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Sample records for hlw glass formulation

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    International Nuclear Information System (INIS)

    Jiang Yaozhong; Tang Baolong; Zhang Baoshan; Zhou Hui

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-12-30

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Kruger, Albert A.

    2013-01-01

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

  8. HLW immobilization in glass

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  10. 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

    in melter operating temperature. Glass composition development was based on one of the HLW waste compositions specified by ORP that has a high concentration of aluminum. Small-scale tests were used to provide an initial screening of various glass formulations with respect to melt rates; more definitive screening was provided by the subsequent DM100 tests. Glass properties evaluated included: viscosity, electrical conductivity, crystallinity, gross glass phase separation and the 7- day Product Consistency Test (ASTM-1285). Glass property limits were based upon the reference properties for the WTP HLW melter. However, the WTP crystallinity limit (< 1 vol% at 950oC) was relaxed slightly as a waste loading constraint for the crucible melts.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  12. 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.

  13. HIGH ALUMINUM HLW GLASSES FOR HANFORD'S WTP

    International Nuclear Information System (INIS)

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

    2009-01-01

    achievements of this program with emphasis on the recent enhancements in Al 2 O 3 loadings in HLW glass and its processing characteristics. Glass formulation development included crucible-scale preparation and characterization of glass samples to assess compliance with all melt processing and product quality requirements, followed by small-scale screening tests to estimate processing rates. These results were used to down-select formulations for subsequent engineering-scale melter testing. Finally, further testing was performed on the DM1200 vitrification system installed at VSL, which is a one-third scale (1.20 m 2 ) pilot melter for the WTP HLW melters and which is fitted with a fully prototypical off-gas treatment system. These tests employed glass formulations with high waste loadings and Al 2 O 3 contents of ∼25 wt%, which represents a near-doubling of the present WTP baseline maximum Al 2 O 3 loading. In addition, these formulations were processed successfully at glass production rates that exceeded the present requirements for WTP HLW vitrification by up to 88%. The higher aluminum loading in the HLW glass has an added benefit in that the aluminum leaching requirements in pretreatment are reduced, thus allowing less sodium addition in pretreatment, which in turn reduces the amount of LAW glass to be produced at the WTP. The impact of the results from this ORP program in reducing the overall cost and schedule for the Hanford waste treatment mission will be discussed

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  15. Development Of Glass Matrices For HLW Radioactive Wastes

    International Nuclear Information System (INIS)

    Jantzen, C.

    2010-01-01

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

  16. 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.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  18. SOURCE TERMS FOR HLW GLASS CANISTERS

    International Nuclear Information System (INIS)

    J.S. Tang

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Kim, Dong-Sang; Vienna, John D.

    2001-01-01

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

  20. 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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  3. Cooling and cracking of technical HLW glass products

    International Nuclear Information System (INIS)

    Kienzler, B.

    1989-01-01

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

  4. Expected behavior of HLW glass in storage

    International Nuclear Information System (INIS)

    McElroy, J.L.

    1975-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-04-01

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

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

    Science.gov (United States)

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

    2017-01-01

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

  7. Chemical compatibility of HLW borosilicate glasses with actinides

    International Nuclear Information System (INIS)

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

    1978-11-01

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

  8. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-15

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

  10. Baseline LAW Glass Formulation Testing

    International Nuclear Information System (INIS)

    Kruger, Albert A.; Mooers, Cavin; Bazemore, Gina; Pegg, Ian L.; Hight, Kenneth; Lai, Shan Tao; Buechele, Andrew; Rielley, Elizabeth; Gan, Hao; Muller, Isabelle S.; Cecil, Richard

    2013-01-01

    The major objective of the baseline glass formulation work was to develop and select glass formulations that are compliant with contractual and processing requirements for each of the LAW waste streams. Other objectives of the work included preparation and characterization of glasses with respect to the properties of interest, optimization of sulfate loading in the glasses, evaluation of ability to achieve waste loading limits, testing to demonstrate compatibility of glass melts with melter materials of construction, development of glass formulations to support ILAW qualification activities, and identification of glass formulation issues with respect to contract specifications and processing requirements

  11. 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.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-06

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Rutledge, V.J.; Maio, V.

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Kruger, A.A.; Jain, V.

    2009-01-01

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

  2. 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.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  5. Glass Ceramic Formulation Data Package

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  6. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-02-27

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

  8. Corrosion behaviour of the WAK-HLW glass

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  9. HLW vitrification in France industrial experience and glass quality

    International Nuclear Information System (INIS)

    Desvaux, J.L.; Delahaye, P.

    1994-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-15

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

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    International Nuclear Information System (INIS)

    Inagaki, Y.; Uehara, S.

    2004-01-01

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

  14. Formulation of heat absorbing glasses

    Directory of Open Access Journals (Sweden)

    Álvarez-Casariego, Pedro

    1996-06-01

    Full Text Available In the thermal exchanges between buildings and environment, glazing is an element of major importance, for it largely influences the so-called Solar Heat Gain and Thermal Losses. These parameters can be modified by applying different type of coatings onto glass surface or by adding colorant compounds during glass melting. The latter is a cheaper way to control the Solar Heat Gain. The knowledge of the laws governing the interaction between colorant compounds and solar radiation, allows us to define glass formulations achieving specific aesthetic requirements and solar energy absorption. In this paper two examples of application of the modelling of glass colorants spectral absorptance are presented. First is addressed to obtaining a glass with high luminous transmittance and low solar energy transmittance, and the other one to obtaining a glass with neutral colour appearance and minimized solar energy transmittance. Calculation formulas are defined together with photometric properties so-obtained. These type of glasses are particularly suitable to be used as building and automotive glazing, for they retain the mechanical characteristics and possibilities of transformation of standard glass.

    En los intercambios de energía entre un edificio y el medio exterior, el vidrio es el elemento de mayor importancia, por su influencia en la Ganancia de Calor Solar y en las Pérdidas Térmicas. Estos parámetros pueden ser modificados mediante el depósito de capas sobre el vidrio o mediante la adición de compuestos absorbentes de la radiación solar. Esta última vía es la más económica para controlar la Ganancia de Calor Solar. El conocimiento de las leyes que gobiernan la interacción de los diversos colorantes con la radiación solar, permite definir formulaciones de vidrios con características especificas de tipo estético y de absorción energética. En este trabajo se presentan dos ejemplos de aplicación de esta modelización de las

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-12

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  18. 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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  20. 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

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

    International Nuclear Information System (INIS)

    Luckscheiter, B.; Nesovic, M.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-04-01

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    1987-10-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  7. Technical Status Report: Preliminary Glass Formulation Report for INEEL HAW

    International Nuclear Information System (INIS)

    Peeler, D.; Reamer, I.; Vienna, J.; Crum, J.A.

    1998-03-01

    This study was performed by a team comprising experts in glass chemistry, glass technology, and statistics at both SRTC and PNNL. This joint effort combined the strengths of each discipline and site to quickly develop a glass formulation for specific INEEL HAW

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

    International Nuclear Information System (INIS)

    Van Iseghem, P.

    1997-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-05-01

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

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

    International Nuclear Information System (INIS)

    Neumann, W.

    1980-01-01

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

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

    International Nuclear Information System (INIS)

    Wang, L.

    1992-08-01

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

  12. 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.

  13. Mechanical properties of bioactive glass putty formulations

    NARCIS (Netherlands)

    van Gestel, N.A.P.; Geurts, J.A.P.; Hulsen, D.J.W.; Hofmann, S.; Ito, K.; van Rietbergen, B.; Arts, J.J.C.

    2016-01-01

    Introduction: Bioactive glass (BAG) has been studied widely and seems to be a very promising biomaterial in regeneration of large bone defects and osteomyelitis treatment, because of its bone bonding and antibacterial properties[1]-[5]. Its high stiffness could potentially also enable mechanical

  14. A new viscosity model for waste glass formulations

    International Nuclear Information System (INIS)

    Sadler, A.L.K.

    1996-01-01

    Waste glass formulation requires prediction, with reasonable accuracy, of properties over much wider ranges of composition than are typically encountered in any single industrial application. Melt viscosity is one such property whose behavior must be predicted in formulating new waste glasses. A model was developed for silicate glasses which relates the Arrhenius activation energy for flow to an open-quotes effectiveclose quotes measure of non-bridging oxygen content in the melt, NBO eff . The NBO eff parameter incorporates the differing effects of modifying cations on the depolymerization of the silicate network. The activation energy-composition relationship implied by the model is in accordance with experimental behavior. The model was validated against two different databases, with satisfactory results

  15. 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.

  16. Final Report. Baseline LAW Glass Formulation Testing, VSL-03R3460-1, Rev. 0

    Energy Technology Data Exchange (ETDEWEB)

    Muller, Isabelle S. [The Catholic University of America, Washington, DC (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); Buechele, Andrew [The Catholic University of America, Washington, DC (United States); Rielley, Elizabeth [The Catholic University of America, Washington, DC (United States); Bazemore, Gina [The Catholic University of America, Washington, DC (United States); Cecil, Richard [The Catholic University of America, Washington, DC (United States); Hight, Kenneth [The Catholic University of America, Washington, DC (United States); Mooers, Cavin [The Catholic University of America, Washington, DC (United States); Lai, Shan-Tao T. [The Catholic University of America, Washington, DC (United States); Kruger, Albert A. [The Catholic University of America, Washington, DC (United States)

    2015-06-18

    The major objective of the baseline glass formulation work was to develop and select glass formulations that are compliant with contractual and processing requirements for each of the LAW waste streams. Other objectives of the work included preparation and characterization of glasses with respect to the properties of interest, optimization of sulfate loading in the glasses, evaluation of ability to achieve waste loading limits, testing to demonstrate compatibility of glass melts with melter materials of construction, development of glass formulations to support ILAW qualification activities, and identification of glass formulation issues with respect to contract specifications and processing requirements.

  17. Glass formulation requirements for DWPF coupled operations using crystalline silicotitanates

    International Nuclear Information System (INIS)

    Harbour, J.R.; Andrews, M.K.

    1997-01-01

    The design basis DWPF flowsheet couples the vitrification of two waste streams: (1) a washed sludge and (2) a hydrolyzed sodium tetraphenylborate precipitate product, PHA. The PHA contains cesium-137 which had been precipitated from the tank supernate with sodium tetraphenylborate. Smaller amounts of strontium and plutonium adsorbed on sodium titanate are also present with the PHA feed. Currently, DWPF is running a sludge-only flowsheet while working towards solutions to the problems encountered with In Tank Precipitation (ITP). The sludge loading for the sludge-only flowsheet and for the anticipated coupled operations is 28 wt% on an oxide basis. For the coupled operation, it is essential to balance the treatment of the two waste streams such that no supernate remains after immobilization of all the sludge. An alternative to ITP and sodium titanate is the removal of Cs-137, Sr-90, and plutonium from the tank supernate by ion exchange using crystalline silicotitanate (CST). This material has been shown to effectively sorb these elements from the supernate. It is also known that CST sorbs plutonium. The loaded CST could then be immobilized with the sludge during vitrification. It has recently been demonstrated that CST loadings approaching 70 wt% for a CST-only glass can be achieved using a borosilicate glass formulation which can be processed by the DWPF melter. Initial efforts on coupled waste streams with simulated DWPF sludge show promise that a borosilicate glass formulation can incorporate both sludge and CST. This paper presents the bases for research efforts to develop a glass formulation which will incorporate sludge and CST at loadings appropriate for DWPF operation

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  19. Glass Formulation Development for INEEL Sodium-Bearing Waste

    International Nuclear Information System (INIS)

    Vienna, J.D.; Schweiger, M.J.; Smith, D.E.; Smith, H.D.; Crum, J.V.; Peeler, D.K.; Reamer, I.A.; Musick, C.A.; Tillotson, R.D.

    1999-01-01

    For about four decades, radioactive wastes have been collected and calcined from nuclear fuels reprocessing at the Idaho Nuclear Technology and Engineering Center (INTEC), formerly Idaho Chemical Processing Plant (ICPP). Over this time span, secondary radioactive wastes have also been collected and stored as liquid from decontamination, laboratory activities, and fuel-storage activities. These liquid wastes are collectively called sodium-bearing wastes (SBW). About 5.7 million liters of these wastes are temporarily stored in stainless steel tanks at the Idaho National Engineering and Environmental Laboratory (INEEL). Vitrification is being considered as an immobilization step for SBW with a number of treatment and disposal options. A systematic study was undertaken to develop a glass composition to demonstrate direct vitrification of INEEL's SBW. The objectives of this study were to show the feasibility of SBW vitrification, not a development of an optimum formulation. The waste composition is relatively high in sodium, aluminum, and sulfur. A specific composition and glass property restrictions, discussed in Section 2, were used as a basis for the development. Calculations based on first-order expansions of selected glass properties in composition and some general tenets of glass chemistry led to an additive (fit) composition (68.69 mass % SiO 2 , 14.26 mass% B 2 O 3 , 11.31 mass% Fe 2 O 3 , 3.08 mass% TiO 2 , and 2.67 mass % Li 2 O) that meets all property restrictions when melted with 35 mass % of SBW on an oxide basis, The glass was prepared using oxides, carbonates, and boric acid and tested to confirm the acceptability of its properties. Glass was then made using waste simulant at three facilities, and limited testing was performed to test and optimize processing-related properties and confirm results of glass property testing. The measured glass properties are given in Section 4. The viscosity at 1150 C, 5 Pa·s, is nearly ideal for waste-glass processing in

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

    International Nuclear Information System (INIS)

    Rolland, Severine

    2012-01-01

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

  1. PNL vitrification technology development project glass formulation strategy for LLW vitrification

    International Nuclear Information System (INIS)

    Kim, D.; Hrma, P.R.; Westsik, J.H. Jr.

    1996-03-01

    This Glass Formulation Strategy describes development approaches to optimize glass compositions for Hanford's low-level waste vitrification between now and the projected low-level waste facility start-up in 2005. The objectives of the glass formulation task are to develop optimized glass compositions with satisfactory long-term durability, acceptable processing characteristics, adequate flexibility to handle waste variations, maximize waste loading to practical limits, and to develop methodology to respond to further waste variations

  2. Glass Formulation and Fabrication Laboratory, Building 864, Hazards assessment document

    Energy Technology Data Exchange (ETDEWEB)

    Banda, Z.; Wood, C.L.

    1995-08-01

    The Department of Energy Order 5500.3A requires facility-specific hazards assessments be prepared, maintained, and used for emergency planning purposes. This hazards assessment document describes the chemical and radiological hazards associated with the Glass Formulation and Fabrication Laboratory, Building 864. The entire inventory was screened according to the potential airborne impact to onsite and offsite individuals. The air dispersion model, ALOHA, estimated pollutant concentrations downwind from the source of a release, taking into consideration the toxicological and physical characteristics of the release site, the atmospheric conditions, and the circumstances of the release. The greatest distances at which a postulated facility event will produce consequences exceeding the ERPG-2 threshold is 96 meters. The highest emergency classification is a Site Area Emergency. The Emergency Planning Zone is 100 meters.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-20

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

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

    International Nuclear Information System (INIS)

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

    2004-07-01

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

  5. 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.

  6. "Glass Formulation and Testing with TWRS LAW Simulants," Final Report to Duratek Inc. and BNFL Inc.

    Energy Technology Data Exchange (ETDEWEB)

    Muller, Isabelle S. [The Catholic University of America, Washington, DC (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); Buechele, Andrew C. [The Catholic University of America, Washington, DC (United States); Kim, C. [The Catholic University of America, Washington, DC (United States); Lai, Shan-Tao T. [The Catholic University of America, Washington, DC (United States); Del Rosario, G. [The Catholic University of America, Washington, DC (United States); Yan, Q. [The Catholic University of America, Washington, DC (United States); Kruger, Albert A. [The Catholic University of America, Washington, DC (United States)

    2015-06-22

    This report presents the results of glass formulation development with TWRS LAW simulants that was conducted at the Vitreous State Laboratory of The Catholic University of America during TWRS Phase I.

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

    International Nuclear Information System (INIS)

    Jacobs, R.A.

    1998-01-01

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

  8. 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.

  9. Development of an Alternative Glass Formulation for Vitrification of Excess Plutonium

    International Nuclear Information System (INIS)

    MARRA, JAMES

    2006-01-01

    The Department of Energy Office of Environmental Management (DOE/EM) plans to conduct the Plutonium Disposition Project at the Savannah River Site (SRS) to disposition excess weapons-usable plutonium. A plutonium glass waste form is a leading candidate for immobilization of the plutonium for subsequent disposition in a geologic repository. A reference glass composition (Lanthanide Borosilicate (La 2 O 3 -B 2 O 3 -SiO 2 (LaBS))-Frit B) was developed and testing with the LaBS Frit B composition is underway to provide data to support the Yucca Mountain License Application process. The objective of this task was to investigate alternative frit compositions and/or processing conditions that may improve the performance of the reference Frit B-LaBS glass in the repository. The current LaBS Frit B composition was used as the baseline for alternative glass formulation development efforts. A review of the literature and past high actinide concentration glass development efforts was conducted to formulate candidate compositions for testing. Glass science principles were also utilized to determine candidate frit components that may meet task objectives. Additionally, glass processing methods (e.g. slow cooling or induced heat treatment) were investigated as potential means to improve the glass durability and/or minimize fissile material and neutron absorber separation. Based on these analyses, a series of candidate surrogate glasses were fabricated and analyzed. One composition was then selected for fabrication with PuO 2 and subsequently analyzed. A phase equilibrium approach, developed from the assessment of previous high lanthanide glass formulations, was used to recommend modifications to the SRNL Frit B composition. A specific recommendation to increase Ln 2 O 3 content with concurrent reduction of Al 2 O 3 and SiO 2 content proved to be successful in improving the melting behavior and component solubility of the glass. This change moved the formulation from a

  10. 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.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-12-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  15. Final Report. LAW Glass Formulation to Support AP-101 Actual Waste Testing, VSL-03R3470-2, Rev. 0

    Energy Technology Data Exchange (ETDEWEB)

    Muller, I. S. [The Catholic University of America, Washington, DC (United States); Pegg, I. L. [The Catholic University of America, Washington, DC (United States); Rielley, Elizabeth [The Catholic University of America, Washington, DC (United States); Carranza, Isidro [The Catholic University of America, Washington, DC (United States); Hight, Kenneth [The Catholic University of America, Washington, DC (United States); Lai, Shan-Tao T. [The Catholic University of America, Washington, DC (United States); Mooers, Cavin [The Catholic University of America, Washington, DC (United States); Bazemore, Gina [The Catholic University of America, Washington, DC (United States); Cecil, Richard [The Catholic University of America, Washington, DC (United States); Kruger, Albert A. [The Catholic University of America, Washington, DC (United States)

    2015-06-22

    The main objective of the work was to develop and select a glass formulation for vitrification testing of the actual waste sample of LAW AP-101 at Battelle - Pacific Northwest Division (PNWD). Other objectives of the work included preparation and characterization of glasses to demonstrate compliance with contract and processing requirements, evaluation of the ability to achieve waste loading requirements, testing to demonstrate compatibility of the glass melts with melter materials of construction, comparison of the properties of simulant and actual waste glasses, and identification of glass formulation issues with respect to contract specifications and processing requirements.

