.05. In artificial saliva, none of the materials were observed to release aluminum. CONCLUSION: It was concluded that storage media and method of analysis should be taken into account when the fluoride and aluminum release from dental materials is assessed.

  3. Heat Transfer in Glass, Aluminum, and Plastic Beverage Bottles

    Science.gov (United States)

    Clark, William M.; Shevlin, Ryan C.; Soffen, Tanya S.

    2010-01-01

    This paper addresses a controversy regarding the effect of bottle material on the thermal performance of beverage bottles. Experiments and calculations that verify or refute advertising claims and represent an interesting way to teach heat transfer fundamentals are described. Heat transfer coefficients and the resistance to heat transfer offered…

  4. Discussing compliance. Summary report from discussions with Robert Bernero and Chris Whipple regarding compliance with the Swedish HLW Regulations from meetings in Stockholm May 3 and 4, 1999

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Mikael

    1999-06-01

    Summary report from discussions with Robert Bernero and Chris Whipple regarding compliance with the Swedish HLW Regulations from meetings in Stockholm. The report also contains bibliographical information and preliminary observations made by Robert Bernero and Chris Whipple.

  5. Discussing compliance. Summary report from discussions with Robert Bernero and Chris Whipple regarding compliance with the Swedish HLW Regulations from meetings in Stockholm May 3 and 4, 1999

    International Nuclear Information System (INIS)

    Summary report from discussions with Robert Bernero and Chris Whipple regarding compliance with the Swedish HLW Regulations from meetings in Stockholm. The report also contains bibliographical information and preliminary observations made by Robert Bernero and Chris Whipple

  6. A review of regional and in-situ underground geoscientific investigations related to the geological disposal of HLW in Japan

    International Nuclear Information System (INIS)

    In accordance with a generic approach being applied for Japan's research and development program for disposal of high-level radioactive waste, geoscientific investigations ranging from regional groundwater flow modeling to microscopic observations of nuclide migration are underway. Results from these investigations will serve as a scientific basis to ensure the role of geological environment in the context of geological disposal of HLW. This paper provides an overview of some of these investigations conducted by PNC

  7. Aluminum microstructures on anodic alumina for aluminum wiring boards.

    Science.gov (United States)

    Jha, Himendra; Kikuchi, Tatsuya; Sakairi, Masatoshi; Takahashi, Hideaki

    2010-03-01

    The paper demonstrates simple methods for the fabrication of aluminum microstructures on the anodic oxide film of aluminum. The aluminum sheets were first engraved (patterned) either by laser beam or by embossing to form deep grooves on the surface. One side of the sheet was then anodized, blocking the other side by using polymer mask to form the anodic alumina. Because of the lower thickness at the bottom part of the grooves, the part was completely anodized before the complete oxidation of the other parts. Such selectively complete anodizing resulted in the patterns of metallic aluminum on anodic alumina. Using the technique, we fabricated microstructures such as line patterns and a simple wiring circuit-board-like structure on the anodic alumina. The aluminum microstructures fabricated by the techniques were embedded in anodic alumina/aluminum sheet, and this technique is promising for applications in electronic packaging and devices. PMID:20356280

  8. Compatibility of candidate overpack materials with deep argillaceous HLW disposal environments

    International Nuclear Information System (INIS)

    The Belgian R and D programme on the disposal of high level radioactive waste has been focused on the qualification of deep argillaceous formations for HLW disposal, because they have a number of inherent attractive characteristics and also because they are sufficiently abundant to cover the Belgian needs. A large number of corrosion resistant materials as well as some corrosion allowance materials have been tested. The laboratory test program included accelerated tests as well as exposure tests in simulated repository conditions. Initial ''field'' experiments have been performed in a near surface clay quarry. However, in order to obtain realistic corrosion rate estimates corrosion experiments with simultaneous monitoring of the clay environment parameters have been started in a 230 meters deep underground laboratory constructed in the Boom clay formation at Mol. Two types of experimental devices have been designed. In the first type coupons are mounted on an internally heated tubular holder and are directly exposed to the solid clay. In a second type of test a purge gas is used to extract corrosive products from the clay and circulate them subsequently over a number of metal coupons. Relevant parameters such as pH and Eh are continuously monitored for evaluating the evolution of soil agressivity after the initial chemical and mechanical disturbance of the clay, caused by the operations required for the introduction of the experimental devices. (author)

  9. Technology assessment of disposal alternatives to determine a reference geological repository system for HLW

    International Nuclear Information System (INIS)

    This study is to determine the most promising alternative, that will be developed further as a reference HLW repository system, by comparing the 7 alternatives that were proposed based on the spent fuel packaging options concerning the characteristics of spent PWR and CANDU fuel generated from the domestic NPP and the waste package arrangements and repository layout options. It should be determined by comparing the proposed alternatives from the aspects of technology, safety and economics. In this study, however, the comparison of alternatives was just based on the technology assessment because of the lack of the relevant information. The comparison criteria includes the degree of difficulty, development and maturity of the technology to be applied in repository system construction, operation, retrieval, etc. and the safety during the repository construction and operation. Based on such comparison criteria, the alternative comparison study was performed by a typical pair-wise comparison method. The result showed that, from the aspect of the construction, vertical emplacement options ranked high so that HSA and HCop ranked first and second, respectively. On the other hand, from the aspect of operation, the vertical emplacement options ranked high and VSA and VAT were ranked first and second. Depending upon the degree of importance of construction and operation of the repository, the final results of the alternatives comparison could be changed. (author). 19 refs., 6 tabs., 13 figs

  10. Long term safety assessment of disposal concepts for HLW and spent fuel in rock salt

    International Nuclear Information System (INIS)

    Disposal of high-level radioactive waste in Germany is planned to take place in deep salt repositories. Direct disposal of spent fuel is taken into account as well as disposal of reprocessed high-level waste. Borehole and drift emplacement techniques are considered for both HLW and spent fuel. The following disposal concepts are taken into account: pure borehole emplacement with waste in steel drums or canisters, pure drift emplacement with waste in Pollux casks, and combined drift and borehole emplacement. Each of these concepts was investigated for different ratios of waste from direct disposal and from reprocessing. Long term safety assessments were performed with scenarios of brine intrusion via possible anhydrite veins and from brine pockets in the surrounding salt rock. A survey of the disposal concepts as well as the models used by the computer code for long term safety assessment will be presented. The safety assessments investigated the influence of the geometrical layout of the emplacement sites, the layout temperature of the entire repository and the different behavior of spent fuel and reprocessed waste. The calculations were performed both deterministic and probabilistic. The deterministic calculations were performed with best estimate values of the input parameters and in some variants with locally varied values of selected parameters

  11. Post Closure Long Term Safely of the Initial Container Failure Scenario for a Potential HLW Repository

    International Nuclear Information System (INIS)

    A waste container, one of the key components of a multi-barrier system in a potential high level radioactive waste (HLW) repository in Korea ensures the mechanical stability against the lithostatic pressure of a deep geologic medium and the swelling pressure of the bentonite buffer. Also, it delays potential release of radionuclides for a certain period of time, before it is corroded by intruding impurities. Even though the material of a waste container is carefully chosen and its manufacturing processes are under quality assurance processes, there is a possibility of initial defects in a waste container during manufacturing. Also, during the deposition of a waste container in a repository, there is a chance of an incident affecting the integrity of a waste container. In this study, the appropriate Features, Events, and Processes (FEP's) to describe these incidents and the associated scenario on radionuclide release from a container to the biosphere are developed. Then the total system performance assessment on the Initial waste Container Failure (ICF) scenario was carried out by the MASCOT-K, one of the probabilistic safety assessment tools KAERI has developed. Results show that for the data set used in this paper, the annual individual dose for the ICF scenario meets the Korean regulation on the post closure radiological safety of a repository.

  12. Stress analysis of HLW containers. Preliminary ring test exercise Compas project

    International Nuclear Information System (INIS)

    This document describes the series of experiments and associated calculations performed as the Compas preliminary ring test exercise. A number of mild steel rings, representative of sections through HLW containers, some notched and pre-cracked, were tested in compression right up to and beyond their ultimate load. The Compas project partners independently modelled the behaviour of these rings using their finite element codes. Four different ring types were tested, and each test was repeated three times. For three of the ring types, the three test repetitions gave identical results. The fourth ring, which was not modelled by the partners, had a 4 mm thick layer of weld metal deposited on its surface. The three tests on this ring did not give identical results and suggested that the effect of welding methods should be addressed at a later stage of the project. Fracture was not found to be a significant cause of ring failure. The results of the ring tests were compared with the partners predictions, and additionally some time was spent assessing where the use of the codes could be improved. This exercise showed that the partners codes have the ability to produce results within acceptable limits. Most codes were unable to model stable crack growth. There were indications that some codes would not be able to cope with a significantly more complex three-dimensional analysis

  13. Study on stability of HLW repository in the middle region of Shule river fault zone

    International Nuclear Information System (INIS)

    The authors studies the geological features and stability of the middle region of Shule river fault zone by means of the interpretation of remote sensing images, geological researches and geophysical prospecting. The knowledge of Shule river fault zone is much improved. The structural plane features of the trunk fault of the middle region of Shule river fault zone are observed at hand-outcrop, and its mechanics character is preliminary determined as compression. The result of VLF electromagnetic measurement on this area is consistent with the result of the interpretation of remote sensing images, and geological researches. The knowledge of the space distribution law of the middle region of Shule river zone.is confirmed Some deep geological information is obtained from analysing and studying deep crust structure and earth temperature background by use of the regional geophysical and earth temperature date. The comment on stability of the middle region of the Shule river fault zone indicates that its middle region and the north mountain which is located at the north of Shule river fault zone have better stability. It is a perfect area for being a site of HLW repository

  14. Seismic design evaluation guidelines for buried piping for the DOE HLW Facilities

    International Nuclear Information System (INIS)