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

    Science.gov (United States)

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

    2017-09-01

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

  17. Evaluation of the potential of waste fondant glass in formulations of ceramic pasta

    International Nuclear Information System (INIS)

    Soares Filho, J.E.; Santos, L.L. dos; Nascimento, R.M. do; Feitosa, A.O.; Dutra, R.P.S.

    2014-01-01

    An increasing amount of waste generated and deposited on the environment, many unspecified decomposition with time, as is the case of the glass. Thinking about it, the purpose of this study is to evaluate the power of the flux residue on glass formulations porcelains, as a flux to feldspar replacement. This study was performed in comparison with a standard formulation. The raw materials were characterized in the diffraction X-ray fluorescence and X-ray thermal differential analysis, and determination of the technological properties of water absorption, linear contraction, ignition loss, apparent porosity and apparent specific gravity in the formulation standard and replacement of feldspar in different percentages of waste and processing conditions. Specimens of the formulations were subjected to assay of three points. Results indicate that the residue glass has the potential of being used as a flux material in the composition of the ceramic body reduces the apparent porosity and according to the technology of water absorption property. The ceramic mass standard was classified as semi-stoneware, the BIIa group, and after the addition of the residue in any of the three percentages evaluated was classified as sandstone, belonging to the group BIb.(author)

  18. Formulation of a candidate glass for use as an acceptance test standard material

    International Nuclear Information System (INIS)

    Ebert, W.L.; Strachan, D.M.; Wolf, S.F.

    1998-04-01

    In this report, the authors discuss the formulation of a glass that will be used in a laboratory testing program designed to measure the precision of test methods identified in the privatization contracts for the immobilization of Hanford low-activity wastes. Tests will be conducted with that glass to measure the reproducibility of tests and analyses that must be performed by glass producers as a part of the product acceptance procedure. Test results will be used to determine if the contractually required tests and analyses are adequate for evaluating the acceptability of likely immobilized low-activity waste (ILAW) products. They will also be used to evaluate if the glass designed for use in these tests can be used as an analytical standard test material for verifying results reported by vendors for tests withg ILAW products. The results of those tests and analyses will be presented in a separate report. The purpose of this report is to document the strategy used to formulate the glass to be used in the testing program. The low-activity waste reference glass LRM that will be used in the testing program was formulated to be compositionally similar to ILAW products to be made with wastes from Hanford. Since the ILAW product compositions have not been disclosed by the vendors participating in the Hanford privatization project, the composition of LRM was formulated based on simulated Hanford waste stream and amounts of added glass forming chemicals typical for vitrified waste forms. The major components are 54 mass % SiO 2 , 20 mass % Na 2 O, 10 mass % Al 2 O 3 , 8 mass % B 2 O 3 , and 1.5 mass % K 2 O. Small amounts of other chemicals not present in Hanford wastes were also included in the glass, since they may be included as chemical additives in ILAW products. This was done so that the use of LRM as a composition standard could be evaluated. Radionuclides were not included in LRM because a nonradioactive material was desired

  19. 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

  20. Technetium Chemistry in HLW

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  3. Technical Status Report: Preliminary Glass Formulation Report for INEEL HAW. Revision 1

    International Nuclear Information System (INIS)

    Peeler, D.; Reamer, I.; Vienna, J.; Crum, J.A.

    1998-03-01

    Preliminary glass formulation work has been initiated at Pacific Northwest National Laboratory (PNNL) and the Savannah River Technology Center (SRTC) to support immobilization efforts of Idaho National Engineering and Environmental Laboratory (INEEL) high activity waste (HAW). Based on current pretreatment flow sheet assumptions, several glasses were fabricated and tested using an average 'All Blend' waste stream composition which is dominated by the presence of ZrO 2 (i.e., approximately 80 wt percent). The results of this initial work show that immobilization via vitrification is a viable option for a specific INEEL HAW waste stream. Waste loadings of at least 19 wt percent can be achieved for the 'All Blend' stream while maintaining targeted processing and product performance criteria. This waste loading translates into a ZrO 2 content in excess of 15 wt percent in the final glass waste form. Frits developed for this work are based in the alkali borosilicate system. Although the results indicate that vitrification can be used to immobilize the 'All Blend' waste stream, the glass compositions are by no means optimized

  4. Industrial scale-plant for HLW partitioning in Russia

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  5. Silicate Based Glass Formulations for Immobilization of U.S. Defense Wastes Using Cold Crucible Induction Melters

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Gary L.; Kim, Dong-Sang; Schweiger, Michael J.; Marra, James C.; Lang, Jesse B.; Crum, Jarrod V.; Crawford, Charles L.; Vienna, John D.

    2014-05-22

    The cold crucible induction melter (CCIM) is an alternative technology to the currently deployed liquid-fed, ceramic-lined, Joule-heated melter for immobilizing of U.S. tank waste generated from defense related reprocessing. In order to accurately evaluate the potential benefits of deploying a CCIM, glasses must be developed specifically for that melting technology. Related glass formulation efforts have been conducted since the 1990s including a recent study that is first documented in this report. The purpose of this report is to summarize the silicate base glass formulation efforts for CCIM testing of U.S. tank wastes. Summaries of phosphate based glass formulation and phosphate and silicate based CCIM demonstration tests are reported separately (Day and Ray 2013 and Marra 2013, respectively). Combined these three reports summarize the current state of knowledge related to waste form development and process testing of CCIM technology for U.S. tank wastes.

  6. Source term measurements on vitrified HLW

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  7. Letter Report. Proposed Approach for Development of LAW Glass Formulation Correlation, VSL-04L4460-1, Rev. 2

    Energy Technology Data Exchange (ETDEWEB)

    Muller, Isabelle S. [The Catholic University of America, Washington, DC (United States); Diener, Glenn [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); Kruger, Albert A. [The Catholic University of America, Washington, DC (United States)

    2015-06-18

    The main objective of the work is to develop a correlation that employs waste composition information to determine the appropriate waste loading, glass composition, and amounts and types of glass formers. In addressing this objective emphasis has been placed on those compositions that have been validated in DM100 and LAW Pilot Melter testing. This is particularly important in view of the essential role that potential for sulfate phase separation in the melter plays in glass formulation selection. A further objective of this work is to select and test glass compositions in order to augment the existing data set and to test the predictions from the correlation. It should be noted that the intent of the correlation is to provide practical, robust glass formulations that exceed all of the contract and processability requirements; it is not intended to provide the "maximum achievable" waste loading such that at least one of those properties is at its respective limit.

  8. 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

  9. Safety of HLW shipments

    International Nuclear Information System (INIS)

    1998-01-01

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

  10. Glass formulation requirements for Hanford coupled operations using crystalline silicotitanates (CST)

    International Nuclear Information System (INIS)

    Andrews, M.K.; Harbour, J.R.

    1997-01-01

    The U.S. Department of Energy (DOE) through the Richland Operations Office has requested proposals from the private sector for the treatment of waste from the Hanford Waste Tanks. Phase I of this privatization initiative may include a demonstration for treatment and immobilization of both low activity and high-level waste. If the demonstration includes high-level waste, then the Cs-137 waste stream most likely will be combined with the high-level waste sludge to produce a coupled feed for immobilization (most likely vitrification using a borosilicate glass). It appears that pretreatment will involve the removal of cesium (and perhaps strontium and some transuranic radionuclides) from the supernate using an ion exchange material such as crystalline silicotitanate (CST). The ion exchange sorbent (or the eluted Cs-137) can then be combined with the sludge and vitrified in a coupled operation similar to the DWPF process. Alternatively, the cesium-loaded ion exchange sorbent can be vitrified directly to produce a separate glass waste form. SRTC has been involved in an Office of Science and Technology (EM-50) funded project to determine if Cs-137 loaded CST can be successfully incorporated into glass at significant levels. 1 For a waste form which would include only Cs-137 loaded CST, concentrations up to 60 wt% of CST in glass have been achieved. 2 The glass produced from this demonstration is both processable and durable. This CST-only waste form could be used at Hanford if the cesium-loaded CST is vitrified in a separate melter. For coupled feed operations, the CST would be mixed with high-level radioactive sludge from the Hanford tanks. This report provides the basis and the path forward for SRTC's efforts at developing a glass frit formulation which will incorporate both Hanford sludge and cesium-loaded CST for a coupled flowsheet. The goal of this work is to demonstrate the feasibility of vitrification as a method for immobilization of coupled feed (specifically

  11. 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.

  12. 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

  13. The basic corrosion mechanisms of HLW glasses

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  14. The Impact Of The Mcu Life Extension Solvent On Dwpf Glass Formulation Efforts

    International Nuclear Information System (INIS)

    Peeler, D.; Edwards, T.

    2011-01-01

    As a part of the Actinide Removal Process (ARP)/Modular Caustic Side Solvent Extraction Unit (MCU) Life Extension Project, a next generation solvent (NG-CSSX), a new strip acid, and modified monosodium titanate (mMST) will be deployed. The strip acid will be changed from dilute nitric acid to dilute boric acid (0.01 M). Because of these changes, experimental testing with the next generation solvent and mMST is required to determine the impact of these changes in 512-S operations as well as Chemical Process Cell (CPC), Defense Waste Processing Facility (DWPF) glass formulation activities, and melter operations at DWPF. To support programmatic objectives, the downstream impacts of the boric acid strip effluent (SE) to the glass formulation activities and melter operations are considered in this study. More specifically, the impacts of boric acid additions to the projected SB7b operating windows, potential impacts to frit production temperatures, and the potential impact of boron volatility are evaluated. Although various boric acid molarities have been reported and discussed, the baseline flowsheet used to support this assessment was 0.01M boric acid. The results of the paper study assessment indicate that Frit 418 and Frit 418-7D are robust to the implementation of the 0.01M boric acid SE into the SB7b flowsheet (sludge-only or ARP-added). More specifically, the projected operating windows for the nominal SB7b projections remain essentially constant (i.e., 25-43 or 25-44% waste loading (WL)) regardless of the flowsheet options (sludge-only, ARP added, and/or the presence of the new SE). These results indicate that even if SE is not transferred to the Sludge Receipt and Adjustment Tank (SRAT), there would be no need to add boric acid (from a trim tank) to compositionally compensate for the absence of the boric acid SE in either a sludge-only or ARP-added SB7b flowsheet. With respect to boron volatility, the Measurement Acceptability Region (MAR) assessments also

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  16. 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.

  17. 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.

  18. Safety assessment of HLW geological disposal system

    International Nuclear Information System (INIS)

    Naito, Morimasa

    2006-01-01

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

  19. The Impact of the Proposed delta Gp Limits on Glass Formulation Efforts: Part II. Experimental Results

    International Nuclear Information System (INIS)

    PEELER, DAVID

    2004-01-01

    The Savannah River National Laboratory (SRNL) has initiated studies to assess alternative durability options that may provide access to compositional regions of interest in support of the accelerated cleanup mission at the Defense Waste Processing Facility (DWPF). One of the options being pursued is the redefinition of the durability model acceptability limits. Some of the conservative steps used in establishing the current limits without compromising the high confidence required for meeting the specification on the waste form quality were identified and eliminated. The results led to a set of three new Property Acceptability Region (PAR) values for the preliminary glass dissolution estimator that has the potential to allow access to compositional regions of interest to improve melt rate or waste loading. Although these limits are available for implementation, there is no driving force to do so with the current sludge batch (i.e., the current Frit 418 - Sludge Batch 3 (SB3) system is TL limited). The objectives of this task were to investigate (and generate) the incentive of applying the proposed durability limits in the Product Composition Control System (PCCS) from a glass formulation perspective. Glass compositions were identified or developed to transition into and through the region of GP acceptability as defined by the current and proposed durability limits. The progression through the newly defined acceptability region was accomplished by increasing the total alkali in the glass via higher alkali frits and/or waste loading (WL). The focus of this report is on the measured durability response as it compares to model predictions to assess the applicability and/or potential conservatism of the various limits or durability approaches. The normalized boron release values (NL [B] g/L) for the study glasses ranged from approximately 1.0 g/L to 2.0 g/L. The Product Consistency Test (PCT) responses provide evidence that implementation of the proposed GP limits will

  20. Final Report - Enhanced LAW Glass Formulation Testing, VSL-07R1130-1, Rev. 0, dated 10/05/07

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Pegg, I. L.; Matlack, K. S.; Joseph, I.; Muller, I. S.; Gong, W.

    2013-11-13

    The principal objective of this work was to extend the glass formulation methodology developed in the earlier work [2, 5, 6] for Envelope A, B and C waste compositions for development of compliant glass compositions targeting five high sodium-sulfur waste loading regions. This was accomplished through a combination of crucible-scale tests, and tests on the DM10 melter system. The DM10 was used for several previous tests on LAW compositions to determine the maximum feed sulfur concentrations that can be processed without forming secondary sulfate phases on the surface of the melt pool. This melter is the most efficient melter platform for screening glass compositions over a wide range of sulfate concentrations and therefore was selected for the present tests. The tests were conducted to provide information on melter processing characteristics and off-gas data, including sulfur incorporation and partitioning. As described above, the main objective was to identify the limits of waste loading in compliant glass formulations spanning the range of expected Na{sub 2}O and SO{sub 3} concentrations in the LAW glasses.

  1. 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.

  2. 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.

  3. Formulation and synthesis by melting process of titanate enriched glass-ceramics and ceramics

    International Nuclear Information System (INIS)

    Advocat, T.; Fillet, C.; Lacombe, J.; Bonnetier, A.; McGlinn, P.

    1999-01-01

    The main objective of this work is to provide containment for the separated radionuclides in stable oxide phases with proven resistance to leaching and irradiation damage and in consequence to obtain a glass ceramic or a ceramic material using a vitrification process. Sphene glass ceramic, zirconolite glass ceramic and zirconolite enriched ceramic have been fabricated and characterized by XRD, SEM/EDX and DTA

  4. HLW Tank Space Management, Final Report

    International Nuclear Information System (INIS)

    Sessions, J.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  6. 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

  7. Characterization of cutting soda-lime glass sludge for the formulation of red ceramic products

    International Nuclear Information System (INIS)

    Filogonio, P.H.C.; Reis, A.S.; Louzada, D.M.; Della, V.P.

    2014-01-01

    Considering previous works that have demonstrated the feasibility of soda-lime glass incorporation into red ceramics, this paper aims to determine the potential for incorporation of cutting soda-lime glass sludge in red ceramic manufacturing. Therefore, the waste was characterized by X-ray fluorescence, X-ray diffraction, particle size distribution and thermal behavior. The results confirm the chemical and mineralogical similarity between waste and soda-lime glass. Because of this similarity, it is concluded that the soda-lime glass waste has the capability to be used in the manufacturing of red ceramics. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-09-26

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    International Nuclear Information System (INIS)

    Herman, C

    2007-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-02-01

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

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

    International Nuclear Information System (INIS)

    CAROL, JANTZEN

    2005-01-01

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

  13. Silicate glasses

    International Nuclear Information System (INIS)

    Lutze, W.

    1988-01-01

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

  14. COMPILATION AND MANAGEMENT OF ORP GLASS FORMULATION DATABASE, VSL-12R2470-1 REV 0

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Pasieka, Holly K. [The Catholic University of America, Washington, DC (United States); Muller, Isabelle [The Catholic University of America, Washington, DC (United States); Gilbo, Konstantin [The Catholic University of America, Washington, DC (United States); Perez-Cardenas, Fernando [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); Kot, Wing K. [The Catholic University of America, Washington, DC (United States)

    2012-12-13

    The present report describes the first steps in the development of a glass property-composition database for WTP LAW and HL W glasses that includes all of the data that were used in the development of the WTP baseline models and all of the data collected subsequently as part of WTP enhancement studies perfonned for ORP. The data were reviewed to identifY some of the more significant gaps in the composition space that will need to be filled to support waste processing at Hanford. The WTP baseline models have been evaluated against the new data in terms of range of validity and prediction perfonnance.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-13

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  17. 12 Flasktransport of vitrified High Level Waste (HLW)

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  18. 12 Flasktransport of vitrified High Level Waste (HLW)

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  1. Risk assessment of amorphous silicon dioxide nanoparticles in a glass cleaner formulation

    Science.gov (United States)

    Scheel, Julia; Karsten, Stefan; Stelter, Norbert; Wind, Thorsten

    2013-01-01

    Since nanomaterials are a heterogeneous group of substances used in various applications, risk assessment needs to be done on a case-by-case basis. Here the authors assess the risk (hazard and exposure) of a glass cleaner with synthetic amorphous silicon dioxide (SAS) nanoparticles during production and consumer use (spray application). As the colloidal material used is similar to previously investigated SAS, the hazard profile was considered to be comparable. Overall, SAS has a low toxicity. Worker exposure was analysed to be well controlled. The particle size distribution indicated that the aerosol droplets were in a size range not expected to reach the alveoli. Predictive modelling was used to approximate external exposure concentrations. Consumer and environmental exposure were estimated conservatively and were not of concern. It was concluded based on the available weight-of-evidence that the production and application of the glass cleaner is safe for humans and the environment under intended use conditions. PMID:22548260

  2. Glass formulation development and offgas analysis of microwave melter powder samples

    International Nuclear Information System (INIS)

    Semones, G.B.; Hoffman, C.R.; Phillips, J.A.

    1994-04-01

    Production of nuclear materials for defense applications has resulted in the accumulation of vast amounts of nuclear waste. This contaminated waste is in a variety of forms that require subsequent reprocessing to isolate and encapsulate the nuclear (e.g., uranium, plutonium, strontium, cesium, and americium) and toxic (e.g., lead, chromium, and cadmium) constituents. The encapsulating material must possess good chemical and mechanical durability to resist leaching of the nuclear and toxic constituents into the environment during permanent storage at a waste repository. Glass is an ideal encapsulating material because its open structure allows the introduction of different waste forms and the final vitreous product possesses a high degree of chemical stability. Microwave heating and melting is a relatively new advancement in glass processing which uses microwave radiation to heat the glass formers to adequate temperatures for sintering or melting. An advantage to this technique is that it enables more rapid heating than traditional heating mechanisms. This decrease in cycle time may help to limit exposure to workers encapsulating radioactive and/or toxic waste

  3. Nepheline Crystallization in Nuclear Waste Glasses: Progress toward acceptance of high-alumina formulations

    International Nuclear Information System (INIS)

    McCloy, John S.; Schweiger, Michael J.; Rodriguez, Carmen P.; Vienna, John D.

    2011-01-01

    We have critically compiled and analyzed historical data for investigating the quantity of nepheline (NaAlSiO4) precipitation as a function of composition in simulated nuclear waste glasses. To understand composition we used two primary methods: (1) investigating the Al2O3-SiO2-Na2O ternary with filtering for different B2O3 levels and (2) creating a quadrant system consisting of compositions reduced to two metric numbers. These metrics are (1) the nepheline discriminator (ND) which depends only on the SiO2 content by weight normalized to the total weight of the Al2O3-SiO2-Na2O sub-mixture and (2) the optical basicity (OB) which contains contributions from all constituents in the glass. Nepheline precipitation is expected to be suppressed at high SiO2 levels (ND >0.62) or at low basicities (OB 5 wt% B2O3. The OB concept can help further refine regions of nepheline-free glass formation.

  4. Electrical conductivity and modulus formulation in zinc modified bismuth boro-tellurite glasses

    Science.gov (United States)

    Dhankhar, Sunil; Kundu, R. S.; Dult, Meenakshi; Murugavel, S.; Punia, R.; Kishore, N.

    2016-09-01

    The ac conductivity of zinc modified tellurium based quaternary glasses having composition 60 TeO2-10 B2O3-(30 - x) Bi2O3-x ZnO; x = 10, 15, 20, 25 and 30 has been investigated in the frequency range 10-1-105 Hz and in temperature range 483-593 K. Frequency and temperature dependent ac conductivity found to obey Jonscher power law modified by Almond-West. DC conductivity, crossover frequency and frequency exponent have been estimated from the fitting of the experimental data of conductivity with Jonscher power law modified by Almond-West. The ac conductivity and its frequency exponent have been analyzed by various theoretical models. In presently studied glasses ac conduction takes place via tunneling of overlapping large polaron tunneling. Activation energy is found to be increased with increase in zinc content and dc conduction takes place via variable range hopping proposed by Mott with some modification suggested by Punia et al. The value of the stretched exponent ( β) obtained by fitting of M^' ' }} reveals the presence of non-Debye type relaxation. Scaling spectra of ac conductivity and electric modulus collapse into a single master curve for all compositions and temperatures, reveals the presence of composition and temperature independent conduction and relaxation process in these glasses. Activation energy of conduction ( W) and electric modulus ( E R ) are nearly equal, indicating that polaron have to overcome the same energy barrier during conduction as well as relaxation processes.

  5. 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.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  7. 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.

  8. 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....

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

    International Nuclear Information System (INIS)

    Nagano, Koji

    2003-01-01

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

  10. 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.

  11. A dose of HLW reality

    International Nuclear Information System (INIS)

    Payne, J.

    1993-01-01

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

  12. Vitrification of HLW in cold crucible melter

    International Nuclear Information System (INIS)

    Bordier, G.

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Magnin, M.

    2009-09-01

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

  14. Evaluation of the potential of waste fondant glass in formulations of ceramic pasta; Avaliacao do potencial fundente do residuo de vidro em formulacoes de massas ceramicas

    Energy Technology Data Exchange (ETDEWEB)

    Soares Filho, J.E.; Santos, L.L. dos; Nascimento, R.M. do, E-mail: elsinhofilho@gmail.com [Universidade Federal do Rio Grande do Norte (PPGCEM/UFRN), RN (Brazil). Programa de Pos-Graduacao em Ciencia e Engenharia de Materiais; Feitosa, A.O.; Dutra, R.P.S. [Universidade Federal da Paraiba (PPCEM/UFPB), PB (Brazil). Programa de Pos-Graduacao em Ciencia e Engenharia de Materiais

    2014-07-01

    An increasing amount of waste generated and deposited on the environment, many unspecified decomposition with time, as is the case of the glass. Thinking about it, the purpose of this study is to evaluate the power of the flux residue on glass formulations porcelains, as a flux to feldspar replacement. This study was performed in comparison with a standard formulation. The raw materials were characterized in the diffraction X-ray fluorescence and X-ray thermal differential analysis, and determination of the technological properties of water absorption, linear contraction, ignition loss, apparent porosity and apparent specific gravity in the formulation standard and replacement of feldspar in different percentages of waste and processing conditions. Specimens of the formulations were subjected to assay of three points. Results indicate that the residue glass has the potential of being used as a flux material in the composition of the ceramic body reduces the apparent porosity and according to the technology of water absorption property. The ceramic mass standard was classified as semi-stoneware, the BIIa group, and after the addition of the residue in any of the three percentages evaluated was classified as sandstone, belonging to the group BIb.(author)

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K.; Edwards, T.

    2010-11-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-12-29

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

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

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

    International Nuclear Information System (INIS)

    Shah, J.G.; Yalmali, V.S.; Tawde, Manisha; Mishra, R.

    2006-01-01

    The need of nuclear power as an energy source requires the solution of many problems. One of the most important is fixation of high level radioactive waste (HLW) in suitable borosilicate glass formulation. The major issue with this process is maximum waste loading in the final vitrified product without compromising on long term product characteristics. The electrical resistivity measurement at high temperature could not be measured with good precision using standard parallel plate electrode configuration due to error in cell constant measurement. Hence a high accuracy, calibration free technique consisting of co-axial electrodes was employed

  2. 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.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  4. Molecular glasses for nuclear waste encapsulation

    International Nuclear Information System (INIS)

    Ropp, R.C.

    1982-01-01

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

  5. Counter current decantation washing of HLW sludge

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  6. Strategic management of HLW repository projects

    International Nuclear Information System (INIS)

    Bartlett, J.W.

    1984-01-01

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

  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)

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

    2009-01-01

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

  9. 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)

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

    International Nuclear Information System (INIS)

    Weber, W.J.

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Kitamoto, Asashi; Mulyanto.

    1995-01-01

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

  12. HLW Long-term Management Technology Development

    International Nuclear Information System (INIS)

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

    2010-02-01

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

  13. Wastes vitrification by plasma torch: study of a glass formulation compatible with a wide range of B wastes

    International Nuclear Information System (INIS)

    Poitou, S.; Richaud, D.; Fiquet, O.; Gramondi, P.; Massit, H.

    2001-01-01

    Within the context of radioactive waste management, CEA has equipped itself with a 'PLASMARC' device. The central element of this device is a plasma torch treatment furnace. It has been implemented and validated for the vitrification of low level radioactive wastes. Meanwhile, the plasma torch presents potentially interests for immobilizing under an inert form in vitreous matrices, B wastes which are generally divided and of complex chemical composition. The application of this process to this type of wastes has been studied here. The obtained results show that with the plasma torch it is possible to make glasses with a high amount of silicon and aluminium oxide and which are adapted to the treatment / packaging of the B wastes. (O.M.)

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

    International Nuclear Information System (INIS)

    Steininger, W.

    2006-01-01

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

  15. Characterization of dynamics in complex lyophilized formulations: I. Comparison of relaxation times measured by isothermal calorimetry with data estimated from the width of the glass transition temperature region.

    Science.gov (United States)

    Chieng, Norman; Mizuno, Masayasu; Pikal, Michael

    2013-10-01

    The purposes of this study are to characterize the relaxation dynamics in complex freeze dried formulations and to investigate the quantitative relationship between the structural relaxation time as measured by thermal activity monitor (TAM) and that estimated from the width of the glass transition temperature (ΔT(g)). The latter method has advantages over TAM because it is simple and quick. As part of this objective, we evaluate the accuracy in estimating relaxation time data at higher temperatures (50 °C and 60 °C) from TAM data at lower temperature (40 °C) and glass transition region width (ΔT(g)) data obtained by differential scanning calorimetry. Formulations studied here were hydroxyethyl starch (HES)-disaccharide, HES-polyol, and HES-disaccharide-polyol at various ratios. We also re-examine, using TAM derived relaxation times, the correlation between protein stability (human growth hormone, hGH) and relaxation times explored in a previous report, which employed relaxation time data obtained from ΔT(g). Results show that most of the freeze dried formulations exist in single amorphous phase, and structural relaxation times were successfully measured for these systems. We find a reasonably good correlation between TAM measured relaxation times and corresponding data obtained from estimates based on ΔT(g), but the agreement is only qualitative. The comparison plot showed that TAM data are directly proportional to the 1/3 power of ΔT(g) data, after correcting for an offset. Nevertheless, the correlation between hGH stability and relaxation time remained qualitatively the same as found with using ΔT(g) derived relaxation data, and it was found that the modest extrapolation of TAM data to higher temperatures using ΔT(g) method and TAM data at 40 °C resulted in quantitative agreement with TAM measurements made at 50 °C and 60 °C, provided the TAM experiment temperature, is well below the Tg of the sample. Copyright © 2013 Elsevier B.V. All rights

  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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  17. 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

  18. The borosilicate glass for 'PAMELA'

    International Nuclear Information System (INIS)

    Schiewer, E.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-09-01

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

  20. SNF/HLW Transfer System Description Document

    International Nuclear Information System (INIS)

    W. Holt

    2005-01-01

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

  1. Stress analysis of HLW containers. Compas project

    International Nuclear Information System (INIS)

    1989-01-01

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

  2. HLW Flexible jumper materials compatibility evaluation

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-13

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

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  5. R and D on HLW Partitioning in Russia

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    H. Marr

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  8. 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.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Ninkovic, M. M.; Raicevic, J.

    1995-01-01

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

  11. Active geothermal systems as natural analogs of HLW repositories

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Buelow, S.

    1997-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

    and Environmental Laboratory (INEEL), is providing preliminary data on the CCIM technology using a small laboratory unit at the Khlopin Radium Institute in St. Petersburg Russia with INEEL Sodium Bearing Waste surrogate. The task includes both the baseline borosilicate glass and a new iron-phosphate glass developed at the University of Missouri-Rolla, which may offer significant advantages in compatibility with greater concentrations of highly refractory oxides. This project is integrating two disparate advances to develop a system with strong potential for benefit to the Department of Energy. Collaborative development of basic physical parameter data on the CCIM using promising glass formulations is being conducted by University of Missouri - Rolla, Russian and American researchers

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

    International Nuclear Information System (INIS)

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

    2011-01-01

    This report provides data, analyses, and conclusions from a series of tests that were conducted at the Vitreous State Laboratory of The Catholic of America (VSL) to determine the processing rates that are achievable with AZ-101 HLW simulants and corresponding melter feeds on a DuraMelter 100 (DM100) vitrification system. One of the most critical pieces of information in determining the required size of the RPP-WTP HLW melter is the specific glass production rate in terms of the mass of glass that can be produced per unit area of melt surface per unit time. The specific glass production rate together with the waste loading (essentially, the ratio of waste-in to glass-out, which is determined from glass formulation activities) determines the melt area that is needed to achieve a given waste processing rate with due allowance for system availability. Tests conducted during Part B1 (VSL-00R2590-2) on the DM1000 vitrification system installed at the Vitreous State Laboratory of The Catholic University of America showed that, without the use of bubblers, glass production rates with AZ-101 and C-106/AY-102 simulants were significantly lower than the Project design basis rate of 0.4 MT/m 2 /d. Conversely, three-fold increases over the design basis rate were demonstrated with the use of bubblers. Furthermore, an un-bubbled control test using a replica of the melter feed used in cold commissioning tests at West Valley reproduced the rates that were observed with that feed on the WVDP production melter. More recent tests conducted on the DM1200 system, which more closely represents the present RPP-WTP design, are in general agreement with these earlier results. Screening tests conducted on the DM10 system have provided good indications of the larger-scale processing rates with bubblers (for both HL W and LAW feeds) but significantly overestimated the DM1000 un-bubbled rate observed for C-106/AY-102 melter feeds. This behavior is believed to be a consequence of the role of

  15. 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

  16. Mechanical properties of nuclear waste glasses

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  17. Glass leaching performance

    International Nuclear Information System (INIS)

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

    1983-05-01

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

  18. Prediction of the Formulation Dependence of the Glass Transition Temperature for Amine-Epoxy Copolymers Using a Quantitative Structure-Property Relationship Based on the AM1 Method

    National Research Council Canada - National Science Library

    Morrill, Jason

    2004-01-01

    A designer Quantitative Structure-Property Relationsbip (QSPR) based upon molecular properties calculated using the AM1 semi-empirical quantum mechanical metbod was developed to predict the glass transition temperature (Tg...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-08-01

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

  20. 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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    2000-01-01

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

  3. Solidification of HLLW by glass-ceramic process

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  4. GLASS FORMULATION TESTING TO INCREASE SULFATE INCORPORATION - Final Report VSL-04R4960-1, Rev 0, 2/28/05, Vitreous State Laboratory, The Catholic University of American, Washington, D.C.

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; MATLACK KS

    2012-02-07

    by Optima Chemicals according to VSL specifications. Sufficient feed was prepared to produce over nineteen hundred kilograms of glass during melter tests. The nominal reductant concentration (stoichiometric ratio of 0.5 {approx} 1 mole sucrose per 16 mole NOx or 3 mole carbon per 4 mole NOx) was maintained in all the tests by the addition of sugar at VSL. The DM 10 was used to screen the optimized glass formulation with two alternative aluminum sources (kyanite and zeolite) over a wide range of target sulfur concentrations. Subsequently, based on the DM10 results, nine 12- to 34-hour DM100 tests were conducted; six with kyanite as the aluminum additive at glass sulfur concentrations ranging from 0.75 to 1.5 wt.% SO{sub 3}, and the other three with zeolite as the aluminum additive at glass sulfur concentrations ranging from 0.75 to 1.5 wt. % SO{sub 3}. The DM 100-WV melter was used in order to provide a direct comparison with the LAW tests previously conducted on the same melter. Key operating parameters such as glass temperature and production rate were held constant to investigate the sulfur incorporation into the glass and the effects of varying the aluminum additive source. The bubbling rate was adjusted to achieve a production rate of 2000 kg/m{sup 2}/day with a near-complete cold cap (90-100% of melt surface covered with feed). Quantitative measurements of glass production rates, melter operating conditions (temperatures, pressures, power, flows, etc.), and off-gas characteristics (NOx, SO{sub 2}, CO, particulate load and composition, and acid gases) were made for each test. Glass samples taken from the glass pool and the discharge chamber were inspected throughout testing to determine the limit of salt-free operation in the melter.

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

    International Nuclear Information System (INIS)

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Molecke, M.A.

    1977-01-01

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

  7. TWRS HLW interim storage facility search and evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Calmus, R.B., Westinghouse Hanford

    1996-05-16

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

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

    International Nuclear Information System (INIS)

    Bartlett, J.W.

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Vinjamuri, K.

    1990-01-01

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

  10. Final Report - Glass Formulation Testing to Increase Sulfate Volatilization from Melter, VSL-04R4970-1, Rev. 0, dated 2/24/05

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Matlack, K. A.; Pegg, I. L.; Gong, W.

    2013-11-13

    The principal objectives of the DM100 and DM10 tests were to determine the impact of four different organics and one inorganic feed additive on sulfate volatilization and to determine the sulfur partitioning between the glass and the off-gas system. The tests provided information on melter processing characteristics and off-gas data including sulfur incorporation and partitioning. A series of DM10 and DM100 melter tests were conducted using a LAW Envelope A feed. The testing was divided into three parts. The first part involved a series of DM10 melter tests with four different organic feed additives: sugar, polyethylene glycol (PEG), starch, and urea. The second part involved two confirmatory 50-hour melter tests on the DM100 using the best combination of reductants and conditions based on the DM10 results. The third part was performed on the DM100 with feeds containing vanadium oxide (V{sub 2}O{sub 5}) as an inorganic additive to increase sulfur partitioning to the off-gas. Although vanadium oxide is not a reductant, previous testing has shown that vanadium shows promise for partitioning sulfur to the melter exhaust, presumably through its known catalytic effect on the SO{sub 2}/SO{sub 3} reaction. Crucible-scale tests were conducted prior to the melter tests to confirm that the glasses and feeds would be processable in the melter and that the glasses would meet the waste form (ILAW) performance requirements. Thus, the major objectives of these tests were to: Perform screening tests on the DM10 followed by tests on the DM100-WV system using a LAW -Envelope A feed with four organic additives to assess their impact on sulfur volatilization. Perform tests on the DM100-WV system using a LAW -Envelope A feed containing vanadium oxide to assess its impact on sulfur volatilization. Determine feed processability and product quality with the above additives. Collect melter emissions data to determine the effect of additives on sulfur partitioning and melter emissions

  11. R and D programme for HLW disposal in Japan

    International Nuclear Information System (INIS)

    Tsuboya, Takao

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Zheng Hualing; Chen Shi; Sun Donghui

    2007-01-01

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

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

    International Nuclear Information System (INIS)

    Ha, B.C.