    This paper presents the seismic design and evaluation guidelines for underground piping for the Department of Energy (DOE) High-Level-Waste (HLW) Facilities. The underground piping includes both single and double containment steel pipes and concrete pipes with steel lining, with particular emphasis on the double containment piping. The design and evaluation guidelines presented in this paper follow the generally accepted beam-on-elastic-foundation analysis principle and the inertial response calculation method, respectively, for piping directly in contact with the soil or contained in a jacket. A standard analysis procedure is described along with the discussion of factors deemed to be significant for the design of the underground piping. The following key considerations are addressed: the design feature and safety requirements for the inner (core) pipe and the outer pipe; the effect of soil strain and wave passage; assimilation of the necessary seismic and soil data; inertial response calculation for the inner pipe; determination of support anchor movement loads; combination of design loads; and code comparison. Specifications and justifications of the key parameters used, stress components to be calculated and the allowable stress and strain limits for code evaluation are presented

  15. Study on evaluation method of potential impact of natural phenomena on a HLW disposal system

    International Nuclear Information System (INIS)

    Japan Atomic Energy Agency (JAEA) have developed a formal evaluation method to assess the potential impact of natural phenomena (earthquakes and faulting; volcanism; uplift, subsidence, denudation and sedimentation; climatic and sea-level changes) on a high level radioactive waste (HLW) disposal system. This study focuses on identifying key T-H-M-C-G (thermal - hydrological - mechanical - chemical - geometrical) process associated with perturbations at description of potential changes of the geological environment. In this report, we revised the framework as a part of the total system performance assessment for two purposes: the first one is quantification of relationship of characteristic of natural phenomena between geological environmental conditions (THMCG), and the other one is quantification of relationship of THMCG condition between parameters of performance assessment. On the other hand, we applied the revised framework to all natural phenomena. As a result, to apply the revised framework, we could show that information integration could carry out efficiently. In particular, we introduced the view of change of situations about phenomena such as climate changes which the grade of impact cannot grasp easily as compared with volcanism. Thereby, information integration was attained by the common framework. Furthermore, we could show that suitable scenarios might be chosen by information integration. (author)

  16. Environmental risk assessment: its contribution to criteria development for HLW disposal

    International Nuclear Information System (INIS)

    Principles for radioactive waste management have been provided by the International Atomic Energy Agency in Safety Series No.111-F, which was published in 1995. This has been a major step forward in the process of achieving acceptance for proposals for disposal of radioactive waste, for example, for High Level Waste disposal in deep repositories. However, these principles have still to be interpreted and developed into practical radiation protection criteria. Without prejudicing final judgements on the acceptability of waste proposals, an important aspect is that practical demonstration of compliance (or the opposite) with these criteria must be possible. One of the IAEA principles requires that radioactive waste shall be managed in such a way as to provide an acceptable level of protection of the environment. There has been and continues to be considerable debate as to how to demonstrate compliance with such a principle. This paper briefly reviews the current status and considers how experience in other areas of environmental protection could contribute to criteria development for HLW disposal

  17. Strategy for safety case development: impact of a volunteering approach to siting a japanese HLW repository

    International Nuclear Information System (INIS)

    NUMO strategy for safety case development is constrained by a staged siting approach, which has been initiated by a call for volunteer municipalities to host the HLW repository. For each site, the safety case is an important factor to be considered at the selection steps which narrow down towards the preferred repository location. This is particularly challenging, however, as every site requires a tailored repository concept, with associated performance assessment and an individual site evaluation programme all of which evolve with gradually increasing understanding of the host environment. In order to maintain flexibility without losing focus, NUMO has developed a formalized tailoring procedure, termed the NUMO Structured Approach (NSA). The NSA guides the interaction of the key site characterisation, repository design and performance assessment groups and is facilitated by tools to help the decision making associated with the tailoring process (e.g. a requirements management system) and with comparison of siting and design options (e.g. multi-attribute analysis). Pragmatically, the post-closure safety case will initially emphasize near-field processes and a robust engineering barrier system, considering the limited geological information at early stages. This will be complemented by a more realistic assessment of total system performance, as needed to compare options. In addition, efforts to rigorously assess operational phase safety and the practicality of assuring quality of the constructed engineered barriers are components of the total safety case which are receiving particular attention now, as they may better discriminate between sites while information is still limited. (authors)

  18. Development of a computer tool to support scenario analysis for safety assessment of HLW geological disposal

    International Nuclear Information System (INIS)

    In 'H12 Project to Establishing Technical Basis for HLW Disposal in Japan' a systematic approach that was based on an international consensus was adopted to develop scenarios to be considered in performance assessment. Adequacy of the approach was, in general term, appreciated through the domestic and international peer review. However it was also suggested that there were issues related to improving transparency and traceability of the procedure. To achieve this, improvement of scenario analysis method has been studied. In this study, based on an improvement method for treatment of FEP interaction a computer tool to support scenario analysis by specialists of performance assessment has been developed. Anticipated effects of this tool are to improve efficiency of complex and time consuming scenario analysis work and to reduce possibility of human errors in this work. This tool also enables to describe interactions among a vast number of FEPs and the related information as interaction matrix, and analysis those interactions from a variety of perspectives. (author)

  19. Storage of HLW in engineered structures: air-cooled and water-cooled concepts

    International Nuclear Information System (INIS)

    A comparative study on an air-cooled and a water-cooled intermediate storage of vitrified, highly radioactive waste (HLW) in overground installations has been performed by Nukem and Belgonucleaire respectively. In the air-cooled storage concept the decay heat from the storage area will be removed using natural convection. In the water-cooled storage concept the decay heat is carried off by a primary and secondary forced-cooling system with redundant and diverse devices. The safety study carried out by Nukem used a fault tree method. It shows that the reliability of the designed water-cooled system is very high and comparable to the inherent, safe, air-cooled system. The impact for both concepts on the environment is determined by the release route, but even during accident conditions the release is far below permissible limits. The economic analysis carried out by Belgonucleaire shows that the construction costs for both systems do not differ very much, but the operation and maintenance costs for the water-cooled facility are higher than for the air cooled facility. The result of the safety and economic analysis and the discussions with the members of the working group have shown some possible significant modifications for both systems, which are included in this report. The whole study has been carried out using certain national criteria which, in certain Member States at least, would lead to a higher standard of safety than can be justified on any social, political or economic grounds

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

    International Nuclear Information System (INIS)

    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. 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.

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

    International Nuclear Information System (INIS)

    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

  3. Cutting glass by laser

    Science.gov (United States)

    Kang, Hyoung-Shik; Hong, Soon-Kug; Oh, Seok-Chang; Choi, Jong-Yoon; Song, Min-Gyu

    2002-02-01

    In FPD (Flat Panel Display) devices, the diamond wheel has been used to scribe glass by means of mechanical contact which needs grinding and cleaning processes to remove particles, glass chips, surface cracks and sharp edges. In recent years, laser glass technology that is different from the conventional method of cutting glass by melting, has been researched and utilizes cutting glass by thermal shock. Laser glass cutting by thermal shock can produce cracks in glass by surface cooling after laser heating on glass by means of stress slope on glass surface. When this technology is applied in FPD manufacturing devices, it has several advantages compared to conventional methods as follows: a) non-contact glass cutting: almost no glass chip occurs. b) according to circumstances, grinding and cleaning can be omitted. c) system maintenance can be simplified.

  4. Raman and X-ray absorption spectroscopy studies of chromium–phosphorus interactions in high-bismuth high-level waste glasses

    International Nuclear Information System (INIS)

    High-level waste (HLW) glasses containing bismuth, phosphorus, and chromium were investigated using Raman and X-ray absorption spectroscopy (XAS). The novel and practically important occurrence of foaming on cooling of these melts is associated with P and Cr from the HLW. In response, glasses were synthesized where Bi2O3 and P2O5 contents were varied independently. Relationships between P and Cr were found, where as P2O5-content increases, chromate Cr–O stretch Raman modes diminish intensity, while Cr XAS shows that Cr reduces, from 50% Cr6+ + 50% Cr3+ to nearly 20% Cr6+ + 80% Cr3+, explaining the chromate mode behavior. In the most P2O5-rich glass, the chromate Cr–O distance increases by approximately 0.10 Å, which may indicate bonding between CrO4 and PO4 tetrahedra, similar to that in chromo-phosphates. The presence of chromo-phosphate domains in HLW melts can be linked to oxygen generation as a source of the foaming

  5. 21 CFR 73.1645 - Aluminum powder.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 1 2010-04-01 2010-04-01 false Aluminum powder. 73.1645 Section 73.1645 Food and... ADDITIVES EXEMPT FROM CERTIFICATION Drugs § 73.1645 Aluminum powder. (a) Identity. (1) The color additive aluminum powder shall be composed of finely divided particles of aluminum prepared from virgin aluminum....

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

    International Nuclear Information System (INIS)

    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

  7. Plutonium immobilization plant using glass in new facilities at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    DiSabatino, A.

    1998-06-01

    The Plutonium Immobilization Plant (PIP) accepts plutonium (Pu) from pit conversion and from non-pit sources and, through a glass immobilization process, converts the plutonium into an immobilized form that can be disposed of in a high level waste (HLW) repository. This immobilization process is shown conceptually in Figure 1-1. The objective is to make an immobilized form, suitable for geologic disposal, in which the plutonium is as inherently unattractive and inaccessible as the plutonium in spent fuel from commercial reactors.

  8. SODIUM ENVIRONMENTS IN GLASS

    OpenAIRE

    Greaves, G.

    1981-01-01

    Sodium environments have been measured directly in several oxide glasses using EXAFS at the sodium K edge. The existence of local structure around sodium in glass contradicts the conventional Zachariasen model. Analysis of the EXAFS indicates there are significant differences relating to the glass modifier : glass former chemistry - the details of which demonstrate similarities with crystalline silicates and borates.

  9. Inverted glass harp

    Science.gov (United States)

    Quinn, Daniel B.; Rosenberg, Brian J.

    2015-08-01

    We present an analytical treatment of the acoustics of liquid-filled wine glasses, or "glass harps." The solution is generalized such that under certain assumptions it reduces to previous glass harp models, but also leads to a proposed musical instrument, the "inverted glass harp," in which an empty glass is submerged in a liquid-filled basin. The versatility of the solution demonstrates that all glass harps are governed by a family of solutions to Laplace's equation around a vibrating disk. Tonal analyses of recordings for a sample glass are offered as confirmation of the scaling predictions.