    1997-01-01

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

  14. Durable Glass For Thousands Of Years

    International Nuclear Information System (INIS)

    Jantzen, C.

    2009-01-01

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

  15. 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.

  16. Immobilization of hazardous and radioactive waste into glass structures

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Feldman, D.L.

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  19. Spent fuel and HLW transportation the French experience

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

    Blomqvist, G.

    1977-07-01

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

  3. Glass science tutorial: Lecture No. 7, Waste glass technology for Hanford

    International Nuclear Information System (INIS)

    Kruger, A.A.

    1995-07-01

    This paper presents the details of the waste glass tutorial session that was held to promote knowledge of waste glass technology and how this can be used at the Hanford Reservation. Topics discussed include: glass properties; statistical approach to glass development; processing properties of nuclear waste glass; glass composition and the effects of composition on durability; model comparisons of free energy of hydration; LLW glass structure; glass crystallization; amorphous phase separation; corrosion of refractories and electrodes in waste glass melters; and glass formulation for maximum waste loading

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    International Nuclear Information System (INIS)

    Amoroso, J.

    2011-01-01

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

  6. DEFENSE HIGH LEVEL WASTE GLASS DEGRADATION

    International Nuclear Information System (INIS)

    Ebert, W.

    2001-01-01

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

  7. Demonstration of pyrometallurgical processing for metal fuel and HLW

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  8. Development of gap filling technique in HLW repository

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  9. Historical fuel reprocessing and HLW management in Idaho

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  10. Comparing technical concepts for disposal of Belgian vitrified HLW

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  11. Improvement of database on glass dissolution

    International Nuclear Information System (INIS)

    Hayashi, Maki; Sasamoto, Hiroshi; Yoshikawa, Hideki

    2008-03-01

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

  12. Liquidus Temperature Data for DWPF Glass

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-12-29

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  18. 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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  1. Turning nuclear waste into glass

    Energy Technology Data Exchange (ETDEWEB)

    Pegg, Ian L.

    2015-02-15

    Vitrification has emerged as the treatment option of choice for the most dangerous radioactive waste. But dealing with the nuclear waste legacy of the Cold War will require state-of-the-art facilities and advanced glass formulations.

  2. 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.

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

    International Nuclear Information System (INIS)

    Ove Arup and Partners

    1990-01-01

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

  4. Technical and economic optimization study for HLW waste management

    International Nuclear Information System (INIS)

    Deffes, A.

    1989-01-01

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

  5. Public Perspectives in the Japanese HLW Disposal Program

    International Nuclear Information System (INIS)

    Inatsugu, Shigefumi; Takeuchi, Mitsuo; Kato, Toshiaki

    2006-01-01

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

  6. Thermal analysis of the vertical disposal for HLW

    International Nuclear Information System (INIS)

    Zhao Honggang; Wang Ju; Liu Yuemiao; Su Rui

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Grogan, H.

    1985-03-01

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

  8. Compas project stress analysis of HLW containers intermediate testwork

    International Nuclear Information System (INIS)

    Ove Arup and Partners London

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Ha, B.C.

    1993-01-01

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

  10. Tc Chemistry in HLW: Role of Organic Complexants

    International Nuclear Information System (INIS)

    Hess, Nancy S.; Conradsen, Steven D.

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    Kitayama, Kazumi; Sakabe, Yasushi; Ishiguro, Katsuhiko

    2007-01-01

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

  12. Leaching behavior of simulated high-level waste glass

    International Nuclear Information System (INIS)

    Kamizono, Hiroshi

    1987-03-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  15. Corrosion resistance of metal materials for HLW canister

    International Nuclear Information System (INIS)

    Furuya, Takashi; Muraoka, Susumu; Tashiro, Shingo

    1982-02-01

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

  16. Novel waste forms for HLW and ILW immobilisation

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, W. L.; Chemical Engineering

    2006-09-30

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  19. 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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  1. 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.

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

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

    International Nuclear Information System (INIS)

    Lee, Sung Ho; Hwang, Yong Soo

    2008-12-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  6. 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

  7. Modelling spent fuel and HLW behaviour in repository conditions

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    International Nuclear Information System (INIS)

    McKee, R.W.

    1983-02-01

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

  9. 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.

  10. Electrochromic Glasses.

    Science.gov (United States)

    1980-07-31

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

  11. Colloid formation during waste glass corrosion

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-12-01

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

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

    International Nuclear Information System (INIS)

    Piccolo, S.F.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    McKenzie, W.F.

    1990-01-01

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

  19. 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.

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

    International Nuclear Information System (INIS)

    Trifunac, A.D.

    1998-01-01

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

  1. Vitrification of HLW produced by uranium/molybdenum fuel reprocessing in cogema's cold crucible melter

    International Nuclear Information System (INIS)

    Quang, R. Do; Petitjean, V.; Hollebeque, F.; Pinet, O.; Flament, T.; Prodhomme, A.; Dalcorso, J. P.

    2003-01-01

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

  2. Vitrification of HLW Produced by Uranium/Molybdenum Fuel Reprocessing in COGEMA's Cold Crucible Melter

    International Nuclear Information System (INIS)

    Do Quang, R.; Petitjean, V.; Hollebecque, F.; Pinet, O.; Flament, T.; Prod'homme, A.

    2003-01-01

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

  3. Glass consistency and glass performance

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

    Jain, V.; Pan, Y.M.

    2004-01-01

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

  5. Colloidal glasses

    Indian Academy of Sciences (India)

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

  6. Assessing the performance of the Nagra HLW disposal concept

    International Nuclear Information System (INIS)

    Smith, P.; Zuidema, P.; McKinley, I.G.

    1995-01-01

    This article outlines the procedures used in safety assessment and illustrates their application in evaluating the performance of a high-level waste repository. Nagra's general safety assessment methodology has five main components: formulating the aims of the analysis, defining the safety concept, scenario development, consequence analysis and interpretation of results. A safety analysis based on conservative assumptions shows that the engineered barriers of the high-level waste repository are very effective in preventing release of radionuclides; this alone is sufficient to ensure that regulatory requirements can be met. The function of the host rock is to provide a favourable environment for the engineered barrier system. (author) 8 figs

  7. 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.

  8. An empirical modeling tool and glass property database in development of US-DOE radioactive waste glasses

    International Nuclear Information System (INIS)

    Muller, I.; Gan, H.

    1997-01-01

    An integrated glass database has been developed at the Vitreous State Laboratory of Catholic University of America. The major objective of this tool was to support glass formulation using the MAWS approach (Minimum Additives Waste Stabilization). An empirical modeling capability, based on the properties of over 1000 glasses in the database, was also developed to help formulate glasses from waste streams under multiple user-imposed constraints. The use of this modeling capability, the performance of resulting models in predicting properties of waste glasses, and the correlation of simple structural theories to glass properties are the subjects of this paper. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-01

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

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

    International Nuclear Information System (INIS)

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

    1995-03-01

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

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

    International Nuclear Information System (INIS)

    Cunnane, J.C.

    1994-03-01

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

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

    International Nuclear Information System (INIS)

    JANTZEN, CAROL

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Kitamoto, A.; Mulyanto

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Ove Arup and Partners, London

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Liu Xiaodong; Luo Taian; Zhu Guoping; Chen Qingchun

    2007-12-01

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

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

    International Nuclear Information System (INIS)

    Liu Linqing; Yu Yunxiang

    1994-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-12-15

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

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

    International Nuclear Information System (INIS)

    Sasaki, Noriaki

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-04-01

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

  1. 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....

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Fujiwara, Osamu; Sanga, Tomoji; Moriya, Toshifumi

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Lang-Lenton, J.

    2007-01-01

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

  6. Accelerated Leach Testing of GLASS: ALTGLASS Version 3.0

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-31

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

  7. Accelerated Leach Testing of GLASS: ALTGLASS Version 3.0

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  8. Application of intelligence based uncertainty analysis for HLW disposal

    International Nuclear Information System (INIS)

    Kato, Kazuyuki

    2003-01-01

    Safety assessment for geological disposal of high level radioactive waste inevitably involves factors that cannot be specified in a deterministic manner. These are namely: (1) 'variability' that arises from stochastic nature of the processes and features considered, e.g., distribution of canister corrosion times and spatial heterogeneity of a host geological formation; (2) 'ignorance' due to incomplete or imprecise knowledge of the processes and conditions expected in the future, e.g., uncertainty in the estimation of solubilities and sorption coefficients for important nuclides. In many cases, a decision in assessment, e.g., selection among model options or determination of a parameter value, is subjected to both variability and ignorance in a combined form. It is clearly important to evaluate both influences of variability and ignorance on the result of a safety assessment in a consistent manner. We developed a unified methodology to handle variability and ignorance by using probabilistic and possibilistic techniques respectively. The methodology has been applied to safety assessment of geological disposal of high level radioactive waste. Uncertainties associated with scenarios, models and parameters were defined in terms of fuzzy membership functions derived through a series of interviews to the experts while variability was formulated by means of probability density functions (pdfs) based on available data set. The exercise demonstrated applicability of the new methodology and, in particular, its advantage in quantifying uncertainties based on expert's opinion and in providing information on dependence of assessment result on the level of conservatism. In addition, it was also shown that sensitivity analysis could identify key parameters in reducing uncertainties associated with the overall assessment. The above information can be used to support the judgment process and guide the process of disposal system development in optimization of protection against

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Wald, J.W.

    1985-10-01

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

  12. 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....

  13. 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....

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

  15. 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.

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

    International Nuclear Information System (INIS)

    Wald, J.W.

    1985-10-01

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

  17. Radionuclides containment in nuclear glasses. An overview

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  18. Thermodynamic model of natural, medieval and nuclear waste glass durability

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Plodinec, M.J.

    1983-01-01

    A thermodynamic model of glass durability based on hydration of structural units has been applied to natural glass, medieval window glasses, and glasses containing nuclear waste. The relative durability predicted from the calculated thermodynamics correlates directly with the experimentally observed release of structural silicon in the leaching solution in short-term laboratory tests. By choosing natural glasses and ancient glasses whose long-term performance is known, and which bracket the durability of waste glasses, the long-term stability of nuclear waste glasses can be interpolated among these materials. The current Savannah River defense waste glass formulation is as durable as natural basalt from the Hanford Reservation (10 6 years old). The thermodynamic hydration energy is shown to be related to the bond energetics of the glass. 69 references, 2 figures, 1 table

  19. GLASS BOX

    National Research Council Canada - National Science Library

    Curtis, Laura

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-04-01

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

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

    International Nuclear Information System (INIS)

    Ryder, D.E.

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Lee, Youn-Myoung; Hwang, Yong-Soo

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Aurah, Mirwaise Y.; Roberts, Mark A.

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Hara, Shigeo; Terai, Ryohei; Yamanaka, Hiroshi

    1983-01-01

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

  10. Factors influencing chemical durability of nuclear waste glasses

    International Nuclear Information System (INIS)

    Feng, Xiangdong; Bates, J.K.

    1993-01-01

    A short summary is given of our studies on the major factors that affect the chemical durability of nuclear waste glasses. These factors include glass composition, solution composition, SA/V (ratio of glass surface area to the volume of solution), radiation, and colloidal formation. These investigations have enabled us to gain a better understanding of the chemical durability of nuclear waste glasses and to accumulate.a data base for modeling the long-term durability of waste glass, which will be used in the risk assessment of nuclear waste disposal. This knowledge gained also enhances our ability to formulate optimal waste glass compositions

  11. 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.

  12. A new glass option for parenteral packaging.

    Science.gov (United States)

    Schaut, Robert A; Peanasky, John S; DeMartino, Steven E; Schiefelbein, Susan L

    2014-01-01

    Glass is the ideal material for parenteral packaging because of its chemical durability, hermeticity, strength, cleanliness, and transparency. Alkali borosilicate glasses have been used successfully for a long time, but they do have some issues relating to breakage, delamination, and variation in hydrolytic performance. In this paper, alkali aluminosilicate glasses are introduced as a possible alternative to alkali borosilicate glasses. An example alkali aluminosilicate glass is shown to meet the compendial requirements, and to have similar thermal, optical, and mechanical attributes as the current alkali borosilicate glasses. In addition, the alkali aluminosilicate performed as well or better than the current alkali borosilicates in extractables tests and stability studies, which suggests that it would be suitable for use with the studied liquid product formulation. The physical, mechanical, and optical properties of glass make it an ideal material for packaging injectable drugs and biologics. Alkali borosilicate glasses have been used successfully for a long time for these applications, but there are some issues. In this paper, alkali aluminosilicate glasses are introduced as a possible alternative to alkali borosilicate glasses. An example alkali aluminosilicate glass is shown to meet the requirements for packaging injectable drugs and biologics, and to be suitable for use with a particular liquid drug. © PDA, Inc. 2014.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  14. 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.

  15. Glass as a matrix for SRP high-level defense waste

    International Nuclear Information System (INIS)

    Wiley, J.R.; Bibler, N.E.; Dukes, M.D.; Plodinec, M.J.

    1980-01-01

    Work done at Savannah River Laboratory and elsewhere that has led to development of glass as a candidate for solidifying Savannah River Plant waste is summarized. Areas of development described are glass formulation and fabrication, and leaching and radiation effects

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    International Nuclear Information System (INIS)

    Kitamoto, A.; Mulyanto

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, J.; Noseck, U.

    2015-06-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-14

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-15

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

  4. Proposal for geological site selection for L/ILW and HLW repositories. Statement of requirements, procedure and results. Technical report 08-03

    International Nuclear Information System (INIS)

    2008-10-01

    Important steps in the process of managing radioactive wastes have already been implemented in Switzerland. These include the handing and packaging of the waste, waste characterisation and documentation of waste inventories and interim storage along with associated transport. In terms of preparing for deep geological disposal, the necessary scientific and technical work is well advanced and the feasibility of constructing geological repositories that provide the required long-term safety has been successfully demonstrated for all waste types arising in Switzerland. Sufficient knowledge is available to allow the next steps in the selection of repository sites to be defined. The legal framework is also in place and organisational measures have been provided that will allow the tasks to be performed in the coming years to be implemented efficiently. The selection of geological siting regions and sites for repositories in Switzerland will be conducted in three stages. Stage 1 ends with the definition of geological siting regions within which the repository projects will be elaborated in more detail in stages 2 and 3. This report documents and justifies the siting proposals prepared by Nagra for the repositories for low- and intermediate-level waste (L/ILW) and high-level waste (HLW). Formulation of these proposals is conducted in five steps: 1) The waste inventory, which includes reserves for future developments, is allocated to the L/ILW and HLW repositories; 2) Based on this waste allocation, 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

  5. 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.

  6. Crystallization Formulation Lab

    Data.gov (United States)

    Federal Laboratory Consortium — The Crystallization Formulation Lab fills a critical need in the process development and optimization of current and new explosives and energetic formulations. The...

  7. Reciclagem de vidro laminado: utilização dos vidros de baixa granulometria como carga abrasiva na formulação de vernizes de alto tráfego para pisos de madeira Recycling of laminated glass: the use of low granulometry glasses as abrasive filler in the formulation of varnishes for wood flooring application

    Directory of Open Access Journals (Sweden)

    Isabella M. Vargas

    2007-06-01

    Full Text Available O vidro laminado, especialmente o proveniente dos pára-brisas, é de difícil reciclagem após sua utilização, sobretudo o resíduo de vidro constituído de partículas pequenas (com baixa granulometria. O vidro muito fino se caracteriza como um resíduo, uma vez que não pode ser recuperado na indústria vidreira, pois nos grandes fornos de fusão de vidro evita-se a utilização deste material para impedir que ocorra o fenômeno de arraste de pó pela chaminé, bem como a deposição de vidro fundido nas paredes internas do forno e do regenerador[1]. Neste estudo duas granulometrias de vidro foram utilizadas - malha 200 e 325 mesh - na formulação de vernizes baseados num oligômero uretano-acrilado e usado para pintura e proteção de pisos de madeira. A cura desses vernizes foi por meio de radiação ultravioleta. Os resultados encontrados mostraram que a resistência à abrasão do verniz com a adição de vidro fino aumentou significativamente.Windshields are difficult to recycle after use, especially the residue of fine glass (with small granulometry. This material is characterized as a residue since it cannot be recovered in the glass industry, because in the furnaces of glass fusion fine glass is avoided to prevent deposition on the internal walls of the furnaces and the drag of this fine powder through the chimney[1]. In this work two different granulometries of this fine glass - 200 mesh and 325 mesh - were used as a filler in a varnish based on a urethane-acrylate oligomer used to cover and protect wood flooring. The cure of this varnish was through ultraviolet radiation. The results showed increased abrasion resistance for the varnish with addition of the fine glass.

  8. Glass compositions

    Energy Technology Data Exchange (ETDEWEB)

    France, P W

    1985-05-30

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

  9. BNFL Report Glass Formers Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Schumacher, R.F.

    2000-07-27

    The objective of this task was to obtain powder property data on candidate glass former materials, sufficient to guide conceptual design and estimate the cost of glass former handling facilities as requested under Part B1 of BNFL Technical and Development Support. Twenty-nine glass forming materials were selected and obtained from vendors for the characterization of their physical properties, durability in caustic solution, and powder flow characteristics. A glass former was selected based on the characterization for each of the ten oxide classes required for Envelope A, B, and C mixtures. Three blends (A, B, and C) were prepared based on formulations provided by Vitreous State Laboratory and evaluated with the same methods employed for the glass formers. The properties obtained are presented in a series of attached Tables. It was determined that five of the ten glass formers, (kyanite, iron oxide, titania, zircon, and zinc oxide) have the potential to cause some level of solids f low problems. The problems might include arching or ratholing in the silo/hopper. In addition, all of the blends may require consideration for their handling.

  10. BNFL Report Glass Formers Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Schumacher, R.F.

    2000-07-27

    The objective of this task was to obtain powder property data on candidate glass former materials, sufficient to guide conceptual design and estimate the cost of glass former handling facilities as requested under Part B1 of BNFL Technical and Development Support. Twenty-nine glass forming materials were selected and obtained from vendors for the characterization of their physical properties, durability in caustic solution, and powder flow characteristics. A glass former was selected based on the characterization for each of the ten oxide classes required for Envelope A, B, and C mixtures. Three blends (A, B, and C) were prepared based on formulations provided by Vitreous State Laboratory and evaluated with the same methods employed for the glass formers. The properties obtained are presented in a series of attached Tables. It was determined that five of the ten glass formers, (kyanite, iron oxide, titania, zircon, and zinc oxide) have the potential to cause some level of solids f low problems. In addition, all of the blends may require consideration for their handling. A number of engineering considerations and recommendations were prepared based on the experimental findings, experience, and other process considerations. Recommendations for future testing are included. In conjunction with future work, it is recommended that a professional consultant be engaged to guide and assist with testing and design input.

  11. BNFL Report Glass Formers Characterization

    International Nuclear Information System (INIS)

    Schumacher, R.F.

    2000-01-01

    The objective of this task was to obtain powder property data on candidate glass former materials, sufficient to guide conceptual design and estimate the cost of glass former handling facilities as requested under Part B1 of BNFL Technical and Development Support. Twenty-nine glass forming materials were selected and obtained from vendors for the characterization of their physical properties, durability in caustic solution, and powder flow characteristics. A glass former was selected based on the characterization for each of the ten oxide classes required for Envelope A, B, and C mixtures. Three blends (A, B, and C) were prepared based on formulations provided by Vitreous State Laboratory and evaluated with the same methods employed for the glass formers. The properties obtained are presented in a series of attached Tables. It was determined that five of the ten glass formers, (kyanite, iron oxide, titania, zircon, and zinc oxide) have the potential to cause some level of solids f low problems. In addition, all of the blends may require consideration for their handling. A number of engineering considerations and recommendations were prepared based on the experimental findings, experience, and other process considerations. Recommendations for future testing are included. In conjunction with future work, it is recommended that a professional consultant be engaged to guide and assist with testing and design input

  12. BNFL Report Glass Formers Characterization

    International Nuclear Information System (INIS)

    Schumacher, R.F.

    2000-01-01

    The objective of this task was to obtain powder property data on candidate glass former materials, sufficient to guide conceptual design and estimate the cost of glass former handling facilities as requested under Part B1 of BNFL Technical and Development Support. Twenty-nine glass forming materials were selected and obtained from vendors for the characterization of their physical properties, durability in caustic solution, and powder flow characteristics. A glass former was selected based on the characterization for each of the ten oxide classes required for Envelope A, B, and C mixtures. Three blends (A, B, and C) were prepared based on formulations provided by Vitreous State Laboratory and evaluated with the same methods employed for the glass formers. The properties obtained are presented in a series of attached Tables. It was determined that five of the ten glass formers, (kyanite, iron oxide, titania, zircon, and zinc oxide) have the potential to cause some level of solids f low problems. The problems might include arching or ratholing in the silo/hopper. In addition, all of the blends may require consideration for their handling

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

    International Nuclear Information System (INIS)

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

    2002-03-01

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

  14. 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)

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

    2011-01-01

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

  15. 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

  16. Preliminary Technology Maturation Plan for Immobilization of High-Level Waste in Glass Ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, John D.; Crum, Jarrod V.; Sevigny, Gary J.; Smith, G L.

    2012-09-30

    A technology maturation plan (TMP) was developed for immobilization of high-level waste (HLW) raffinate in a glass ceramics waste form using a cold-crucible induction melter (CCIM). The TMP was prepared by the following process: 1) define the reference process and boundaries of the technology being matured, 2) evaluate the technology elements and identify the critical technology elements (CTE), 3) identify the technology readiness level (TRL) of each of the CTE’s using the DOE G 413.3-4, 4) describe the development and demonstration activities required to advance the TRLs to 4 and 6 in order, and 5) prepare a preliminary plan to conduct the development and demonstration. Results of the technology readiness assessment identified five CTE’s and found relatively low TRL’s for each of them: • Mixing, sampling, and analysis of waste slurry and melter feed: TRL-1 • Feeding, melting, and pouring: TRL-1 • Glass ceramic formulation: TRL-1 • Canister cooling and crystallization: TRL-1 • Canister decontamination: TRL-4 Although the TRL’s are low for most of these CTE’s (TRL-1), the effort required to advance them to higher values. The activities required to advance the TRL’s are listed below: • Complete this TMP • Perform a preliminary engineering study • Characterize, estimate, and simulate waste to be treated • Laboratory scale glass ceramic testing • Melter and off-gas testing with simulants • Test the mixing, sampling, and analyses • Canister testing • Decontamination system testing • Issue a requirements document • Issue a risk management document • Complete preliminary design • Integrated pilot testing • Issue a waste compliance plan A preliminary schedule and budget were developed to complete these activities as summarized in the following table (assuming 2012 dollars). TRL Budget Year MSA FMP GCF CCC CD Overall $M 2012 1 1 1 1 4 1 0.3 2013 2 2 1 1 4 1 1.3 2014 2 3 1 1 4 1 1.8 2015 2 3 2 2 4 2 2.6 2016 2 3 2 2 4 2 4

  17. A state-of-the-art report on the formulation and characterization of synroc

    International Nuclear Information System (INIS)

    Jung, C. H.; Park, J. Y.; Oh, S. J.; Kim, H. Y.; Kim, I. T.

    1998-09-01

    Synroc (Synthetic Rock), a titanate-based ceramic originally proposed by Prof. A. Ringwood (ANU) and designed for the immobilization of high level nuclear waste (HLW), consists of three principal phases such as hollandite, zirconolite and perovskite. Nearly all the fission products and actinides in HLW can be incorporated as solid-solution in at least one of these phase. The preferred form of Synroc can be obtained up to 20 % of high level waste calcine to form dilute solid solution. The constituent minerals, or close structural analogues, have survived in a wide range of geochemical environments for periods of 20-2000 Myr while immobilizing the same elements present in nuclear waste. A dense, compact, and mechanically strong form of Synroc can be formed by hot pressing reactive precursor powders at about 1200 dg C and 20 MPa. In this state-of-the-art report, formulation method and characterization of Syroc with respect to the crystal structure, the consisting substances, types, etc. were reviewed. Additionally, a new promising powder process, C ombustion Process , was proposed and the properties of the combustion-synthesized powder were described. An international cooperative program between JAERI and ANSTO, and US patents for early Synroc research in Australia were also introduced. From the literatures review, Synroc is expected to have advantages in using as an immobilizer of HLW. Therefore, a systematic research to develop the Synroc is needed. (author). 53 refs., 2 tabs., 16 figs

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-02-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  1. Glasses obtained from industrial wastes

    International Nuclear Information System (INIS)

    Bortoluzzi, D.; Oliveira Fillho, J.; Uggioni, E.; Bernardin, A.M.

    2009-01-01

    This paper deals with the study of the vitrification mechanism as an inertization method for industrial wastes contaminated with heavy metals. Ashes from coal (thermoelectric), wastes from mining (fluorite and feldspar) and plating residue were used to compose vitreous systems planed by mixture design. The chemical composition of the wastes was determined by XRF and the formulations were melted at 1450 deg C for 2h using 10%wt of CaCO 3 (fluxing agent). The glasses were poured into a mold and annealed (600 deg C). The characteristic temperatures were determined by thermal analysis (DTA, air, 20 deg C/min) and the mechanical behavior by Vickers microhardness. As a result, the melting temperature is strongly dependent on silica content of each glass, and the fluorite residue, being composed mainly by silica, strongly affects Tm. The microhardness of all glasses is mainly affected by the plating residue due to the high iron and zinc content of this waste. (author)

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  3. Wastes vitrification by plasma torch: study of a glass formulation compatible with a wide range of B wastes; Vitrification des dechets par torche a plasma: recherche d'une formulation de verre compatible avec un large eventail de dechets B

    Energy Technology Data Exchange (ETDEWEB)

    Poitou, S.; Richaud, D.; Fiquet, O.; Gramondi, P.; Massit, H. [CEA Cadarache, Dept. d' Entreposage et de Stockage des Dechets, 13 - Saint-Paul-lez-Durance (France)

    2001-07-01

    Within the context of radioactive waste management, CEA has equipped itself with a 'PLASMARC' device. The central element of this device is a plasma torch treatment furnace. It has been implemented and validated for the vitrification of low level radioactive wastes. Meanwhile, the plasma torch presents potentially interests for immobilizing under an inert form in vitreous matrices, B wastes which are generally divided and of complex chemical composition. The application of this process to this type of wastes has been studied here. The obtained results show that with the plasma torch it is possible to make glasses with a high amount of silicon and aluminium oxide and which are adapted to the treatment / packaging of the B wastes. (O.M.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

  5. Glass: Rotary Electric Glass Furnace

    Energy Technology Data Exchange (ETDEWEB)

    Recca, L.

    1999-01-29

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

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

    International Nuclear Information System (INIS)

    Rogers, V.

    1993-03-01

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

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

    International Nuclear Information System (INIS)

    Hahn, P. S.; Cho, W. J.; Kwon, S.

    2006-01-01

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

  8. Role of international collaboration in PNC's R ampersand D programme for HLW disposal

    International Nuclear Information System (INIS)

    Masuda, Sumio; Umeki, Hiroyuki; Yamakawa, Minoru

    1996-01-01

    PNC has been active in promoting international cooperation in connection with the Japanese HLW disposal programme, based on both a bilateral and multilateral approach. Both types of cooperation are extremely useful; in particular, bilateral cooperation has the advantage of providing opportunities for in-depth discussions in mutual areas of interest. By way of contrast, multilateral cooperation also provides an international arena for broader discussion and corroboration of output from individual R ampersand D programmes. International collaboration also provides young researchers with an opportunity to learn from experience. Depending on the issues to be tackled, appropriate forms of collaboration have been integrated into PNC's strategy for maximizing output. The lessons learned from collaboration are very valuable and can be used directly in their programme to enhance its credibility. The format of collaboration has also been extensively developed: it has been found that resources can be utilized more effectively by sharing them appropriately

  9. Depth optimization for the Korean HLW repository System within a discontinuous and saturated granitic rock mass

    International Nuclear Information System (INIS)

    Kim, Jhin Wung; Bae, Dae Seok; Choi, Jong Won

    2005-12-01

    The present study is to evaluate the material properties of the compacted bentonite, backfill material, canister cast iron insert, and the rock mass for the Korean HLW repository system. These material properties are either measured, or taken from other countries, through the evaluation of the thermal, hydraulic, and mechanical interaction behavior of a repository. After the evaluation of the material properties, the most appropriate and economical depth as well as the layout of a single layer repository is to be recommended. Material properties used for the granitic rock mass, rock joints, PWR spent fuel, disposal canister, compacted bentonite, backfill material, and ground water are the data collected domestically, and foreign data are used for some of the data not available domestically. The repository model includes a saturated granitic rock mass with joints, PWR spent fuel in a disposal canister surrounded by compacted bentonite inside a deposition hole, and backfill material in the rest of the space within a repository cavern

  10. Technical Standards on the Safety Assessment of a HLW Repository in Other Countries

    International Nuclear Information System (INIS)

    Lee, Sung Ho; Hwang, Yong Soo

    2009-01-01

    The basic function of HLW disposal system is to prevent excessive radio-nuclides being leaked from the repository in a short time. To do this, many technical standards should be developed and established on the components of disposal system. Safety assessment of a repository is considered as one of technical standards, because it produces quantitative results of the future evolution of a repository based on a reasonably simplified model. In this paper, we investigated other countries' regulations related to safely assessment focused on the assessment period, radiation dose limits and uncertainties of the assessment. Especially, in the investigation process of the USA regulations, the USA regulatory bodies' approach to assessment period and peak dose is worth taking into account in case of a conflict between peak dose from safety assessment and limited value in regulation.

  11. Influence of the reprocessing flow sheet on the HLW solidification technology

    International Nuclear Information System (INIS)

    Baetsle, L.H.

    1981-01-01

    The introduction of Pu recycled LWR and CMFBR fuel will require the addition of a second dissolution step to quantitation recover Pu. If process modifications can be brought to the head-end procedures it is advisable to remove Ru, Te, Mo, Pd by high performance centrifugation and to volatilize soluble RuNO(NO 3 ) 2 by sparing with ozone. This changes improve the liquid extraction efficiency and simplify the off gas treatment during calcination and vitrification of HAWC. The conversion of HAW to HAWC by evaporation is accompagnied by some-volatilization of Ru and Cs. Organic reductants reduce the Ru volatilization. The introduction of salt free reagents during feed adjustment steps will decrease Na content in the HLW. The main impact of the use of salt free reagent will have its bearing on the LAW and ILIW treatment and conditioning. (DG)

  12. HLW disposal by fission reactors; calculation of trans-mutation rate and recycle

    International Nuclear Information System (INIS)

    Mulyanto

    1997-01-01

    Transmutation of MA (Minor actinide) and LLFPS (long-lived fission products) into stable nuclide or short-lived isotopes by fission reactors seem to become an alternative technology for HLW disposal. in this study, transmutation rate and recycle calculation were developed in order to evaluate transmutation characteristics of MA and LLFPs in the fission reactors. inventory of MA and LLFPs in the transmutation reactors were determined by solving of criticality equation with 1-D cylindrical geometry of multigroup diffusion equations at the beginning of cycle (BOC). transmutation rate and burn-up was determined by solving of depletion equation. inventory of MA and LLFPs was calculated for 40 years recycle. From this study, it was concluded that characteristics of MA and LLFPs in the transmutation reactors can be evaluated by recycle calculation. by calculation of transmutation rate, performance of fission reactor for transmutation of MA or LLFPs can be discussed

  13. An Ilustrative Nuclide Release Behavior from an HLW Repository due to an Earthquake Event

    International Nuclear Information System (INIS)

    Lee, Youn-Myoung; Hwang, Yong-Soo; Choi, Jong-Won

    2008-01-01

    Program for the evaluation of a high-level waste repository which is conceptually modeled. During the last few years, programs developed with the aid of AMBER and GoldSim by which nuclide transports in the near- and far-field of a repository as well as transport through the biosphere under various normal and disruptive release scenarios could be modeled and evaluated, have been continuously demonstrated. To show its usability, as similarly done for the natural groundwater flow scheme, influence of a possible disruptive event on a nuclide release behavior from an HLW repository system caused naturally due to an earthquake has been investigated and illustrated with the newly developed GoldSim program

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

    International Nuclear Information System (INIS)

    Hwang, Y. S.; Park, J. B.; Kang, C. H.

    2002-01-01

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

  15. Current status of preparing buffer/backfill block in HLW disposal abroad

    International Nuclear Information System (INIS)

    Yan Ming; Wang Xuewen; Zhang Huyuan

    2014-01-01

    There is an urgent need for China to commence the full-scale compaction test, resolving the preparation problem for buffer/backfill blocks when underground research laboratory project is planned for High Level Radioactive Waste (HLW) disposal. The foreign countries have some research about the preparation of buffer/backfill blocks in engineered barrier systems. The foreign research shows that installation of clay blocks with sector shape at waste pollution area is a feasible engineering method. Compacted clay blocks need to be cured in a cabinet with controlled temperature and humidity to avoid desiccation and surface powdering. A freeze mixing method, mixing powdered-ice and cooled bentonite, can be operated more easily and obtain more uniform hydration than the traditional mixing of water and bentonite. It is helpful to review and adsorb the foreign research results for the design of full-scale test of bentonite compaction. (authors)

  16. Glass binder development for a glass-bonded sodalite ceramic waste form

    International Nuclear Information System (INIS)

    Riley, Brian J.; Vienna, John D.; Frank, Steven M.; Kroll, Jared O.; Peterson, Jacob A.

    2017-01-01

    This paper discusses work to develop Na_2O-B_2O_3-SiO_2 glass binders for immobilizing LiCl-KCl eutectic salt waste in a glass-bonded sodalite waste form following electrochemical reprocessing of used metallic nuclear fuel. In this paper, five new glasses with ~20 mass% Na_2O were designed to generate waste forms with high sodalite. The glasses were then used to produce ceramic waste forms with a surrogate salt waste. The waste forms made using these new glasses were formulated to generate more sodalite than those made with previous baseline glasses for this type of waste. The coefficients of thermal expansion for the glass phase in the glass-bonded sodalite waste forms made with the new binder glasses were closer to the sodalite phase in the critical temperature region near and below the glass transition temperature than previous binder glasses used. Finally, these improvements should result in lower probability of cracking in the full-scale monolithic ceramic waste form, leading to better long-term chemical durability.