  10. Engineering for Operation of a Future Belgian Deep Geological Repository for ILW and HLW - 12379

    International Nuclear Information System (INIS)

    In Belgium, an advanced conceptual design is being elaborated for deep geologic disposal of high level waste (HLW) and for low and intermediate level waste (LILW) not amenable for surface disposal. The concept is based on a shielded steel and concrete container for disposal of HLW, i.e., the Super-container. LILW will be disposed of in separately designed concrete caissons. The reference host rock is the Boom Clay, a poorly indurated clay formation in northeastern Belgium. Investigations into the potential host rock are conducted at the HADES underground research laboratory in Mol, Belgium. In 2009 the Belgian Agency for Management of Radioactive Waste and Enriched Fissile Materials (ONDRAF/NIRAS) initiated a four year research project aimed at confirming the fundamental feasibility of building and operating a repository. The goal of the program is to demonstrate at a detailed conceptual level that the proposed geologic disposal system can be safely constructed, operated, and progressively closed. Part of the broader research efforts being conducted includes evaluations optimization of the waste transportation shaft, subsurface transportation system, ventilation system, and evaluation of backfilling and sealing concepts for the repository design. The potential for implementation of a waste retrieval strategy encompassing the first 100 years after emplacement is also considered. In the framework of a four year research program aimed at confirming the fundamental feasibility of building and operating a repository in poorly indurated clay design studies have been underway to optimize the waste transportation shaft, subsurface transportation system, and ventilation system. Additionally backfilling and sealing concepts proposed for the potential repository have been reviewed in conjunction with impacts related to the potential future inclusion of a retrievability requirement in governing regulations. The main engineering challenges in the Belgian repository concept are

  11. Methodology of fuel cycles long-term safety assessment of SNF/HLW geological disposal

    International Nuclear Information System (INIS)

    Methodology for the long-term safety assessment of nuclear fuel cycles is given in the presented doctoral thesis. The aim of work was to develop a geological repository model for disposal of spent nuclear fuel (SNF) and high level waste (HLW) using an appropriate software code able to calculate the influence of partitioning and transmutation in advanced fuel cycles. The first step in this process was specifying of indicators which can be used to quantify the radiological impact of each fuel cycle. Indicators such as annual effective dose and radiotoxicity of inventory have been quantitatively analysed to determine the potential risk and radiological consequences associated with production of SNF/HLW. Advanced fuel types bring a number of advantages in comparison to uranium oxide fuel UO2 used worldwide nowadays in terms of safety improvement due to minor actinides transmutation and non-proliferation aspects as well. Within the scope of work, three different fuel cycles are compared from the point of view of long-term safety of deep geological repository. The first considered fuel cycle is the currently used open fuel cycle (UOX) which uses only U-FA (Uranium Fuel Assembly). The second assessed cycle is a closed fuel cycle (MOX) with MOX-FA (Mixed OXides Fuel Assembly) and the third considered one is a partially closed fuel cycle (IMF) with IMC-FA (Inert Matrix Combined Fuel Assembly). Description and input data of advanced fuel cycles have been gained by participation in the EC project RED-IMPACT. Results were calculated using code AMBER, which is a flexible software tool that allows building dynamic compartmental models to represent the migration and fate of contaminants in a system, for example in the surface and sub-surface environment. Contaminants in solid, liquid and gaseous phases can be considered. AMBER gives the user the flexibility to define any number of compartments; any number of contaminants and associated decays; deterministic, probabilistic and

  12. DESIGN OF A CONCRETE SLAB FOR STORAGE OF SNF AND HLW CASKS

    Energy Technology Data Exchange (ETDEWEB)

    J. Bisset

    2005-02-14

    This calculation documents the design of the Spent Nuclear Fuel (SNF) and High-Level Waste (HLW) Cask storage slab for the Aging Area. The design is based on the weights of casks that may be stored on the slab, the weights of vehicles that may be used to move the casks, and the layout shown on the sketch for a 1000 Metric Ton of Heavy Metal (MTHM) storage pad on Attachment 2, Sht.1 of the calculation 170-C0C-C000-00100-000-00A (BSC 2004a). The analytical model used herein is based on the storage area for 8 vertical casks. To simplify the model, the storage area of the horizontal concrete modules and their related shield walls is not included. The heavy weights of the vertical storage casks and the tensile forces due to pullout at the anchorages will produce design moments and shear forces that will envelope those that would occur in the storage area of the horizontal modules. The design loadings will also include snow and live loads. In addition, the design will also reflect pertinent geotechnical data. This calculation will document the preliminary thickness and general reinforcing steel requirements for the slab. This calculation also documents the initial design of the cask anchorage. Other slab details are not developed in this calculation. They will be developed during the final design process. The calculation also does not include the evaluation of the effects of cask drop loads. These will be evaluated in this or another calculation when the exact cask geometry is known.

  13. Key Technologies Analyses to Develop a Deep Borehole disposal Concept for HLW

    International Nuclear Information System (INIS)

    A deep geological disposal system, the disposal depth is about 500 m below ground, is considered as the safest method to isolate the spent fuels(SF) or high-level radioactive waste(HLW) from the human environment with the best available technology at present time. The disposal safety of this system has been demonstrated with underground research laboratory and some advanced countries such as Finland and Sweden are implementing their disposal project on commercial stage. However, if these high-level radioactive wastes can be disposed of in deeper and more stable rock formation than deep geological disposal depth, it has several advantages. 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. And the key technologies and challenges in development of this disposal method with the nuclear environment were analyzed. In this paper, the general concept of deep borehole disposal for spent fuels or high level radioactive wastes which has been developed by some countries according to the rapid advance in the development of borehole drilling technology, as an alternative method to the deep geological disposal method, was reviewed. And the key technologies, such as drilling technology of large diameter borehole, packaging and emplacement technology, sealing technology and performance/safety analyses technologies, and their challenges in development of deep borehole disposal system were analyzed

  14. Recycling glass packaging

    OpenAIRE

    Monica Delia DOMNICA; Leila BARDAªUC

    2015-01-01

    From the specialized literature it follows that glass packaging is not as used as other packages, but in some industries are highly needed. Following, two features of glass packaging will become important until 2017: the shape of the glass packaging and glass recycling prospects in Romania. The recycling of glass is referred to the fact that it saves energy, but also to be in compliance with the provisions indicating the allowable limit values for the quantities of lead and cadmium.

  15. Photoemission study of tris(8-hydroxyquinoline) aluminum/aluminum oxide/tris(8-hydroxyquinoline) aluminum interface

    International Nuclear Information System (INIS)

    The evolution of the interface electronic structure of a sandwich structure involving aluminum oxide and tris(8-hydroxyquinoline) aluminum (Alq), i.e. (Alq/AlOx/Alq), has been investigated with photoemission spectroscopy. Strong chemical reactions have been observed due to aluminum deposition onto the Alq substrate. The subsequent oxygen exposure releases some of the Alq molecules from the interaction with aluminum. Finally, the deposition of the top Alq layer leads to an asymmetry in the electronic energy level alignment with respect to the AlOx interlayer

  16. 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.

  17. Neurofibrillary pathology and aluminum in Alzheimer's disease

    OpenAIRE

    Shin, R. W.; Lee, V. M. Y; Trojanowski, J Q

    1995-01-01

    Since the first reports of aluminum-induced neurofibrillary degeneration in experimental animals, extensive studies have been performed to clarify the role played by aluminum in the pathogenesis of Alzheimer's disease (AD). Additional evidence implicating aluminum in AD includes elevated levels of aluminum in the AD brain, epidemiological data linking aluminum exposure to AD, and interactions between aluminum and protein components in the pathological lesions o...

  18. New Co-containing glass ceramics saturable absorbers for 1.5-μm solid state lasers

    Science.gov (United States)

    Malyarevich, Alexander M.; Denisov, Igor A.; Yumashev, Konstantin V.; Chuvaeva, Tamara I.; Dymshits, Olga S.; Onushchenko, Alexei A.; Zhilin, Alexander A.

    2001-03-01

    New saturable absorber Q-switch for 1.54 %mum Er: glass laser is present. The saturable absorber is transparent glass ceramic containing magnesium-aluminum spinel nanocrystallites doped with tetrahedrally coordinated Co2+ ions. Q-switched pulses of up to 5.5 mJ in energy and 80 ns in duration at 1.54 micrometers were achieved. Relaxation time of the 4A2 to 4T1(4F) transition bleaching was measured to be (450+/- 150)ns. Ground-state absorption cross-sections at 1.54 micrometers wavelength were estimated to be (3.2+/- 0.4)*10-19 cm2 and (5.0+/- 0.6)X10-20 cm2, respectively. Results of study absorption and luminescence spectra of different glass ceramics on the base of magnesium-aluminum, zinc-aluminum, lithium-aluminum spinel nanocrystallites doped with tetrahedrally coordinated Co2+ ions are also analyzed.

  19. Structural classification of phosphate glasses

    International Nuclear Information System (INIS)

    A structural classification of phosphate glasses is proposed. Following types of phosphate glasses are distinguished: discontinuous polymeric structure glasses (phosphate and mixed chains and rings containing glasses), continuous spatial network structure glasses (ultraphosphate and mixed network glasses) and non-polymeric structure glasses (oxide-halide and halide glasses, stuffed with ortho- and pyrophosphate-like groups). Type of the structure determines in a considerable degree the relation between glass composition and properties. (author). 25 refs

  20. Thermal phase stability of some simulated Defense waste glasses

    International Nuclear Information System (INIS)

    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 4500C to 11000C. Additionally, a specimen of each glass was cooled at a constant cooling rate of 70C/hour from an 11000C melt down to 5000C where it was removed from the furnace. The following were observed. The slow cooling rate of 70C/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 CeO2 and (Ni, Mn, Fe)2O4 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 5000C there is but little devitrification occurring implying that glass canisters stored at 3000C may be kinetically stable despite not being thermodynamically so

  1. Hualu Aluminum Will Construct Large Coal-Power-Aluminum Aluminum Processing Industrial Chain

    Institute of Scientific and Technical Information of China (English)

    2015-01-01

    The reporter learned from relevant departments of Baiyin City that in order to further push forward industrial upgrading,fulfill expansion and consolidation of the enterprise,Gansu Hualu Aluminum Co.,Ltd(Hualu Aluminum)will implement Out-Of-City-Into-Park project,