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

    International Nuclear Information System (INIS)

    Sreten Mastilovic

    2001-01-01

    The objective of this calculation is to determine the structural response of the standard high-level waste (HLW) canister and the canister containing the cans of immobilized plutonium (Pu) (''can-in-canister'' [CIC] throughout this document) subjected to drop DBEs (design basis events) during the handling operation. The evaluated DBE in the former case is 7-m (23-ft) vertical (flat-bottom) drop. In the latter case, two 2-ft (0.61-m) corner (oblique) drops are evaluated in addition to the 7-m vertical drop. These Pu CIC calculations are performed at three different temperatures: room temperature (RT) (20 C), T = 200 F = 93.3 C , and T = 400 F = 204 C ; in addition to these the calculation characterized by the highest maximum stress intensity is performed at T = 750 F = 399 C as well. The scope of the HLW canister calculation is limited to reporting the calculation results in terms of: stress intensity and effective plastic strain in the canister, directional residual strains at the canister outer surface, and change of canister dimensions. The scope of Pu CIC calculation is limited to reporting the calculation results in terms of stress intensity, and effective plastic strain in the canister. The information provided by the sketches from Reference 26 (Attachments 5.3,5.5,5.8, and 5.9) is that of the potential CIC design considered in this calculation, and all obtained results are valid for this design only. This calculation is associated with the Plutonium Immobilization Project and is performed by the Waste Package Design Section in accordance with Reference 24. It should be noted that the 9-m vertical drop DBE, included in Reference 24, is not included in the objective of this calculation since it did not become a waste acceptance requirement. AP-3.124, ''Calculations'', is used to perform the calculation and develop the document

  18. Nitrate glass

    International Nuclear Information System (INIS)

    Kirilenko, I.A.; Vinogradov, E.E.

    1977-01-01

    Experimental evidence on behaviour of nitrate glasses is reviewed in terms of relationships between the presence of water in vitrescent nitrate systems and the properties of the systems. The glasses considered belong to systems of Mg(NO 3 ) 2 - Nd(NO 3 ) 3 ; Hg(NO 3 ) 2 -Nd(NO 3 ) 3 ; NaNO 3 -Mg(NO 3 ) 2 -Nd(NO 3 ) 3 ; M-Zn(NO 3 ) 3 , where M is a mixture of 20% mass NaNO 3 and 80% mass Mg(NO 3 ) 2 , and Zn is a rare earth ion. Nitrate glass is shown to be a product of dehydration. Vitrification may be regarded as a resusl of formation of molecular complexes in the chain due to hydrogen bonds of two types, i.e. water-water, or water-nicrate group. Chain formation, along with low melting points of the nitrates, hinder crystallization of nitrate melts. Provided there is enough water, this results in vitrification

  19. Systems approach to nuclear waste glass development

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1986-01-01

    Development of a host solid for the immobilization of nuclear waste has focused on various vitreous wasteforms. The systems approach requires that parameters affecting product performance and processing be considered simultaneously. Application of the systems approach indicates that borosilicate glasses are, overall, the most suitable glasses for the immobilization of nuclear waste. Phosphate glasses are highly durable; but the glass melts are highly corrosive and the glasses have poor thermal stability and low solubility for many waste components. High-silica glasses have good chemical durability, thermal stability, and mechanical stability, but the associated high melting temperatures increase volatilization of hazardous species in the waste. Borosilicate glasses are chemically durable and are stable both thermally and mechanically. The borosilicate melts are generally less corrosive than commercial glasses, and the melt temperature miimizes excessive volatility of hazardous species. Optimization of borosilicate waste glass formulations has led to their acceptance as the reference nuclear wasteform in the United States, United Kingdom, Belgium, Germany, France, Sweden, Switzerland, and Japan

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

    International Nuclear Information System (INIS)

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

    1986-09-01

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

  1. Simulation of HTM processes in buffer-rock barriers based on the French HLW disposal concept

    International Nuclear Information System (INIS)

    Li, Xiaoshuo; Roehlig, Klaus-Juergen; Zhang, Chunliang

    2012-01-01

    Document available in extended abstract form only. The main objectives of this paper are to gain experience with modelling and analysis of HTM processes in clay rock and bentonite buffer surrounding heat-generating radioactive waste. The French concept for HLW disposal in drifts with backfilled bentonite buffer considered in numerical calculations which are carried out by using the computer code CODE-BRIGHT developed by the Technical University of Catalonia in Barcelona. The French repository designed by ANDRA is located in the middle of the Callovo-Oxfordian argillaceous formation (COX) of 250 m thickness at a depth of 500 to 630 m below the surface. The French concept has been simplified at this simulation work. A drift is considered to be excavated at a depth of 500 m below the surface. It has a diameter of 2.2 m and a length of 20 m. A large volume of the rock mass around the drift is taken into account by an axisymmetric model of 100 m radius and 100 m length. In fact, this model represents a cylindrical rock-buffer-system with the central axis of the containers, as shown in Figure 1. Some points are selected in the buffer and the rock along the radial line (dash yellow line) in the middle of the drift for recording HTM parameters with time. The display and analysis of the results at this paper are chiefly along this line. The simulation work has been divided to two time steps. At the first step, the drift excavation and ventilation is simulated by reducing the stress normal to the drift wall down to zero and circulating gas along the drift wall with relative humidity of 85 %. Following the drift excavation and ventilation, the HLW containers and the bentonite are emplaced in the drift as the second step of the simulation. This is simulated by simultaneously applying the initial conditions of the buffer and the decayed heat emitting from the waste containers as thermal boundary conditions. Two materials (Clay rock and bentonite buffer) are taken into account

  2. 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.

  3. 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)

    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

  4. Glass optimization for vitrification of Hanford Site low-level tank waste

    International Nuclear Information System (INIS)

    Feng, X.; Hrma, P.R.; Westsik, J.H. Jr.

    1996-03-01

    The radioactive defense wastes stored in 177 underground single-shell tanks (SST) and double-shell tanks (DST) at the Hanford Site will be separated into low-level and high-level fractions. One technology activity underway at PNNL is the development of glass formulations for the immobilization of the low-level tank wastes. A glass formulation strategy has been developed that describes development approaches to optimize glass compositions prior to the projected LLW vitrification facility start-up in 2005. Implementation of this strategy requires testing of glass formulations spanning a number of waste loadings, compositions, and additives over the range of expected waste compositions. The resulting glasses will then be characterized and compared to processing and performance specifications yet to be developed. This report documents the glass formulation work conducted at PNL in fiscal years 1994 and 1995 including glass formulation optimization, minor component impacts evaluation, Phase 1 and Phase 2 melter vendor glass development, liquidus temperature and crystallization kinetics determination. This report also summarizes relevant work at PNNL on high-iron glasses for Hanford tank wastes conducted through the Mixed Waste Integrated Program and work at Savannah River Technology Center to optimize glass formulations using a Plackett-Burnam experimental design

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

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, Joseph V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Freedman, Vicky L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-09-28

    Approximately 50 million gallons of high-level radioactive mixed waste has accumulated in 177 buried single- and double-shell tanks at the Hanford Site in southeastern Washington State as a result of the past production of nuclear materials, primarily for defense uses. The United States Department of Energy (DOE) is proceeding with plans to permanently dispose of this waste. Plans call for separating the tank waste into high-level waste (HLW) and low-activity waste (LAW) fractions, which will be vitrified at the Hanford Waste Treatment and Immobilization Plant (WTP). Principal radionuclides of concern in LAW are 99Tc, 129I, and U, while non-radioactive contaminants of concern are Cr and nitrate/nitrite. HLW glass will be sent off-site to an undetermined federal site for deep geological disposal while the much larger volume of immobilized low-activity waste will be placed in the on-site, near-surface Integrated Disposal Facility (IDF).

  6. Direction of CRT waste glass processing: electronics recycling industry communication.

    Science.gov (United States)

    Mueller, Julia R; Boehm, Michael W; Drummond, Charles

    2012-08-01

    Cathode Ray Tube, CRT, waste glass recycling has plagued glass manufacturers, electronics recyclers and electronics waste policy makers for decades because the total supply of waste glass exceeds demand, and the formulations of CRT glass are ill suited for most reuse options. The solutions are to separate the undesirable components (e.g. lead oxide) in the waste and create demand for new products. Achieving this is no simple feat, however, as there are many obstacles: limited knowledge of waste glass composition; limited automation in the recycling process; transportation of recycled material; and a weak and underdeveloped market. Thus one of the main goals of this paper is to advise electronic glass recyclers on how to best manage a diverse supply of glass waste and successfully market to end users. Further, this paper offers future directions for academic and industry research. To develop the recommendations offered here, a combination of approaches were used: (1) a thorough study of historic trends in CRT glass chemistry; (2) bulk glass collection and analysis of cullet from a large-scale glass recycler; (3) conversations with industry members and a review of potential applications; and (4) evaluation of the economic viability of specific uses for recycled CRT glass. If academia and industry can solve these problems (for example by creating a database of composition organized by manufacturer and glass source) then the reuse of CRT glass can be increased. Copyright © 2012 Elsevier Ltd. All rights reserved.

  7. Investigations into the stabilisation of drugs by sugar glasses : I. Tablets prepared from stabilised alkaline phosphatase

    NARCIS (Netherlands)

    Eriksson, H.J.C.; Hinrichs, W.L.J.; van Veen, B.; Somsen, G.W.; de Jong, G.J.; Frijlink, H.W.

    2002-01-01

    The aim of this study was to investigate the formulation of sugar glass stabilised alkaline phosphatase from bovine intestine (BIAP) into tablets. Two major subjects of tablet formulation were investigated. First, the compaction behaviour of the inulin sugar glass was investigated. Secondly, the

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

    Energy Technology Data Exchange (ETDEWEB)

    DUNCAN, G.P.

    2000-02-28

    The primary purpose of this business case is to provide Operations and Maintenance with a detailed transfer process review for the first High Level Waste (HLW) feed delivery to the Privatization Contractor (PC), AZ-101 batch transfer to PC. The Team was chartered to identify improvements that could be implemented in the field. A significant penalty can be invoked for not providing the quality, quantity, or timely delivery of HLW feed to the PC.

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Lindstroem Jensen, K.E.

    1987-01-01

    The study, which is a support to the Pagis project, covers three tasks concerning the evaluation of the Danish salt dome Mors (variant disposal site): evaluation of the human intrusion scenario where a cavern is excavated near the HLW-repository by solution mining technique. The waste is supposed to be leached during the operation period until the abandoned cavern is closed by convergence and the contaminated brine is pressed up into the overburden. Evaluation of the brine intrusion scenario, where the HLW-repository is inadvertently located close to a major brine pocket which subsequently releases its brine content through defects in the repository to the discharge stream for the catchment area. Collection and description of hydrological data of surface and deep layers (down to circa 700 metres) in the repository region. The data will be used by GSF to calculate the radionuclide migration in the geosphere

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

    Energy Technology Data Exchange (ETDEWEB)

    Perry, Frank Vinton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kelley, Richard E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-09-14

    The DOE Spent Fuel and Waste Technology (SWFT) R&D Campaign is supporting research on crystalline rock, shale (argillite) and salt as potential host rocks for disposal of HLW and SNF in a mined geologic repository. The distribution of these three potential repository host rocks is limited to specific regions of the US and to different geologic and hydrologic environments (Perry et al., 2014), many of which may be technically suitable as a site for mined geologic disposal. This report documents a regional geologic evaluation of the Pierre Shale, as an example of evaluating a potentially suitable shale for siting a geologic HLW repository. This report follows a similar report competed in 2016 on a regional evaluation of crystalline rock that focused on the Superior Province of the north-central US (Perry et al., 2016).

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

    International Nuclear Information System (INIS)

    McKinley, Ian G.; Apted, Mick; Umeki, Hiroyuki; Kawamura, Hideki

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Conradt, R.; Roggendorf, H.

    1985-01-01

    The purpose of the reported work was to determine the corrosion behaviour of the inactive HLW glass SM 58 LW 11 in Q-solution at temperatures up to 200 0 C and elevated pressures up to 13 MPa. In particular, a parametric study on the effects of time, temperature, pressure, crystallization, metallic impurities a.o. was performed. Further tests helped to identify the rate determining steps in the entire process and the most likely long-term corrosion law. (orig./RB)

  14. Waste vitrification: prediction of acceptable compositions in a lime-soda-silica glass-forming system

    International Nuclear Information System (INIS)

    Gilliam, T.M.; Jantzen, C.M.

    1996-10-01

    A model is presented based upon calculated bridging oxygens which allows the prediction of the region of acceptable glass compositions for a lime-soda-silica glass-forming system containing mixed waste. The model can be used to guide glass formulation studies (e.g., treatability studies) or assess the applicability of vitrification to candidate waste streams

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Sik; Kim, Kil Jung; Yang, Young Jin; Kim, Sung Chun [KOPEC, Taejeon (Korea)

    2002-03-01

    To estimate Total Life Cycle Cost (TSLCC) for Korea HLW Repository through localization of element cost data, we review and re-organize each basic element cost data for reference repository system, localize various element cost and finally estimate TSLCC considering economic parameters. As results of the study, TSLCC is estimated as 17,167,689 million won, which includes costs for site preparation, surface facilities, underground facilities and management/integration. Since HLW repository Project is an early stage of pre-conceptual design at present, the information of design and project information are not enough to perform cost estimate and cost localization for the Project. However, project cost structure is re-organized based on the local condition and Total System Life Cycle Cost is estimated using the previous cost data gathered from construction experience of the local nuclear power plant. Project results can be used as basic reference data to assume total construction cost for the local HLW repository and should be revised to more reliable cost data with incorporating detail project design information into the cost estimate in a future. 20 refs. (Author)

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

    International Nuclear Information System (INIS)

    Kim, Eun Ka; Suh, In Suk; Ro, Seong Gy; Yoo, Kun Joong; Yoo, Jae Hyung; Cho, Sung Soo

    2010-12-01

    The purpose of this study is to suggest development direction of related technologies to analyze patented technology filed as a leading technology and to identify the technology trend for developing HLW management technology suitable for atomic industrial environment in Korea. For patent analysis of HLW management technology, international patent data were collected. And international application number, patent share of applicant and nationality, annual number of applications, application trends of assignees and detail technology, and frequency of patent citations / citations-to were analyzed by statistical analysis. Technical level and competitiveness through quantitative analysis by indicators of patent analysis were confirmed. And technology developments of blank technology, similarity analysis, the point of the main patent and a range of patent rights were analyzed through in-depth analysis. Trends of the patented technology of our country and world patent technology in such results have been identified, and statistical data on patents were secured. Especially in HLW management technology, patent application in Korea compared ti United States, Japan and European Union was began much later for the '90s, and are showing the annual increase on trend of patent application. Patent trend in Korea corresponds to development generation, while declining in foreign patent. The result of this study will be usefully applied to setting a development direction and blank technology of patent technology to pursue future in Korea

  18. CRYSTALLIZATION IN MULTICOMPONENT GLASSES

    Energy Technology Data Exchange (ETDEWEB)

    KRUGER AA; HRMA PR

    2009-10-08

    In glass processing situations involving glass crystallization, various crystalline forms nucleate, grow, and dissolve, typically in a nonuniform temperature field of molten glass subjected to convection. Nuclear waste glasses are remarkable examples of multicomponent vitrified mixtures involving partial crystallization. In the glass melter, crystals form and dissolve during batch-to-glass conversion, melter processing, and product cooling. Crystals often agglomerate and sink, and they may settle at the melter bottom. Within the body of cooling glass, multiple phases crystallize in a non-uniform time-dependent temperature field. Self-organizing periodic distribution (the Liesegnang effect) is common. Various crystallization phenomena that occur in glass making are reviewed.

  19. Crystallization In Multicomponent Glasses

    International Nuclear Information System (INIS)

    Kruger, A.A.; Hrma, P.R.

    2009-01-01

    In glass processing situations involving glass crystallization, various crystalline forms nucleate, grow, and dissolve, typically in a nonuniform temperature field of molten glass subjected to convection. Nuclear waste glasses are remarkable examples of multicomponent vitrified mixtures involving partial crystallization. In the glass melter, crystals form and dissolve during batch-to-glass conversion, melter processing, and product cooling. Crystals often agglomerate and sink, and they may settle at the melter bottom. Within the body of cooling glass, multiple phases crystallize in a non-uniform time-dependent temperature field. Self-organizing periodic distribution (the Liesegnang effect) is common. Various crystallization phenomena that occur in glass making are reviewed.

  20. Bioactive type glass-ceramics within incorporated aluminium

    International Nuclear Information System (INIS)

    Volzone, C.; Stabile, F.M.; Ortiga, J.

    2012-01-01

    Bioactive glass-ceramics are used as biomaterials for the reparation of bone tissue. They are prepared, generally, by bioglass of specific composition for each particular use. The aluminium addition in the formulation at very small quantities influences on the structural properties. Two glass-ceramics obtained by P 2 O 5 -Na 2 O-CaO-SiO 2 formulation within aluminium (0.5 % in Al 2 O 3 base) added through a reactive alumina and purified feldspar were analyzed. The results showed structural differences between both glass-ceramics. (author)

  1. Investigation of lead-iron-phosphate glass for SRP waste

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1986-10-01

    The search for a host solid for the immobilization of nuclear waste has focused on various vitreous waste forms. Recently, lead-iron-phosphate (LIP) glasses have been proposed for solidification of all types of HLLW. Investigation of this glass for vitrification of SRP waste demonstrated that the phosphate glass is incompatible with the current borosilicate glass technology. The durability of LIP glasses in deionized water was comparable to current borosilicate waste glass formulations, and the LIP glass has a low melt temperature. However, many of the defense waste constituents have low solubility in the phosphate melt, producing an inhomogeneous product. Also, the LIP melt is highly corrosive which prevents the use of current melter materials, in particular Inconel 690, and thus requires more exotic materials of construction such as platinum

  2. Time evolution of dissolved oxygen and redox conditions in a HLW repository

    International Nuclear Information System (INIS)

    Wersin, P.; Spahiu, K.; Bruno, J.