  2. Development of measurement system for the verification of the performance assessment of H.L.W. repository

    International Nuclear Information System (INIS)

    Full text of publication follows: The performance of the H.L.W. (High Level Waste) repository depends on the functions of the multi-barrier system that consists on both engineered barrier and natural barrier of geosphere. In particular, the latter barrier plays an important role for isolation of radioactive materials and this function gives an incentive to dispose H.L.W. into deep geologic environment. One of the key functions of the deep geologic formation is subjected by the nearly stagnant flow of underground-water. The preliminary estimation, for instance, suggests the flow rate is as low as 10-10 - 10-8 m/s which is several orders of magnitude lower than the detection limit by conventional measurement techniques. This objection may underestimate the repository performance due to the lack of the actual data of such a low flow rate. The objective of this study is to develop the in-situ measurement method of such a low flow rate in deep underground with a new concept using the solid particle tracer coupled with the ultrasonic detection system. The solid particle tracer is designed capable to adjust its density, which is identical to the underground-water. The tracer is injected into a borehole and flows along with the underground-water. The moving track of the tracer is then monitored by the three-dimensional ultrasonic detection system, which utilizes the matrix sensor and the high-speed numerical visualization device. The integrated flow detection system is designed applicable to wide range of flow rate 10-10 - 10-5 m/s consequently this newly developed system can enhance the confidence of the isolation capability of the H.L.W. disposal system. (authors)

  3. Issues at Stake When Considering Long Term Storage of HLW a Comprehensive Approach to Designing the Facility

    International Nuclear Information System (INIS)

    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 dual-barrier system be considered. In the case of spent fuel, CEA's early assessment of the long-term behavior of cladding shows that it may not 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

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

    International Nuclear Information System (INIS)

    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)

  5. Need for USA high level waste (HLW) alternate geological repository (AGR) and for a different methodology to enhance its acceptance

    International Nuclear Information System (INIS)

    In early February 2010, the administration stopped work and withdrew the Department of Energy (DOE) application for a construction permit for the Yucca Mountain geological repository from the Nuclear Regulatory Commission (NRC). Also, a 'blue ribbon' Commission was appointed to explore alternatives for storage, processing, and disposal, including evaluation of advanced fuel cycles and to provide a final report in 24 months. That decision, however, failed to recognize that: (1) the U.S. will need an early alternate geological repository (AGR) for its HLW irrespective of the findings of the 'blue ribbon' Commission; (2) the once-through spent fuel inventory from commercial nuclear power reactors will continue to rise and so will the damages against the government for its failure to remove spent fuel from reactors sites, as specified in contracts; (3) there are prepackaged DOE and nuclear weapons HLW ready for shipment to a repository which must be taken into account because of government penalties for failure to do so; (4) the current Nuclear Waste Policy Act (NWPA) needs to be modified to allow the early search and approval of Alternate Geological Repository (AGR) and for an interim centralized HLW storage facility to reduce government liabilities; and (5) the methodology used to license Yucca Mountain needs to undergo serious modifications, including a different non-politicized management and siting credo. This paper reviews and discusses all the preceding shortcomings and proposes significant changes to pursue AGR as soon as possible and to get site approval by the NRC first under a formal, stepwise, well-structured risk-informed decision approach as recommended.

  6. Setting up a safe deep repository for long-lived HLW and ILW in Russia: Current state of the works

    International Nuclear Information System (INIS)

    The concept of RW disposal in Russia in accordance with the Federal Law 'On Radioactive Waste Management and Amendments to Specific Legal Acts of the Russian Federation' No. 190-FL dated 11 July 2011, is oriented at the ultimate disposal of waste, without an intent for their subsequent retrieval. The law 190-FL has it as follows: - A radioactive waste repository is a radioactive waste storage facility intended for disposal of the radioactive wastes without an intent for their subsequent retrieval. - Disposal of solid long-lived high-level waste and solid long-lived intermediate-level waste is carried out in deep repositories for radioactive waste. - Import into the Russian Federation of radioactive waste for the purpose of its storage, processing and disposal, except for spent sealed sources of ionising radiation originating from the Russian Federation, is prohibited. For safe final disposal of long-lived HLW and ILW, it is planned to construct a deep repository for radioactive waste (DRRW) in a low-pervious monolith rock massif in the Krasnoyarsk region in the production territory of the Mining and Chemical Combine (FSUE 'Gorno-khimicheskiy kombinat'). According to the IAEA recommendations and in line with the international experience in feasibility studies for setting up of HLW and SNF underground disposal facilities, the first mandatory step is the construction of an underground research laboratory. An underground laboratory serves the following purposes: - itemised research into the characteristics of enclosing rock mass, with verification of massive material suitability for safe disposal of long-lived HLW and ILW; - research into and verification of the isolating properties of an engineering barrier system; - development of engineering solutions and transportation and process flow schemes for construction and running of a future RW ultimate isolation facility. (authors)

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

    International Nuclear Information System (INIS)

    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)

  8. lead glass brick

    CERN Multimedia

    When you look through the glass at a picture behind, the picture appears raised up because light is slowed down in the dense glass. It is this density (4.06 gcm-3) that makes lead glass attractive to physicists. The refractive index of the glass is 1.708 at 400nm (violet light), meaning that light travels in the glass at about 58% its normal speed. At CERN, the OPAL detector uses some 12000 blocks of glass like this to measure particle energies.

  9. Aluminum Zintl anion moieties within sodium aluminum clusters

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Haopeng; Zhang, Xinxing; Ko, Yeon Jae; Grubisic, Andrej; Li, Xiang; Ganteför, Gerd; Bowen, Kit H., E-mail: AKandalam@wcupa.edu, E-mail: kiran@mcneese.edu, E-mail: kbowen@jhu.edu [Department of Chemistry, Johns Hopkins University, Baltimore, Maryland 21218 (United States); Schnöckel, Hansgeorg [Institute of Inorganic Chemistry, Karlsruhe Institute of Technology, 76128 Karlsruhe (Germany); Eichhorn, Bryan W. [Department of Chemistry, University of Maryland at College Park, College Park, Maryland 20742 (United States); Lee, Mal-Soon; Jena, P. [Department of Physics, Virginia Commonwealth University, Richmond, Virginia 23284 (United States); Kandalam, Anil K., E-mail: AKandalam@wcupa.edu, E-mail: kiran@mcneese.edu, E-mail: kbowen@jhu.edu [Department of Physics, West Chester University of Pennsylvania, West Chester, Pennsylvania 19383 (United States); Kiran, Boggavarapu, E-mail: AKandalam@wcupa.edu, E-mail: kiran@mcneese.edu, E-mail: kbowen@jhu.edu [Department of Chemistry, McNeese State University, Lake Charles, Louisiana 70609 (United States)

    2014-02-07

    Through a synergetic combination of anion photoelectron spectroscopy and density functional theory based calculations, we have established that aluminum moieties within selected sodium-aluminum clusters are Zintl anions. Sodium–aluminum cluster anions, Na{sub m}Al{sub n}{sup −}, were generated in a pulsed arc discharge source. After mass selection, their photoelectron spectra were measured by a magnetic bottle, electron energy analyzer. Calculations on a select sub-set of stoichiometries provided geometric structures and full charge analyses for both cluster anions and their neutral cluster counterparts, as well as photodetachment transition energies (stick spectra), and fragment molecular orbital based correlation diagrams.

  10. Experimental Test of Stainless Steel Wire Mesh and Aluminium Alloy With Glass Fiber Reinforcement Hybrid Composite

    OpenAIRE

    Ranga Raj R.,; Velmurugan R

    2015-01-01

    At present, composite materials are mostly used in aircraft structural components, because of their excellent properties like lightweight, high strength to weight ratio, high stiffness, and corrosion resistance and less expensive. In this experimental work, the mechanical properties of laminate, this is reinforced with stainless steel wire mesh, aluminum sheet metal, perforated aluminum sheet metal and glass fibers to be laminate and investigated. The stainless steel wire mesh and...

  11. Spray Rolling Aluminum Strip

    Energy Technology Data Exchange (ETDEWEB)

    Lavernia, E.J.; Delplanque, J-P; McHugh, K.M.

    2006-05-10

    Spray forming is a competitive low-cost alternative to ingot metallurgy for manufacturing ferrous and non-ferrous alloy shapes. It produces materials with a reduced number of processing steps, while maintaining materials properties, with the possibility of near-net-shape manufacturing. However, there are several hurdles to large-scale commercial adoption of spray forming: 1) ensuring strip is consistently flat, 2) eliminating porosity, particularly at the deposit/substrate interface, and 3) improving material yield. Through this program, a new strip/sheet casting process, termed spray rolling, has been developed, which is an innovative manufacturing technique to produce aluminum net-shape products. Spray rolling combines the benefits of twin-roll casting and conventional spray forming, showing a promising potential to overcome the above hurdles associated with spray forming. Spray rolling requires less energy and generates less scrap than conventional processes and, consequently, enables the development of materials with lower environmental impacts in both processing and final products. Spray Rolling was developed as a collaborative project between the University of California-Davis, the Colorado School of Mines, the Idaho National Engineering and Environmental Laboratory, and an industry team. The following objectives of this project were achieved: (1) Demonstration of the feasibility of the spray rolling process at the bench-scale level and evaluation of the materials properties of spray rolled aluminum strip alloys; and (2) Demonstration of 2X scalability of the process and documentation of technical hurdles to further scale up and initiate technology transfer to industry for eventual commercialization of the process.

  12. 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

  13. Oxynitride glass fibers

    Science.gov (United States)

    Patel, Parimal J.; Messier, Donald R.; Rich, R. E.

    1991-01-01

    Research at the Army Materials Technology Laboratory (AMTL) and elsewhere has shown that many glass properties including elastic modulus, hardness, and corrosion resistance are improved markedly by the substitution of nitrogen for oxygen in the glass structure. Oxynitride glasses, therefore, offer exciting opportunities for making high modulus, high strength fibers. Processes for making oxynitride glasses and fibers of glass compositions similar to commercial oxide glasses, but with considerable enhanced properties, are discussed. We have made glasses with elastic moduli as high as 140 GPa and fibers with moduli of 120 GPa and tensile strengths up to 2900 MPa. AMTL holds a U.S. patent on oxynitride glass fibers, and this presentation discusses a unique process for drawing small diameter oxynitride glass fibers at high drawing rates. Fibers are drawn through a nozzle from molten glass in a molybdenum crucible at 1550 C. The crucible is situated in a furnace chamber in flowing nitrogen, and the fiber is wound in air outside of the chamber, making the process straightforward and commercially feasible. Strengths were considerably improved by improving glass quality to minimize internal defects. Though the fiber strengths were comparable with oxide fibers, work is currently in progress to further improve the elastic modulus and strength of fibers. The high elastic modulus of oxynitride glasses indicate their potential for making fibers with tensile strengths surpassing any oxide glass fibers, and we hope to realize that potential in the near future.