    1994-02-01

    The evolution of oxygen in a HLW repository has been studied using presently available geochemical background information. The important processes affecting oxygen migration in the near-field include diffusion and oxidation of pyrite and dissolved Fe(II). The evaluation of time scales of oxygen decrease is carried out with 1. an analytical approach involving the coupling of diffusion and chemical reaction, 2. a numerical geochemical approach involving the application of a newly developed diffusion-extended version of the STEADYQL code. Both approaches yield consistent rates of oxygen decrease and indicate that oxidation of pyrite impurities in the clay is the dominant process. The results obtained fRom geochemical modelling are interpreted in terms of evolution of redox conditions. Moreover, a sensitivity analysis of the major geochemical and physical parameters is performed. These results indicate that the uncertainties associated with reactive pyrite surface area impose the overall uncertainties of prediction of time scales. Thus, the obtained time of decrease to 1% of initial O 2 concentrations range between 7 and 290 years. The elapsed time at which the transition to anoxic conditions occurs is estimated to be within the same time range. Additional experimental information on redox sensitive impurities in the envisioned buffer and backfill material would further constrain the evaluated time scales. 41 refs

  3. Use of Gap-fills in the Buffer and Backfill of an HLW Repository

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Owan; Lee, Min Soo; Choi, Heui Joo [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The buffer and backfill are significant barrier components of the repository. They play the roles of preventing the inflow of groundwater from the surrounding rock, retarding the release of radionuclides from the waste, supporting disposal container against external impacts, and discharging decay heat from the waste. When the buffer and backfill are installed for the HLW repository, there may be gaps between the container and buffer and between the backfill and the wall of disposal tunnels, respectively. These gaps occur because spaces are allowed for ease of the installation of the buffer and backfill in excavated deposition boreholes and disposal tunnels. If the gaps are left without any sealing as they are, however, the buffer and backfill can't accomplish their functions as the barrier components. This paper reviews the gap-fill concepts of the developed foreign countries, and then suggests a gap-fill concept which is applicable for the KRS. The gap-fill is suggested to employ bentonite- based materials with a type of pellet, granule, and pellet-granule mixture. The roller compression method and extrusion-cutting method are applicable for the fabrication of the bentonite pellets which can have the high density and the required amount for use to the buffer and backfill. For the installation of the gap-fill, the pouring and then pressing method and the shotcrete- blowing method are preferable for the gap of the deposition borehole and the gap of the disposal tunnel, respectively.

  4. Performance Assessment Uncertainty Analysis for Japan's HLW Program Feasibility Study (H12)

    International Nuclear Information System (INIS)

    BABA, T.; ISHIGURO, K.; ISHIHARA, Y.; SAWADA, A.; UMEKI, H.; WAKASUGI, K.; WEBB, ERIK K.

    1999-01-01

    Most HLW programs in the world recognize that any estimate of long-term radiological performance must be couched in terms of the uncertainties derived from natural variation, changes through time and lack of knowledge about the essential processes. The Japan Nuclear Cycle Development Institute followed a relatively standard procedure to address two major categories of uncertainty. First, a FEatures, Events and Processes (FEPs) listing, screening and grouping activity was pursued in order to define the range of uncertainty in system processes as well as possible variations in engineering design. A reference and many alternative cases representing various groups of FEPs were defined and individual numerical simulations performed for each to quantify the range of conceptual uncertainty. Second, parameter distributions were developed for the reference case to represent the uncertainty in the strength of these processes, the sequencing of activities and geometric variations. Both point estimates using high and low values for individual parameters as well as a probabilistic analysis were performed to estimate parameter uncertainty. A brief description of the conceptual model uncertainty analysis is presented. This paper focuses on presenting the details of the probabilistic parameter uncertainty assessment

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

    International Nuclear Information System (INIS)

    Kitayama, K.; Ishiguro, K.; Takeuchi, M.; Tsuchi, H.; Kato, T.; Sakabe, Y.; Wakasugi, K.

    2008-01-01

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

  6. Cleanup of a HLW nuclear fuel-reprocessing center using 3-D database modeling technology

    International Nuclear Information System (INIS)

    Sauer, R.C.

    1992-01-01

    A significant challenge in decommissioning any large nuclear facility is how to solidify the large volume of residual high-level radioactive waste (HLW) without structurally interfering with the existing equipment and piping used at the original facility or would require rework due to interferences which were not identified during the design process. This problem is further compounded when the nuclear facility to be decommissioned is a 35 year old nuclear fuel reprocessing center designed to recover usable uranium and plutonium. Facilities of this vintage usually tend to lack full documentation of design changes made over the years and as a result, crude traps or pockets of high-level contamination may not be fully realized. Any miscalculation in the construction or modification sequences could compound the overall dismantling and decontamination of the facility. This paper reports that development of a 3-dimensional (3-D) computer database tool was considered critical in defining the most complex portions of this one-of-a-kind vitrification facility

  7. Nuclide release calculation in the near-field of a reference HLW repository

    International Nuclear Information System (INIS)

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

    2004-01-01

    The HLW-relevant R and D program for disposal of high-level radioactive waste has been carried out at Korea Atomic Energy Research Institute (KAERI) since early 1997 in order to develop a conceptual Korea Reference Repository System for direct disposal of nuclear spent fuel by the end of 2007. A preliminary reference geologic repository concept considering such established criteria and requirements as waste and generic site characteristics in Korea was roughly envisaged in 2003 focusing on the near-field components of the repository system. According to above basic repository concept, which is similar to that of Swedish KBS-3 repository, the spent fuel is first encapsulated in corrosion resistant canisters, even though the material has not yet been determined, and then emplaced into the deposition holes surrounded by high density bentonite clay in tunnels constructed at a depth of about 500 m in a stable plutonic rock body. Not only to demonstrate how much a reference repository is safe in the generic point of view with several possible scenarios and cases associated with a preliminary repository concept by conducting calculations for nuclide release and transport in the near-field components of the repository, even though enough information has not been available that much yet, but also to show a methodology by which a generic safety assessment could be performed for further development of Korea reference repository concept, nuclide release calculation study strongly seems to be necessary

  8. Biosphere Modeling for the Dose Assessment of a HLW Repository: Development of ACBIO

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-01-15

    For the purpose of evaluating a dose rate to an individual due to a long-term release of nuclides from a HLW repository, a biosphere assessment model and an implemented code, ACBIO, based on the BIOMASS methodology have been developed by utilizing AMBER, a general compartment modeling tool. To demonstrate its practicability and usability as well as to observe the sensitivity of the compartment scheme, the concentration, the activity in the compartments as well as the annual flux between the compartments at their peak values, were calculated and investigated. For each case when changing the structure of the compartments and GBIs as well as varying selected input Kd values, all of which seem very important among the others, the dose rate per nuclide release rate is calculated separately and analyzed. From the maximum dose rates, the flux to dose conversion factors for each nuclide were derived, which are used for converting the nuclide release rate appearing from the geosphere through various GBIs to dose rates (Sv/y) for an individual in a critical group. It has also been observed that the compartment scheme, the identification of a possible exposure group and the GBIs could all be highly sensitive to the final consequences in a biosphere modeling.

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

    International Nuclear Information System (INIS)

    Ahner, S.; Dekais, J.J.; Puttke, B.; Staner, P.

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    Makino, Hitoshi; Kawamura, Makoto; Wakasugi, Keiichiro; Okubo, Hiroo; Takase, Hiroyasu

    2007-02-01

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

  11. Biosphere Modeling for the Dose Assessment of a HLW Repository: Development of ACBIO

    International Nuclear Information System (INIS)

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

    2006-01-01

    For the purpose of evaluating a dose rate to an individual due to a long-term release of nuclides from a HLW repository, a biosphere assessment model and an implemented code, ACBIO, based on the BIOMASS methodology have been developed by utilizing AMBER, a general compartment modeling tool. To demonstrate its practicability and usability as well as to observe the sensitivity of the compartment scheme, the concentration, the activity in the compartments as well as the annual flux between the compartments at their peak values, were calculated and investigated. For each case when changing the structure of the compartments and GBIs as well as varying selected input Kd values, all of which seem very important among the others, the dose rate per nuclide release rate is calculated separately and analyzed. From the maximum dose rates, the flux to dose conversion factors for each nuclide were derived, which are used for converting the nuclide release rate appearing from the geosphere through various GBIs to dose rates (Sv/y) for an individual in a critical group. It has also been observed that the compartment scheme, the identification of a possible exposure group and the GBIs could all be highly sensitive to the final consequences in a biosphere modeling

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

    International Nuclear Information System (INIS)

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Lin, Chi-Wen; Antaki, G.; Bandyopadhyay, K.; Bush, S.H.; Costantino, C.; Kennedy, R.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Smith, G.M.; Little, R.H.; Watkins, B.M.

    1999-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Mougnaud, Sarah

    2016-01-01

    High-level radioactive waste (HLW) remaining after spent nuclear fuel reprocessing is immobilized within a glass matrix, eventually destined for geological disposal. Water intrusion into the repository is expected after several thousand years. The alteration of a non-radioactive surrogate for nuclear glass has been extensively studied and it has been determined that successive leaching mechanisms lead to the formation of a 'passivating' alteration layer and to the establishment of a residual rate regime in the long term. However, glass packages are submitted to the radioactivity of confined radioelements. This work focuses on the influence of irradiation on the alteration layer formed during the residual rate regime, in a structural and mechanistic point of view. Three focal areas have been selected. Non-radioactive simple glasses have been leached and externally irradiated in order to determine modifications induced by electronic effects (irradiations with electrons and alpha particles). The same type of glass samples have been previously irradiated with heavy ions and their leaching behavior have been studied in order to assess the impact of ballistic dose cumulated by the glass before water intrusion. Leaching behavior of a complex radioactive glass, doped with an alpha-emitter, has been studied to consider a more realistic situation. (author) [fr

  17. Direction of CRT waste glass processing: Electronics recycling industry communication

    International Nuclear Information System (INIS)

    Mueller, Julia R.; Boehm, Michael W.; Drummond, Charles

    2012-01-01

    Highlights: ► Given a large flow rate of CRT glass ∼10% of the panel glass stream will be leaded. ► The supply of CRT waste glass exceeded demand in 2009. ► Recyclers should use UV-light to detect lead oxide during the separation process. ► Recycling market analysis techniques and results are given for CRT glass. ► Academic initiatives and the necessary expansion of novel product markets are discussed. - Abstract: Cathode Ray Tube, CRT, waste glass recycling has plagued glass manufacturers, electronics recyclers and electronics waste policy makers for decades because the total supply of waste glass exceeds demand, and the formulations of CRT glass are ill suited for most reuse options. The solutions are to separate the undesirable components (e.g. lead oxide) in the waste and create demand for new products. Achieving this is no simple feat, however, as there are many obstacles: limited knowledge of waste glass composition; limited automation in the recycling process; transportation of recycled material; and a weak and underdeveloped market. Thus one of the main goals of this paper is to advise electronic glass recyclers on how to best manage a diverse supply of glass waste and successfully market to end users. Further, this paper offers future directions for academic and industry research. To develop the recommendations offered here, a combination of approaches were used: (1) a thorough study of historic trends in CRT glass chemistry; (2) bulk glass collection and analysis of cullet from a large-scale glass recycler; (3) conversations with industry members and a review of potential applications; and (4) evaluation of the economic viability of specific uses for recycled CRT glass. If academia and industry can solve these problems (for example by creating a database of composition organized by manufacturer and glass source) then the reuse of CRT glass can be increased.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  19. 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...... system; this glass though has a long lifetime before ending up in the waste. Altogether these product types add up to 82% of the production of the European glass industry (IPCC, 2001). Recycling of glass in terms of cleaning and refilling of bottles as well as the use of broken glass in the production...... of new glass containers is well established in the glass industry. This chapter describes briefly howglass is produced and howwaste glass is recycled in the industry. Quality requirements and use of recycled products are discussed, as are the resource and environmental issues of glass recycling....

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

    Energy Technology Data Exchange (ETDEWEB)

    Sevougian, S. David [Advanced Nuclear Energy Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); MacKinnon, Robert J. [Advanced Nuclear Energy Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Leigh, Christi D. [Defense Waste Management Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Hansen, Frank D. [Geoscience Research and Applications Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States)

    2013-07-01

    The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential

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

    International Nuclear Information System (INIS)

    Sevougian, S. David; MacKinnon, Robert J.; Leigh, Christi D.; Hansen, Frank D.

    2013-01-01

    The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential

  2. Glass Transition, Crystallization of Glass-Forming Melts, and Entropy

    Directory of Open Access Journals (Sweden)

    Jürn W. P. Schmelzer

    2018-02-01

    Full Text Available A critical analysis of possible (including some newly proposed definitions of the vitreous state and the glass transition is performed and an overview of kinetic criteria of vitrification is presented. On the basis of these results, recent controversial discussions on the possible values of the residual entropy of glasses are reviewed. Our conclusion is that the treatment of vitrification as a process of continuously breaking ergodicity with entropy loss and a residual entropy tending to zero in the limit of zero absolute temperature is in disagreement with the absolute majority of experimental and theoretical investigations of this process and the nature of the vitreous state. This conclusion is illustrated by model computations. In addition to the main conclusion derived from these computations, they are employed as a test for several suggestions concerning the behavior of thermodynamic coefficients in the glass transition range. Further, a brief review is given on possible ways of resolving the Kauzmann paradox and its implications with respect to the validity of the third law of thermodynamics. It is shown that neither in its primary formulations nor in its consequences does the Kauzmann paradox result in contradictions with any basic laws of nature. Such contradictions are excluded by either crystallization (not associated with a pseudospinodal as suggested by Kauzmann or a conventional (and not an ideal glass transition. Some further so far widely unexplored directions of research on the interplay between crystallization and glass transition are anticipated, in which entropy may play—beyond the topics widely discussed and reviewed here—a major role.

  3. Spin glasses

    International Nuclear Information System (INIS)

    Mookerjee, Abhijit

    1976-01-01

    ''Spin glasses'', are entire class of magnetic alloys of moderate dilution, in which the magnetic atoms are far enough apart to be unlike the pure metal, but close enough so that the indirect exchange energy between them (mediated by the s-d interaction between local moments and conduction electrons) dominates all other energies. Characteristic critical phenomena displayed such as freezing of spin orientation at 'Tsub(c)' and spreading of magnetic ordering, are pointed out. Anomalous behaviour, associated with these critical phenomena, as reflected in : (i) Moessbauer spectroscopy giving hyperfine splitting at Tsub(c), (ii) maxima in susceptibility and remanent magnetism, (iii) thermopower maxima and change in slope, (iv) Characteristic cusp in susceptibility and its removal by very small magnetic fields, and (v) conductivity-resistivity measurements, are discussed. Theoretical developments aimed at explaining these phenomena, in particular, the ideas from percolation and localisation theories, and the approach based on the gellations of polymers, are discussed. Finally, a new approach based on renormalisation group in disordered systems is also briefly mentioned. (K.B.)

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

    International Nuclear Information System (INIS)

    Pritrsky, J.

    2008-02-01

    Methodology for the long-term safety assessment of nuclear fuel cycles is given in the presented doctoral thesis. The aim of work was to develop a geological repository model for disposal of spent nuclear fuel (SNF) and high level waste (HLW) using an appropriate software code able to calculate the influence of partitioning and transmutation in advanced fuel cycles. The first step in this process was specifying of indicators which can be used to quantify the radiological impact of each fuel cycle. Indicators such as annual effective dose and radiotoxicity of inventory have been quantitatively analysed to determine the potential risk and radiological consequences associated with production of SNF/HLW. Advanced fuel types bring a number of advantages in comparison to uranium oxide fuel UO 2 used worldwide nowadays in terms of safety improvement due to minor actinides transmutation and non-proliferation aspects as well. Within the scope of work, three different fuel cycles are compared from the point of view of long-term safety of deep geological repository. The first considered fuel cycle is the currently used open fuel cycle (UOX) which uses only U-FA (Uranium Fuel Assembly). The second assessed cycle is a closed fuel cycle (MOX) with MOX-FA (Mixed OXides Fuel Assembly) and the third considered one is a partially closed fuel cycle (IMF) with IMC-FA (Inert Matrix Combined Fuel Assembly). Description and input data of advanced fuel cycles have been gained by participation in the EC project RED-IMPACT. Results were calculated using code AMBER, which is a flexible software tool that allows building dynamic compartmental models to represent the migration and fate of contaminants in a system, for example in the surface and sub-surface environment. Contaminants in solid, liquid and gaseous phases can be considered. AMBER gives the user the flexibility to define any number of compartments; any number of contaminants and associated decays; deterministic, probabilistic and

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

    International Nuclear Information System (INIS)

    Pritrsky, J.

    2008-01-01

    Methodology for the long-term safety assessment of nuclear fuel cycles is given in the presented doctoral thesis. The aim of work was to develop a geological repository model for disposal of spent nuclear fuel (SNF) and high level waste (HLW) using an appropriate software code able to calculate the influence of partitioning and transmutation in advanced fuel cycles. The first step in this process was specifying of indicators which can be used to quantify the radiological impact of each fuel cycle. Indicators such as annual effective dose and radiotoxicity of inventory have been quantitatively analysed to determine the potential risk and radiological consequences associated with production of SNF/HLW. Advanced fuel types bring a number of advantages in comparison to uranium oxide fuel UO 2 used worldwide nowadays in terms of safety improvement due to minor actinides transmutation and non-proliferation aspects as well. Within the scope of work, three different fuel cycles are compared from the point of view of long-term safety of deep geological repository. The first considered fuel cycle is the currently used open fuel cycle (UOX) which uses only U-FA (Uranium Fuel Assembly). The second assessed cycle is a closed fuel cycle (MOX) with MOX-FA (Mixed OXides Fuel Assembly) and the third considered one is a partially closed fuel cycle (IMF) with IMC-FA (Inert Matrix Combined Fuel Assembly). Description and input data of advanced fuel cycles have been gained by participation in the EC project RED-IMPACT. Results were calculated using code AMBER, which is a flexible software tool that allows building dynamic compartmental models to represent the migration and fate of contaminants in a system, for example in the surface and sub-surface environment. Contaminants in solid, liquid and gaseous phases can be considered. AMBER gives the user the flexibility to define any number of compartments; any number of contaminants and associated decays; deterministic, probabilistic and

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

    International Nuclear Information System (INIS)

    Barney, G.S.