  14. Ultrahigh vacuum system with aluminum

    International Nuclear Information System (INIS)

    A bakeable vacuum chamber (1500C continuous) consists of aluminum alloy beam pipe (6063-T6) and bellows (5052-F) with an aluminum alloy flange (2219-T87) and a metal seal [Helicoflex-HN: pure aluminum (1050) O-ring with an elastic core (Ni base super alloy Inconel 750) which supplies the sealing force] has been constructed. The beam pipe and the flange (6063-T6/2219-T87), and the bellows and the flange (5052-F/2219-T87) were welded by an alternate current (50 Hz) TIG process using an aluminum alloy filler wire (4043). The mechanical properties of the aluminum alloy (2219-T87) is suitable for using the Helicoflex O-ring but the groove surface for the gasket is weak for scratching. Cromium-nitride coating by ion plating method was carried out on the aluminum surface of the gasket groove [thickness: 16 μm, micro Vickers hardness: 1800]. Ordinary stainless steel vacuum system can be replaced by the aluminum vacuum system in an accelerator. (author)

  15. Effect of alumina on the dissolution rate of glasses; Role de l'alumine sur la vitesse initiale de dissolution des verres

    Energy Technology Data Exchange (ETDEWEB)

    Palavit, G.; Montagne, L. [Ecole Nationale Superieure de Chimie de Lille, Lab. de Cristallochimie et Physicochimie du Solide, URA CNRS 0452, 59 - Villeneuve d' Ascq (France)

    1997-07-01

    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{sup 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{sup 3+} is dependant upon the nature of the glass matrix. (author)

  16. [Microbiological corrosion of aluminum alloys].

    Science.gov (United States)

    Smirnov, V F; Belov, D V; Sokolova, T N; Kuzina, O V; Kartashov, V R

    2008-01-01

    Biological corrosion of ADO quality aluminum and aluminum-based construction materials (alloys V65, D16, and D16T) was studied. Thirteen microscopic fungus species and six bacterial species proved to be able to attack aluminum and its alloys. It was found that biocorrosion of metals by microscopic fungi and bacteria was mediated by certain exometabolites. Experiments on biocorrosion of the materials by the microscopic fungus Alternaria alternata, the most active biodegrader, demonstrated that the micromycete attack started with the appearance of exudate with pH 8-9 on end faces of the samples. PMID:18669265

  17. Microstructuring of glasses

    CERN Document Server

    Hülsenberg, Dagmar; Bismarck, Alexander

    2008-01-01

    As microstructured glass becomes increasingly important for microsystems technology, the main application fields include micro-fluidic systems, micro-analysis systems, sensors, micro-actuators and implants. And, because glass has quite distinct properties from silicon, PMMA and metals, applications exist where only glass devices meet the requirements. The main advantages of glass derive from its amorphous nature, the precondition for its - theoretically - direction-independent geometric structurability. Microstructuring of Glasses deals with the amorphous state, various glass compositions and their properties, the interactions between glasses and the electromagnetic waves used to modify it. Also treated in detail are methods for influencing the geometrical microstructure of glasses by mechanical, chemical, thermal, optical, and electrical treatment, and the methods and equipment required to produce actual microdevices.

  18. The Velocity Analysis of Woven Glass Fiber Composites Using Cross-correlation Properties

    International Nuclear Information System (INIS)

    This paper discusses experimental results obtained by the potentiality of cross-correlation function as a tool for analyzing propagation of wave in an aluminum and a woven glass fiber composite. Each propagated wave has its own characteristic time delay, and examination of the cross-correlation of input and output signal give the most proper wave velocity and significant path. Using the above distinctive features, we observed the propagation velocity for the aluminum alloy and a woven glass fiber composite more accurately and easily then the common methods. The fiber locations of this composite also determined by the basis of these results

  19. Chalcogenide glass microsphere laser

    OpenAIRE

    Elliott, Gregor R.; Murugan, G.Senthil; Wilkinson, James S.; Zervas, Michalis N.; Hewak, Daniel W.

    2010-01-01

    Laser action has been demonstrated in chalcogenide glass microsphere. A sub millimeter neodymium-doped gallium lanthanum sulphide glass sphere was pumped at 808 nm with a laser diode and single and multimode laser action demonstrated at wavelengths between 1075 and 1086 nm. The gallium lanthanum sulphide family of glass offer higher thermal stability compared to other chalcogenide glasses, and this, along with an optimized Q-factor for the microcavity allowed laser action to be achieved. When...

  20. Homogeneity of Inorganic Glasses

    DEFF Research Database (Denmark)

    Jensen, Martin; Zhang, L.; Keding, Ralf;

    2011-01-01

    Homogeneity of glasses is a key factor determining their physical and chemical properties and overall quality. However, quantification of the homogeneity of a variety of glasses is still a challenge for glass scientists and technologists. Here, we show a simple approach by which the homogeneity o...

  1. Recycling of Glass

    DEFF Research Database (Denmark)

    Christensen, Thomas Højlund; Damgaard, Anders

    2011-01-01

    Glass is used for many purposes, but in the waste system glass is predominantly found in terms of beverage and food containers with a relatively short lifetime before ending up in the waste. Furthermore there is a large amount of flat glass used in building materials which also ends up in the waste...

  2. Technique for mechanical evaluation of long-term behavior of the near-field of HLW disposal repository

    International Nuclear Information System (INIS)

    The targets of this study are techniques of mechanical evaluation of long-term behavior of the near-field of HLW disposal repository, which are essential for safe construction and exact evaluation of nuclide transport. Since the facility is in great depth and under high temperature by the heat from the HLW, and needs evaluation of its longtime behavior, it is quite different from the conventional underground facilities. Therefore, (1) deep rock stress measurement technique, (2) investigation of mechanical properties of soft rock under high temperature, (3) technique of prediction of longtime settlement of overpack in bentonite buffer and (4) numerical simulation code are studied and developed. (1) The rock stress measurement technique has been improved so that it can take into account the anisotropy of soft rock. (2) Strength, deformability and creep characteristics of soft rock under high temperature up to 80 deg C have been clarified. (3) A method of determining the mechanical properties of bentonite from physicochemical properties has been established and verified through centrifugal settlement tests. (4) A thermo-hydro-mechanical coupling simulation code has been developed which includes the sophisticated creep model for soft rock and swelling model for bentonite. (author)

  3. Magnetic properties of iron films on anodized aluminum underlayer

    International Nuclear Information System (INIS)

    A one-step anodization process was used to prepare the anodic alumina (AA) film on glass. Using the AA as an inserted underlayer, iron films with thickness of tN in the range of 10 ∼ 35 nm were deposited by argon ion sputtering. The iron film deposited on AA underlayer exhibited different magnetic behaviors from the iron film deposited on glass or on aluminum underlayer. The perpendicular coercivity of film deposited on AA underlayer reached a maximal value of about 1 kOe at tN = 30 nm. We believe that the improvement of magnetic properties came from the modulation of the morphology of Fe film by the porous structure of AA underlayer.

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

    International Nuclear Information System (INIS)

    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

  5. Chrome - Free Aluminum Coating System

    Science.gov (United States)

    Bailey, John H.; Gugel, Jeffrey D.

    2010-01-01

    This slide presentation concerns the program to qualify a chrome free coating for aluminum. The program was required due to findings by OSHA and EPA, that hexavalent chromium, used to mitigate corrosion in aerospace aluminum alloys, poses hazards for personnel. This qualification consisted of over 4,000 tests. The tests revealed that a move away from Cr+6, required a system rather than individual components and that the maximum corrosion protection required pretreatment, primer and topcoat.

  6. Refinement of Aluminum Thermal Chrome

    International Nuclear Information System (INIS)

    Refinement of aluminum thermal chrome of the X98.5 mark by a high-temperature annealing in high vacuum is explored experimentally. It is shown that at the temperature of annealing 1150 C during 1...6 hours the content of such interstitial impurity as nitrogen is essentially depressed in chrome, and also the content of aluminum and iron admixtures is noticeably moderated

  7. Gas evolution in aluminum electrolytic capacitors

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-Aleixandre, C.; Albella, J.M.; Martinez-Duart, J.M.

    1984-03-01

    Gas evolution in aluminum electrolytic capacitors constitutes one of their main drawbacks in comparison to other types of capacitors lacking a liquid electrolyte. In this respect, one of the most common causes of failure shown by liquid electrolyte capacitors is electrolyte leakage through the seal or even explosions produced by internal pressure buildup. In order to prevent these hazards, some substances, known as depolarizers, are usually added to the capacitor electrolyte with the purpose of absorbing the hydrogen evolved at the cathode (1, 2). Although the gas evolution problem in electrolytic capacitors has been known for a long time, there is a lack of literature on both direct measurements of the gas evolved and assessments of the amount of depolarizer active for the hydrogen absorption process. Aluminum electrolytic capacitors of 100..mu..F and 40V nominal voltage, miniature type (diam 8 mm, height 18.5 mm), were manufactured under standard specifications. The capacitors were filled with about 0.5 ml of an electrolyte consisting essentially of a solution of boric, adipic, and phosphoric acids in ethylene glycol. Picric acid and p-benzoquinone in molar concentrations of 0.01M and 0.05M, respectively, were added as depolarizers, yielding an electrolyte with a resistivity of about 80 ..cap omega..-cm and a pH of 5.1. The pressure inside the capacitors was monitored by a conventional Ushaped manometer made from a capillary glass tube filled with distilled water. The number of mols of gas generated in the capacitor (/eta/ /SUB g/ ) was calculated from the measured pressure (sensitivity 0.1 mm Hg) and the value of the internal volume of the manometercapacitor system.

  8. Spectroscopic properties of quartz glasses

    International Nuclear Information System (INIS)

    Full text: In this work the results of influence of 60Co gamma-radiation on spectroscopic properties of KI type quartz glasses are presented. The samples were cut in the plate form with 3 mm thickness and 25 mm diameter preliminary, the surfaces were ground and polished. First of all it was necessary to study optical absorption of these glasses and define absorption bands (AB) if they belong to defect or impurity centers. For this purpose the optical absorption spectra and X-ray luminescence spectra (XL) of initial not irradiated samples and then after 60Co gamma-irradiation to the maximal dose of 1.2x108 R at temperature 300 K were measured. The analyses of the absorption spectra showed that at this dose the intensity of AB increases sharply. It is probably caused by increase of the centers which existed in not irradiated samples. After gamma-irradiation of KI type quartz glasses the AB at 215 nm appears in UV spectral range, which is caused by three-coordinated silicon atoms with a non coupled electron (E/-centre). With increasing of irradiation dose the AB at 215 nm increases that testifies about radiation induced creation of structural defects responsible for E/ centers. It has also been established that under gamma-irradiation an additional AB appears at 550 nm, caused by localization of holes on oxygen of those tetrahedrons where aluminum atoms replace silicon, the intensity which strongly depends on the absorbed dose too. These centers can cause optical luminescence. Indeed, XL luminescence spectra contain the bands at 402 nm and 550 nm which have an impurity character. According to the literary data, these bands are absent in crystal quartz. These centers are probably connected with non bridging oxygen ions existing in quartz glasses due to presence of metal impurity. The XL band at 402 nm belongs to - Si-Ge- type defects. The comparison of the results with the literary data allow us to assume that Al and Ge impurities play the special role in appearing of

  9. Fluoride glass fiber optics

    CERN Document Server

    Aggarwal, Ishwar D

    1991-01-01

    Fluoride Glass Fiber Optics reviews the fundamental aspects of fluoride glasses. This book is divided into nine chapters. Chapter 1 discusses the wide range of fluoride glasses with an emphasis on fluorozirconate-based compositions. The structure of simple fluoride systems, such as BaF2 binary glass is elaborated in Chapter 2. The third chapter covers the intrinsic transparency of fluoride glasses from the UV to the IR, with particular emphasis on the multiphonon edge and electronic edge. The next three chapters are devoted to ultra-low loss optical fibers, reviewing methods for purifying and

  10. Multiple Glass Ceilings

    OpenAIRE

    Russo, Giovanni; Hassink, Wolter

    2011-01-01

    Both vertical (between job levels) and horizontal (within job levels) mobility can be sources of wage growth. We find that the glass ceiling operates at both margins. The unexplained part of the wage gap grows across job levels (glass ceiling at the vertical margin) and across the deciles of the intra-job-level wage distribution (glass ceiling at the horizontal margin). This implies that women face many glass ceilings, one for each job level above the second, and that the glass ceiling is a p...

  11. BENEFITS OF VIBRATION ANALYSIS FOR DEVELOPMENT OF EQUIPMENT IN HLW TANKS - 12341

    Energy Technology Data Exchange (ETDEWEB)

    Stefanko, D.; Herbert, J.

    2012-01-10

    Vibration analyses of equipment intended for use in the Savannah River Site (SRS) radioactive liquid waste storage tanks are performed during pre-deployment testing and has been demonstrated to be effective in reducing the life-cycle costs of the equipment. Benefits of using vibration analysis to identify rotating machinery problems prior to deployment in radioactive service will be presented in this paper. Problems encountered at SRS and actions to correct or lessen the severity of the problem are discussed. In short, multi-million dollar cost saving have been realized at SRS as a direct result of vibration analysis on existing equipment. Vibration analysis of equipment prior to installation can potentially reduce inservice failures, and increases reliability. High-level radioactive waste is currently stored in underground carbon steel waste tanks at the United States Department of Energy (DOE) Savannah River Site and at the Hanford Site, WA. Various types of rotating machinery (pumps and separations equipment) are used to manage and retrieve the tank contents. Installation, maintenance, and repair of these pumps and other equipment are expensive. In fact, costs to remove and replace a single pump can be as high as a half million dollars due to requirements for radioactive containment. Problems that lead to in-service maintenance and/or equipment replacement can quickly exceed the initial investment, increase radiological exposure, generate additional waste, and risk contamination of personnel and the work environment. Several different types of equipment are considered in this paper, but pumps provide an initial example for the use of vibration analysis. Long-shaft (45 foot long) and short-shaft (5-10 feet long) equipment arrangements are used for 25-350 horsepower slurry mixing and transfer pumps in the SRS HLW tanks. Each pump has a unique design, operating characteristics and associated costs, sometimes exceeding a million dollars. Vibration data are routinely

  12. Experimental programme to demonstrate the viability of the supercontainer concept for HLW

    International Nuclear Information System (INIS)

    The EIG EURIDICE (a joint venture between the Belgian Organisation for Radioactive Waste Management - ONDRAF/NIRAS - and the Belgian Nuclear Research Centre - SCKoCEN) is responsible for performing large-scale tests, technical demonstrations and experiments to assess the feasibility of a final disposal of vitrified radioactive waste in deep clay layers. This is part of the Belgian Research and Development programme managed by ONDRAF/NIRAS. The current Belgian reference design for vitrified HLW and spent fuel assemblies is the so-called Supercontainer design. The vitrified waste canisters or spent fuel assemblies are enclosed in a carbon steel overpack which has to prevent contact between water from the host formation and the waste during the thermal phase. In order to maintain favourable chemical conditions to avoid corrosion during this period (several hundred or even thousand of years), the overpack is surrounded by a high alkaline concrete buffer of about 70 cm thick. The buffer also provides permanent radiological shielding for the workers, simplifying handling and other operations. All the components of the Supercontainer are constructed in above ground installations, thus creating favourable QA/QC conditions. After the emplacement of the Supercontainers in the disposal galleries, the remaining space will be backfilled. Tests to demonstrate the viability and the construction feasibility of the supercontainer design have been initiated. The viability programme includes Tests to verify the feasibility to construct and emplace the components of the supercontainers, and tests to verify the feasibility to backfill the disposal galleries once the supercontainers are placed. Supercontainer construction: Tests in column to verify the construction feasibility (risk of cracking) of the buffer with two different types of concrete (a self-compacting concrete - SCC - and a rheoplastic concrete RPC) were performed in collaboration with the Belgian concrete factory Socea. A

  13. An Assessment of Using Vibrational Compaction of Calcined HLW and LLW in DWPF Canisters

    International Nuclear Information System (INIS)

    both of them) of applying the vibrational forces? 2) What is best mode of operation: first fill the canister with calcined waste and then vibrate it and refill it again, or apply vibrational forces during the filling process. By optimum or best we mean less creation of stress/strain forces during the volume reduction vibration process. Lessons learnt: This preliminary study shows that; 1) The maximum stress concentration always occurs in the canister wall, however its location varies and depends on the loading condition, and vibration process. 2) The proposed vibrational process would not cause any damages to the granulated calcined waste. 3) The first natural frequency of the longitudinal vibration of the canister is around 400 Hz, which is far away from the applied vibrational frequencies and from possibility of resonance phenomena that may cause damage to the canister 4) The relationship between the maximum internal stress and the frequency of the applied load is not parabolic. 5) The mechanical properties of the granulated calcined nuclear waste have small impact on the internal stress of the canister. Finally, the calculated data suggested that applying vibrational forces will keep the entire canister whole without any indication of development defects, and will have significant economical benefits of handling HLW and LLW in calcined forms, from waste manipulation, storage and transportation

  14. Risk and uncertainty assessment for a potential HLW repository in Korea: TSPA 2006

    International Nuclear Information System (INIS)

    KAERI has worked on the concept development on permanent disposal of HLW and its total system performance assessment since 1997. More than 36 000 MT of spent nuclear fuel from PWR and CANDU reactors is planned to be disposed of in crystalline bed-rocks. The total system performance assessment (TSPA) tools are under development. The KAERI FEP encyclopedia is actively developed to include all potential FEP suitable for Korean geo- and socio conditions. The FEPs are prioritized and then categorized to the intermediate level FEP groups. These groups become elements of the rock engineering system (RES) matrix. Then the sub-scenarios such as a container failure, groundwater migration, solute transport, etc are developed by connecting interactions between diagonal elements of the RES matrix. The full scenarios are developed from the combination of sub-scenarios. For each specific scenario, the assessment contexts and associated assessment method flow charts are developed. All information on these studies is recorded into the web based programme, FEAS (FEP to Assessment through Scenarios.) KAERI applies three basic programmes for the post closure radionuclide transport calculations; MASCOT-K, AMBER, and the new MDPSA under development. The MASCOT-K originally developed by Serco for a LLW repository has been extended extensively by KAERI to simulate release reactions such as congruent and gap releases in spent nuclear fuel. The new MDPSA code is dedicated for the probabilistic assessment of radio-nuclides in multi-dimensions of a fractured porous medium. To acquire input data for TSPA domestic experiment programmes as well as literature survey are performed. The data are stored in the Performance Assessment Input Data system (PAID.) To assure the transparency, traceability, retrievability, reproducibility, and review (T2R3) the web based KAERI QA system is developed. All tasks in TSPA are recorded under the concept of a 'Project' in this web system. Currently, FEAS, PAID

  15. Benefits Of Vibration Analysis For Development Of Equipment In HLW Tanks - 12341

    International Nuclear Information System (INIS)

    Vibration analyses of equipment intended for use in the Savannah River Site (SRS) radioactive liquid waste storage tanks are performed during pre-deployment testing and has been demonstrated to be effective in reducing the life-cycle costs of the equipment. Benefits of using vibration analysis to identify rotating machinery problems prior to deployment in radioactive service will be presented in this paper. Problems encountered at SRS and actions to correct or lessen the severity of the problem are discussed. In short, multi-million dollar cost saving have been realized at SRS as a direct result of vibration analysis on existing equipment. Vibration analysis of equipment prior to installation can potentially reduce inservice failures, and increases reliability. High-level radioactive waste is currently stored in underground carbon steel waste tanks at the United States Department of Energy (DOE) Savannah River Site and at the Hanford Site, WA. Various types of rotating machinery (pumps and separations equipment) are used to manage and retrieve the tank contents. Installation, maintenance, and repair of these pumps and other equipment are expensive. In fact, costs to remove and replace a single pump can be as high as a half million dollars due to requirements for radioactive containment. Problems that lead to in-service maintenance and/or equipment replacement can quickly exceed the initial investment, increase radiological exposure, generate additional waste, and risk contamination of personnel and the work environment. Several different types of equipment are considered in this paper, but pumps provide an initial example for the use of vibration analysis. Long-shaft (45 foot long) and short-shaft (5-10 feet long) equipment arrangements are used for 25-350 horsepower slurry mixing and transfer pumps in the SRS HLW tanks. Each pump has a unique design, operating characteristics and associated costs, sometimes exceeding a million dollars. Vibration data are routinely

  16. Anodized aluminum on LDEF

    Science.gov (United States)

    Golden, Johnny L.

    1993-01-01

    A compilation of reported analyses and results obtained for anodized aluminum flown on the Long Duration Exposure Facility (LDEF) was prepared. Chromic acid, sulfuric acid, and dyed sulfuric acid anodized surfaces were exposed to the space environment. The vast majority of the anodized surface on LDEF was chromic acid anodize because of its selection as a thermal control coating for use on the spacecraft primary structure, trays, tray clamps, and space end thermal covers. Reports indicate that the chromic acid anodize was stable in solar absorptance and thermal emittance, but that contamination effects caused increases in absorptance on surfaces exposed to low atomic oxygen fluences. There were some discrepancies, however, in that some chromic acid anodized specimens exhibited significant increases in absorptance. Sulfuric acid anodized surfaces also appeared stable, although very little surface area was available for evaluation. One type of dyed sulfuric acid anodize was assessed as an optical baffle coating and was observed to have improved infrared absorptance characteristics with exposure on LDEF.

  17. EIS Data Call Report: Plutonium immobilization plant using glass in new facilities at the Savannah River Site

    International Nuclear Information System (INIS)

    The Plutonium Immobilization Plant (PIP) accepts plutonium (Pu) from pit conversion and from non-pit sources and, through a glass immobilization process, converts the plutonium into an immobilized form that can be disposed of in a high level waste (HLW) repository. This immobilization process is shown conceptually in Figure 1-1. The objective is to make an immobilized form, suitable for geologic disposal, in which the plutonium is as inherently unattractive and inaccessible as the plutonium in spent fuel from commercial reactors

  18. Low-aluminum content iron-aluminum alloys

    Energy Technology Data Exchange (ETDEWEB)

    Sikka, V.K.; Goodwin, G.M.; Alexander, D.J. [and others

    1995-06-01

    The low-aluminum-content iron-aluminum program deals with the development of a Fe-Al alloy with aluminum content such as a produce the minimum environmental effect at room temperature. The FAPY is an Fe-16 at. % Al-based alloy developed at the Oak Ridge National Laboratory as the highest aluminum-containing alloy with essentially no environmental effect. The chemical composition for FAPY in weight percent is: aluminum = 8.46, chromium = 5.50, zirconium = 0.20, carbon = 0.03, molybdenum = 2.00, yttrium = 0.10 and iron = 83.71. The ignots of the alloy can be hot worked by extrusion, forging, and rolling processes. The hot-worked cast structure can be cold worked with intermediate anneals at 800{degrees}C. Typical room-temperature ductility of the fine-grained wrought structure is 20 to 25% for this alloy. In contrast to the wrought structure, the cast ductility at room temperature is approximately 1% with a transition temperature of approximately 100 to 150{degrees}C, above which ductility values exceed 20%. The alloy has been melted and processed into bar, sheet, and foil. The alloy has also been cast into slabs, step-blocks of varying thicknesses, and shapes. The purpose of this section is to describe the welding response of cast slabs of three different thicknesses of FAPY alloy. Tensile, creep, and Charpy-impact data of the welded plates are also presented.

  19. 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

  20. Gas evolution behavior of aluminum in mortar

    International Nuclear Information System (INIS)

    As a part of study of leaching behavior for solidified dry low level radioactive waste, gas evolution behavior of aluminum in mortar was investigated, and a plan of our research was proposed. The effect of pH on corrosion rate of aluminum, corrosion product, time dependency of corrosion rate of aluminum in mortar, change of corrosion mechanism, the effects of Na, Ca and Cl ions on corrosion rate of aluminum in mortar and corrosion behavior of aluminum when aluminum was used as sacrificed anode in reinforced concrete were previously clarified. Study of the effects of environmental factors such as pH, kind of ions and temperature on gas evolution behavior of aluminum and the effect of aluminum/carbon steel surface ratio no gas evolution behavior of aluminum were planed. (author). 75 refs

  1. Effect of B2O3 on iron phosphate glass-ceramic wasteforms

    International Nuclear Information System (INIS)

    The effects of doping iron phosphate glass-ceramic wasteforms containing high level waste (HLW) with different B2O3 contents on their structures and properties were investigated. The chemical durability of glass-ceramic wasteforms was measured by dissolution rate (DR) method. The structure of the glass-ceramic wasteforms was analyzed by Fourier turning infrared (FTIR) and X-radiation diffraction (XRD). The results show that glass-ceramic wasteforms have the main crystalline phase of monazite. There are great effects of wasteforms on the chemical durability. The 28 d leaching rate of glass-ceramic wasteforms is 7.81 x 10-9 g/(cm2·min) when 10% (mole percent) Fe2O3 is replaced by B2O3 in wasteforms. Samples exist a large number of [PO4]3-, a small amount of [O2O7]4- and no [PO3]-. Boron in wasteforms samples is present as tetrahedral [BO4] units. (authors)

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

    International Nuclear Information System (INIS)

    Long half life and easy availability from high level wastes make 137Cesium 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 137Cs 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 137Cs loaded on AMP .Phosphate glasses containing Na2O, P2O5, B2O3, Fe2O3, Al2O3 and SiO2 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/cm2/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)

  3. Spatial extension and parameter integration into the safety case - examples from the Hungarian LLW/ILW and HLW disposal projects

    International Nuclear Information System (INIS)

    Exploration of potential sites for radioactive waste disposal of waste types LLW, ILW and HLW is underway worldwide. Nuclear disasters of the last decades and their spatial effects and temporal decay show that the long term safety of radioactive waste disposal systems, including recognition and modelling of future events, is of the highest importance. A key outcome of these site investigations is the generation of information required by Safety Cases, notably the identification of how future events might affect safety. Models developed for studies of safety must describe complex systems over long time frames and thus must deal with measured and calculated parameters and their associated uncertainties. However, there are some basic problems, especially in the first phase of the research, when the limited available information must be extended spatially and the uncertainties might not be well understood. This presentation will discuss two examples in connection with research on potential host rocks for Hungarian LLW/ILW and HLW disposal: the feasibility of parameter spatial extension and estimation of the uncertainty of permeability. The first example pertains to geological research on the Moragy Granite Formation (MGF), a potential host rock for Hungarian LLW/ILW disposal in Bataapati, and is concerned with the interdependent use of geological and geophysical information. During this project, boreholes were drilled and cores were analysed with general and special geological and geophysical methods. Indirect connections between physical parameters provided by geophysical measurements and geological characterisation of the different parts of MGF supported the estimation of spatial variability of the granite body. The second example deals with geological research on the Boda Clay-stone Formation (BCF), a potential rock body for HLW disposal in Hungary and describes a new laboratory method for gas and water permeability measurement. The well-conditioned cores originated

  4. Key issues identified from project TRU-2 on the generic co-location concept of transuranic (TRU) waste and high-level radioactive waste (HLW) repositories in Japan

    International Nuclear Information System (INIS)

    The Federation of the Electric Power Companies of Japan (FEPC) and the Japan Atomic Energy Agency (JAEA) have been collaborating with relevant organisations to promote the safe geological disposal of transuranic (TRU) waste following the already established disposal policy for high-level radioactive waste (HLW) in Japan. A result of this intensive collaborative effort was the production of a recent progress report (TRU-2) which describes the generic R and D for TRU-waste disposal in Japan. In order to improve feasibility and reduce costs and the burden on siting, the concept of co-locating TRU-waste and HLW repositories in a single complex was assessed in detail and compared with the results from several other countries that have also looked at co-location disposal. Heat from HLW, high pH plume(s) from the large amounts of cementitious materials used in the engineered barrier system (EBS) of TRU waste, and nitrates and organic materials in certain types of TRU waste were identified as critical reciprocal influences that might degrade the performance of the TRU/HLW co-location disposal system over the long-term. It was shown that these reciprocal influences could be avoided by establishing a separation distance between the two repositories of approximately 300 meters. (authors)

  5. Formulation of Japanese consensus-building model for HLW geological disposal site determination. 4. The influence of the accurate information on the decision making

    International Nuclear Information System (INIS)

    Investigation has been made to discuss how the accurate scientific information affects the perception of risk. To verify this investigation, dialogue seminars have been held. Based upon the outcomes of these investigations, the analysis of attribution was done to verify the factors affecting the risk perception and acceptance relevant to the consensus-building for HLW geological disposal site determination. (author)

  6. 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)

    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)

  7. Microstructure and mechanical properties of aluminum back contact layers

    Energy Technology Data Exchange (ETDEWEB)

    Popovich, V.A.; Janssen, M.; Richardson, I.M. [Delft University of Technology, Department of Materials Science and Engineering, Delft (Netherlands); van Amstel, T.; Bennett, I.J. [Energy Research Centre of the Netherlands, Solar Energy, PV Module Technology, Petten (Netherlands)

    2011-01-15

    The overall demand to reduce solar energy costs gives a continuous drive to reduce the thickness of silicon wafers. Handling and bowing problems associated with thinner wafers become more and more important, as these can lead to cells cracking and thus to high yield losses. In this paper the microstructure and mechanical properties of the aluminum on the rear side of a solar cell are discussed. It is shown that the aluminum back contact has a complex composite-like microstructure, consisting of five main components: (1) the back surface field layer; (2) a eutectic layer; (3) spherical (3-5 {mu}m) hypereutectic Al-Si particles surrounded by a thin aluminum oxide layer (200 nm); (4) a bismuth-silicate glass matrix; and (5) pores (14 vol%). The Young's modulus of the Al-Si particles is estimated by nanoindentation and the overall Young's modulus is estimated on the basis of bowing measurements. These results are used as input parameters for the improved thermomechanical multiscale model of a solar cell. (author)

  8. Experimental investigation of the effects of aqueous species on the dissolution kinetics of R7T7 glass

    International Nuclear Information System (INIS)

    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. Scaleable Clean Aluminum Melting Systems

    Energy Technology Data Exchange (ETDEWEB)

    Han, Q.; Das, S.K. (Secat, Inc.)

    2008-02-15

    The project entitled 'Scaleable Clean Aluminum Melting Systems' was a Cooperative Research and Development Agreements (CRADAs) between Oak Ridge National Laboratory (ORNL) and Secat Inc. The three-year project was initially funded for the first year and was then canceled due to funding cuts at the DOE headquarters. The limited funds allowed the research team to visit industrial sites and investigate the status of using immersion heaters for aluminum melting applications. Primary concepts were proposed on the design of furnaces using immersion heaters for melting. The proposed project can continue if the funding agency resumes the funds to this research. The objective of this project was to develop and demonstrate integrated, retrofitable technologies for clean melting systems for aluminum in both the Metal Casting and integrated aluminum processing industries. The scope focused on immersion heating coupled with metal circulation systems that provide significant opportunity for energy savings as well as reduction of melt loss in the form of dross. The project aimed at the development and integration of technologies that would enable significant reduction in the energy consumption and environmental impacts of melting aluminum through substitution of immersion heating for the conventional radiant burner methods used in reverberatory furnaces. Specifically, the program would couple heater improvements with furnace modeling that would enable cost-effective retrofits to a range of existing furnace sizes, reducing the economic barrier to application.

  10. Development of new radiopaque glass fiber posts.

    Science.gov (United States)

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

    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 b 0.05)when evaluatedwith one-way ANOVA statistical analysis. According to combustion analyses, the filler content of the tested posts ranges between 58.84wt.% and 86.02wt.%. 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. Futureglass 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. PMID:26652441

  11. Thermo-hydro-mechanical processes in the nearfield around a HLW repository in argillaceous formations. Vol. I. Laboratory investigations

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Chun-Liang; Czaikowski, Oliver; Rothfuchs, Tilmann; Wieczorek, Klaus

    2013-06-15

    All over the world, clay formations are being investigated as host medium for geologic disposal of radioactive waste because of their favourable properties, such as very low hydraulic conductivity against fluid transport, good sorption capacity for retardation of radionuclides, and high potential of self-sealing of fractures. The construction of a repository, the disposal of heat-emitting high-level radioactive waste (HLW), the backfilling and sealing of the remaining voids, however, will inevitably induce mechanical (M), hydraulic (H), thermal (T) and chemical (C) disturbances to the host formation and the engineered barrier system (EBS) over very long periods of time during the operation and post-closure phases of the repository. The responses and resulting property changes of the clay host rock and engineered barriers are to be well understood, characterized, and predicted for assessing the long-term performance and safety of the repository.

  12. Study on a transportation and emplacement system of pre-assembled EBS module for HLW geological disposal

    International Nuclear Information System (INIS)

    HLW disposal is one of the largest issue to utilize Nuclear power safely. In the past study, the concept, which buffer materials and Overpacked waste were transported into underground respectively, have shown. The concept of pre-assembled engineered barrier has advantage to simplify the logistics and emplacement procedure, however there are difficulties to support heavy weight of pre-assembled package by equipment under the condition of little clearance between tunnel and package. In this study, Combination of air bearing and two degree-of-freedom wheels were suggested for transportation, and air jack was suggested for unloading and emplacement system. Also, whole system for transportation and emplacement procedure was designed, and Scale model test was examined to evaluate the feasibility of these concept and functions. (author)

  13. Comparison of sodium zirconium phosphate-structured HLW forms and synroc for high-level nuclear waste immobilization

    International Nuclear Information System (INIS)

    The incorporation of (a) Cs/Sr as simulated heat-generating isotopes contained in Purex reprocessing waste, (b) simulated actinides, and (c) simulated Purex waste in sodium zirconium phosphate (NZP) has been studied. The samples were prepared by sintering, by hot pressing and by hot isostatic pressing in metal bellows containers. The short-term chemical durability of the phosphate-based material containing Purex waste was within an order of magnitude of that for Synroc-C, as measured by 7-day MCC-1 tests at 90 degrees C. The dissolution behavior showed evidence of re-precipitation phenomena, even after times as short as 28 days. Potential for improvement of NZP-based ceramics for HLW management is discussed. 19 refs., 4 figs., 3 tabs

  14. The treatment of gas in the performance assessment for the disposal of HLW and MLW in boom clay

    International Nuclear Information System (INIS)

    In Belgium the Boom Clay is studied as potential host rock for the geological disposal of high level (HLW) and intermediate level radioactive waste (ILW). The Boom Clay has been chosen because of its very low hydraulic conductivity (2 10-12 m/s). Consequently, transport of contaminants in pore water of the Boom Clay is diffusion-controlled whereas advection has a negligible contribution to the overall migration. Also the transport of dissolved gas is very limited. Therefore, when gas is generated this can easily lead to a gas pressure build-up and thus to a safety concern. In the following sections the experimental evidence about gas generation and transport in Boom Clay are briefly summarised, then the approach used to treat the gas issue, followed by the assessment of the gas generation and the assessment of its potential consequences. (authors)

  15. ENRESA strategy for the treatment of gas-related issues for HLW repositories sited in granitic and argillaceous settings

    International Nuclear Information System (INIS)

    Gas generation in HLW repositories is an unavoidable process whose consequences related with the global safety of the system will depend on many issues. Design aspects, materials used, waste form, rock type, behaviour of engineered barriers, etc. will condition the gas generated, amounts and generation rates, generation mechanisms, accumulation or dissipation from the waste form through the near field or transport in the rock. ENRESA has focused the research on gas related issues on those aspects which are relevant to the disposal concept foreseen in the Spanish program for high level radioactive waste management, which is a deep repository in granite, in which carbon steel canisters containing spent fuel assemblies will be placed in horizontal galleries surrounded by a bentonite buffer. According with this repository concept, several research projects have been initiated by ENRESA, most of them within an international cooperation framework. (authors)

  16. Thermodynamics of Glass Melting

    Science.gov (United States)

    Conradt, Reinhard

    First, a model based on linear algebra is described by which the thermodynamic properties of industrial multi-component glasses and glass melts can be accurately predicted from their chemical composition. The model is applied to calculate the heat content of glass melts at high temperatures, the standard heat of formation of glasses from the elements, and the vapor pressures of individual oxides above the melt. An E-fiber glass composition is depicted as an example. Second, the role of individual raw materials in the melting process of E-glass is addressed, with a special focus on the decomposition kinetics and energetic situation of alkaline earth carriers. Finally, the heat of the batch-to-melt conversion is calculated. A simplified reaction path model comprising heat turnover, content of residual solid matter, and an approach to batch viscosity is outlined.

  17. Operational safety and radiation protection considerations in designing an HLW repository in Germany

    International Nuclear Information System (INIS)

    In Germany the reference concept for disposal of heat generating radioactive waste considers emplacing canisters with vitrified waste in deep vertical boreholes drilled from the drifts of a repository mine in salt at a depth of 870 m. Spent fuel is to be disposed of in self-shielding POLLUX casks in horizontal drifts. An optimized disposal concept anticipates emplacing unshielded canisters with vitrified HLW and canisters containing the fuel rods of 3 PWR or 9 BWR fuel assemblies in boreholes with a diameter of 60 cm and a depth of up to 300 m.. In all cases the void space between POLLUX cask and drifts and canisters and borehole wall will be backfilled with crushed salt. (1) Operational Safety: Based on a detailed description of all underground disposal operation steps, the possible impacts on the disposal operations were analysed and the need for further studies determined. The disposal operation steps comprise e.g. rail bound transport from the shaft to the emplacement drift and emplacement process itself. As possible impacts the following occurrences were considered: ventilation failure, power supply failure, rock mechanics impact including cross-section convergence, irregular floor uplift and rock fall, brine and natural gas intrusion, derailing of transport carts and finally internal fire. (2) Radiation Protection: According to the German Atomic Energy Act (AtG), the design, construction and operation of a nuclear site like a final repository has to be licensed by the responsible authority. The Radiological Protection Ordinance and further guidelines i.e. concerning the emission and immission of released radioactive nuclides or the risk analysis of possible failure, build the basis for the licensing procedures. To ensure adequate protection against undue radiation exposure the repository is divided into different radiological protection areas. Generally, the handling of shielded waste packages above und under ground (including all the pathway of transport and

  18. Glass microspheres for brachytherapy

    International Nuclear Information System (INIS)

    We developed the capacity to produce glass microspheres containing in their structure one or more radioactive isotopes useful for brachytherapy. We studied the various facts related with their production: (Rare earth) alumino silicate glass making, glass characterization, microspheres production, nuclear activation through (n,γ) nuclear reactions, mechanical characterization before and after irradiation. Corrosion tests in simulated human plasma and mechanical properties characterization were done before and after irradiation. (author)

  19. Fractography of glass

    CERN Document Server

    Tressler, Richard

    1994-01-01

    As the first major reference on glass fractography, contributors to this volume offer a comprehensive account of the fracture of glass as well as various fracture surface topography Contributors discuss optical fibers, glass containers, and flatglass fractography In addition, papers explore fracture origins; the growth of the original flaws of defects; and macroscopic fracture patterns from which fracture patterns evolve This volume is complete with photographs and schematics

  20. Diamond turning of glass

    Energy Technology Data Exchange (ETDEWEB)

    Blackley, W.S.; Scattergood, R.O.

    1988-12-01

    A new research initiative will be undertaken to investigate the critical cutting depth concepts for single point diamond turning of brittle, amorphous materials. Inorganic glasses and a brittle, thermoset polymer (organic glass) are the principal candidate materials. Interrupted cutting tests similar to those done in earlier research are Ge and Si crystals will be made to obtain critical depth values as a function of machining parameters. The results will provide systematic data with which to assess machining performance on glasses and amorphous materials