    1996-01-01

    At the Hanford Site organic compounds were measured in tank supernate simulant solutions during FY 1995. This solubility information will be used to determine if these organic salts could exist in solid phases (saltcake or sludges) in the waste where they might react violently with the nitrate or nitrite salts present in the tanks. Solubilities of sodium glycolate, succinate, and caproate salts; iron and aluminum and butylphosphate salts; and aluminum oxalate were measured in simulated waste supernate solutions at 25 degree C, 30 degree C, 40 degree C, and 50 degree C. The organic compounds were selected because they are expected to exist in relatively high concentrations in the tanks. The solubilities of sodium glycolate, succinate, caproate, and butylphosphate in HLW tank supernate solutions were high over the temperature and sodium hydroxide concentration ranges expected in the tanks. High solubilities will prevent solid sodium salts of these organic acids from precipitating from tank supernate solutions. The total organic carbon concentrations (YOC) of actual tank supernates are generally much lower than the TOC ranges for simulated supernate solutions saturated (at the solubility limit) with the organic salts. This is so even if all the dissolved carbon in a given tank and supernate is due to only one of these eight soluble compounds (an unlikely situation). Metal ion complexes of and butylphosphate and oxalate in supernate solutions were not stable in the presence of the hydroxide concentrations expected in most tanks. Iron and aluminum dibutylphosphate compounds reacted with hydroxide to form soluble sodium dibutylphosphate and precipitated iron and aluminum hydroxides. Aluminum oxalate complexes were also not stable in the basic simulated supernate solutions. Solubilities of all the organic salts decrease with increasing sodium hydroxide concentration because of the common ion effect of Na+. Increasing temperatures raised the solubilities of the organic

  7. Hydro-mechanical behaviour of crushed COx argillite used as backfilling material in HLW repository

    International Nuclear Information System (INIS)

    Tang Chaosheng; Shi Bin; Cui Yujun; Anh-Minh Tang

    2010-01-01

    At present, the crushed Callovo-Oxfordian (COx) argillite powder is proposed as an alternative backfilling material in France, which will be constructed in the engineering barrier of high-level radioactive waste (HLW) repository. In this investigation, the compression behavior of two crushed COx argillite powders (coarser one and finer one) was studied by running l-D compression tests with several loading-unloading cycles. After the final dry density 2.0 g/cm 3 was reached, the specimen was flooding with distilled water and the evolution of axial stress was studied during saturation process. The effects of initial axial stress level and grain size distribution (GSD) on hydro-mechanical behaviour of compacted specimen were analyzed. The results show that the compression curves are significantly influenced by the GSD of the soils. To obtain the same degree of compaction, the axial stress applied to finer soil is much higher than that of coarser soil. In addition, the compression index of the finer soil is bigger than that of coarser soil. The swelling index at initial water content increases with the dry density and seems to be independent of the GSD. During saturation, the initial lower axial stress causes obvious swelling behavior for both the coarser and finer powder samples and the corresponding axial stress increase gradually. At initial higher axial stress condition, monotone collapse behavior is observed for the coarser powder samples. Whereas the axial stress decrease firstly, then increase and finally decrease again for the finer powder samples. After saturation, the equilibrium axial stresses of finer powder samples are higher than that of coarser powder samples. (authors)

  8. Dissolution of ORNL HLW sludge and partitioning of the actinides using the TRUEX process

    International Nuclear Information System (INIS)

    Spencer, B.B.; Egan, B.Z.; Beahm, E.C.; Chase, C.W.; Dillow, T.A.

    1997-01-01

    Experiments were conducted to evaluate the transuranium extraction (TRUEX) process for partitioning actinides from actual dissolved high-level radioactive waste (HLW) sludge. Samples of sludge from melton Valley Storage Tank W-25 were rinsed with mild caustic (0.2 M NaOH) to reduce the concentrations of nitrates and fission products associated with the interstitial liquid. In one campaign the rinsed sludge was leached in nitric acid, and about 50% of the dry mass of the sludge was dissolved. The resulting solution contained total metal concentrations of ∼ 1.8 M with a nitric acid concentration of 2.9 M. In the other campaign the sludge was neutralized with nitric acid to destroy the carbonates, then leached with 2.6 M NaOH for ∼ 6 h before rinsing with the mild caustic. The sludge was then leached in nitric acid, and about 80% of the sludge dissolved. The resulting solution contained total metal concentrations of ∼ 0.6 M with a nitric acid concentration of 1.7 M. Chemical analyses of both phases were used to evaluate the process. Evaluation was based on two metrics: the fraction of TRU elements removed from the dissolved sludge and comparison of the results with predictions made with the Generic TRUEX Model (GTM). The fractions of Eu, Pu, Cm, Th and U species removed from aqueous solution in only one extraction stage were > 95% and were close to the values predicted by the GTM. Mercury was also found to be strongly extracted, with a one-stage removal of > 92%. In one test, vanadium appeared to be moderately extracted

  9. Time-frames and the demonstration of safety for HLW disposal

    International Nuclear Information System (INIS)

    Watkins, B.; Kessler, J.

    1999-01-01

    An important principle which is often embodied in the criteria for the safe disposal of long-lived radioactive wastes is that a similar level of radiation protection should be provided to future generations as that provided for those alive today. This has resulted in the development of performance assessment methodologies to evaluate the potential long term impacts of HLW disposal on humans, usually in terms of individual dose or risk. However, the actual periods of time over which it is expected that there will be full control over high level waste disposals are extremely short in comparison with the times over which radionuclides in the wastes could potentially move from the deep repository and emerge into the surface environment. This leads to problems in setting quantitative dose or risk based standard appropriate for the short and long term, and in setting the time-frames for which calculations should be carried out. This is especially difficult in view of the uncertainty in predicting changes in human behaviour and changes in the biosphere and geosphere over the time-scales involved. Different assessment time-frames and approaches proposed by IAEA, Nordic countries, Britain and US guidance documents are briefly reviewed. Whilst accepting the basic radiation protection objective of protecting future generations, no international consensus bas been agreed on what time-frames should be used in performance assessments. It is recommended that different time-frames should be associated with different quantitative or qualitative performance measures. As a result, a range of indicators of safety may be appropriate in demonstrating compliance with regulatory performance criteria and the consequent overall assessment context. It is argued that what is required is a simple, robust yet defensible approach to time-frames and performance indicators which can be accepted by the public, regulators and the nuclear industry

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

    International Nuclear Information System (INIS)

    J. Bisset

    2005-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mitsui, Takashi; Miura, Noriaki [IHI Corporation, 1 Shin-Nakahara-cho, Isogo-ku, Yokohama, Kanagawa (Japan); Oowaki, Katsura; Kawaguchi, Isao [IHI Inspection and Instrumentation Co., Ltd, 1 Shin-Nakahara-cho, Isogo-ku, Yokohama, Kanagawa (Japan); Miura, Yasuhiko; Ino, Tooru [Japan Nuclear Fuel Limited, 4-108, Aza Okitsuke, Oaza Obuchi, Rokkasho-Mura, Kamikita-gun, Aomori (Japan)

    2013-07-01

    Since 2008, the equipment for dismantling the used glass melter has been developed in High-level Liquid Waste (HLW) Vitrification Facility in the Japanese Rokkasho Reprocessing Plant (RRP). Due to the high radioactivity of the glass melter, the equipment requires a fully-remote operation in the vitrification cell. The remote fiber laser cutting system was adopted as one of the major pieces of equipment. An output power of fiber laser is typically higher than other types of laser and so can provide high-cutting performance. The fiber laser can cut thick stainless steel and Inconel, which are parts of the glass melter such as casings, electrodes and nozzles. As a result, it can make the whole of the dismantling work efficiently done for a shorter period. Various conditions of the cutting test have been evaluated in the process of developing the remote fiber cutting system. In addition, the expected remote operations of the power manipulator with the laser torch have been fully verified and optimized using 3D simulations. (authors)

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  13. Audits of radiopharmaceutical formulations

    International Nuclear Information System (INIS)

    Castronovo, F.P. Jr.

    1992-01-01

    A procedure for auditing radiopharmaceutical formulations is described. To meet FDA guidelines regarding the quality of radiopharmaceuticals, institutional radioactive drug research committees perform audits when such drugs are formulated away from an institutional pharmacy. All principal investigators who formulate drugs outside institutional pharmacies must pass these audits before they can obtain a radiopharmaceutical investigation permit. The audit team meets with the individual who performs the formulation at the site of drug preparation to verify that drug formulations meet identity, strength, quality, and purity standards; are uniform and reproducible; and are sterile and pyrogen free. This team must contain an expert knowledgeable in the preparation of radioactive drugs; a radiopharmacist is the most qualified person for this role. Problems that have been identified by audits include lack of sterility and apyrogenicity testing, formulations that are open to the laboratory environment, failure to use pharmaceutical-grade chemicals, inadequate quality control methods or records, inadequate training of the person preparing the drug, and improper unit dose preparation. Investigational radiopharmaceutical formulations, including nonradiolabeled drugs, must be audited before they are administered to humans. A properly trained pharmacist should be a member of the audit team

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  15. Reactive decontamination formulation

    Science.gov (United States)

    Giletto, Anthony [College Station, TX; White, William [College Station, TX; Cisar, Alan J [Cypress, TX; Hitchens, G Duncan [Bryan, TX; Fyffe, James [Bryan, TX

    2003-05-27

    The present invention provides a universal decontamination formulation and method for detoxifying chemical warfare agents (CWA's) and biological warfare agents (BWA's) without producing any toxic by-products, as well as, decontaminating surfaces that have come into contact with these agents. The formulation includes a sorbent material or gel, a peroxide source, a peroxide activator, and a compound containing a mixture of KHSO.sub.5, KHSO.sub.4 and K.sub.2 SO.sub.4. The formulation is self-decontaminating and once dried can easily be wiped from the surface being decontaminated. A method for decontaminating a surface exposed to chemical or biological agents is also disclosed.

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

    International Nuclear Information System (INIS)

    Marvy, A.; Ochem, D.

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    Polyakov, Yu.D.; Porsov, A.Yu.; Beigul, V.P.; Palenov, M.V.

    2014-01-01

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

  18. Linear free energy relationships in glass corrosion

    International Nuclear Information System (INIS)

    Abrajano, T.A. Jr.; Bates, J.K.; Bohlke, J.K.

    1988-01-01

    Various theoretical models that have been proposed to correlate glass durability to their composition for a wide variety of silicate, borosilicate, and aluminosilicate glasses are examined. Comparisons are made between the predictions of these models and those of an empirical formulation extracted from existing data in the present work. The empirical approach provides independent confirmation of the relative accuracy of the silica release rate predictions of the different theoretical models in static leaching systems. Extension of the empirical approach used in this work are discussed. 23 refs., 2 figs., 1 tab

  19. Suggestions on selection of clay site as a key alternative of underground repository for HLW geological disposal in China

    International Nuclear Information System (INIS)

    Zheng Hualing; Fu Bingjun; Fan Xianhua; Chen Shi; Sun Donghui

    2006-01-01

    Site selection for the underground repository is a vital problem with respect to the HLW geological disposal. Over the past decades, we have been focusing our attention on granite as a priority in China. However, there are some problems have to be discussed on this matter. In this paper, both experiences gained and lessons learned in the international community regarding the site selection are described. And then, after analyzing a lot of some key factors affecting the site selection, some comments and suggestions on selection of clay site as a key alternative before final decision making in China are presented. (authors)

  20. Natural analogue of redox front formation in near-field environment at post-closure phase of HLW geological disposal

    International Nuclear Information System (INIS)

    Yoshida, Hidekazu; Yamamoto, Koushi; Amano, Yuki

    2005-01-01

    Redox fronts are created in the near field of rocks, in a range of oxidation environments, by microbial activity in rock groundwater. Such fronts, and the associated oxide formation, are usually unavoidable around high level radioactive waste (HLW) repositories, whatever their design. The long term behaviour of these oxides after repositories have been closed is however little known. Here we introduce an analogue of redox front formation, such as 'iron oxide' deposits, known as takashikozo forming cylindrical nodules, and the long term behaviour of secondarily formed iron oxyhydroxide in subsequent geological environments. (author)

  1. Status of Progress Made Toward Safety Analysis and Technical Site Evaluations for DOE Managed HLW and SNF.

    Energy Technology Data Exchange (ETDEWEB)

    Sevougian, S. David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Stein, Emily [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gross, Michael B [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hammond, Glenn Edward [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Frederick, Jennifer M [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mariner, Paul [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-11-01

    The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE) is conducting research and development (R&D) on generic deep geologic disposal systems (i.e., repositories). This report describes specific activities in FY 2016 associated with the development of a Defense Waste Repository (DWR)a for the permanent disposal of a portion of the HLW and SNF derived from national defense and research and development (R&D) activities of the DOE.

  2. 6Li-doped silicate glass for thermal neutron shielding

    International Nuclear Information System (INIS)

    Stone, C.A.; Blackburn, D.H.; Kauffman, D.A.; Cranmer, D.C.; Olmez, I.

    1994-01-01

    Glass formulations are described that contain high concentrations of 6 Li and are suitable for use as thermal neutron shielding. One formulation contained 31 mol% of 6 Li 2 O and 69 mol% of SiO 2 . Studies were performed on a second formulation that contained as much as 37 mol% of 6 Li 2 O and 59 mol% of SiO 2 , with 4 mol% Al 2 O 3 added to prevent crystallization at such high 6 Li 2 O concentrations. These lithium silicate glasses can be formed into a variety of shapes using conventional glass fabrication techniques. Examples include flat plates, disks, hollow cylinders, and other more complex geometries. Both in-beam and in-core experiments have been performed to study the use and durability of Li silicate glasses. In-core experiments show the glass can withstand the intense radiation fields near the core of a reactor. The neutron attenuation of the glasses used in these studies was 90%/mm. In-beam studies show that the glass is effective for reducing the gamma-ray and neutron fields near experiments. ((orig.))

  3. 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.

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

    International Nuclear Information System (INIS)

    Moore, P.; Bergman, W.; Gilbert, H.

    1992-08-01

    We have conducted a survey of the major users of glass-paper HEPA filters in the DOE complex to ascertain the life cycle costs of these filters. Purchase price of the filters is only a minor portion of the costs; the major expenditures are incurred during the removal and disposal of contaminated filters. Through personal interviews, site visits and completion of questionnaires, we have determined the costs associated with the use of HEPA filters in the DOE complex. The total approximate life-cycle cost for a standard (2 in. x 2 in. x 1 in.) glass-paper HEPA filter is $3,000 for one considered low-level waste (LLW), $11,780 for transuranic (TRU) and $15,000 for high-level waste (HLW). The weighted-average cost for a standard HEPA filter in the complex is $4,753

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

    International Nuclear Information System (INIS)

    Loiseau, P.; Caurant, D.; Touet, I.; Destre, Y.; Fillet, C.

    2000-01-01

    December 1991 legislation in France has spurred research on enhanced separation and conditioning or transmutation of long-lived radionuclides from high level radioactive wastes (HLW). In this field, we have studied zirconolite-based glass-ceramics in which the crystalline phase (zirconolite: CaZrTi 2 O 7 ) aimed to preferentially incorporate minor actinides is embedded in a glassy calcium aluminosilicate matrix. At the laboratory scale, the crystallization of the parent glass is carried out thanks to a two-step thermal treatment: a nucleation stage followed by a growth stage. This paper presents the evolution of the crystallization, followed by scanning electron microscopy (SEM) and X-ray diffraction (XRD), with the temperature of the crystal growth thermal treatment, in the range 950 deg. C - 1350 deg. C. (authors)

  6. Preparation of radiopharmaceutical formulations

    International Nuclear Information System (INIS)

    Simon, J.; Garlich, J.R.; Frank, R.K.; McMillan, K.

    1998-01-01

    Radiopharmaceutical formulations for complexes comprising at least one radionuclide complexed with a ligand, or its physiologically-acceptable salts thereof, especially 153 samarium-ethylenediaminetetramethylenephosphonic acid, which optionally contains a divalent metal ion, e.g. calcium, and is frozen, thawed, and then administered by injection. Alternatively, the radiopharmaceutical formulations must contain the divalent metal and are frozen only if the time before administration is sufficiently long to cause concern for radiolysis of the ligand. 2 figs., 9 tabs

  7. Tariff formulation and equalization

    International Nuclear Information System (INIS)

    Svartsund, Trond

    2003-01-01

    The primary goal of the transmission tariff is to provide for socioeconomic use of the transmission grid. The present tariff structure is basically right. The responsibility for the formulation of the tariff resides with the local grid owner. This must take place in agreement with the current regulations which are passed by the authorities. The formulation must be adaptable to the local requirements. EBL (Norwegian Electricity Industry Association) is content with the current regulations

  8. Glass and nuclear wastes

    International Nuclear Information System (INIS)

    Sombret, C.

    1982-10-01

    Glass shows interesting technical and economical properties for long term storage of solidified radioactive wastes by vitrification or embedding. Glass composition, vitrification processes, stability under irradiation, thermal stability and aqueous corrosion are studied [fr

  9. 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.

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

    Science.gov (United States)

    Bansal, Narottam P.; Choi, Sung R.

    2007-01-01

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

  11. Conceptual design for vitrification of HLW at West Valley using a rotary calciner/metallic melter

    International Nuclear Information System (INIS)

    Giraud, J.P.; Conord, J.P.; Saverot, P.M.

    1984-01-01

    The CEA has had an extensive research program in the field of vitrification technology for over 24 years, and several testing facilities were used throughout all phases of development and engineering: The Vulcain facility comprises a vitrification hot cell and four auxiliary hot cells. Vulcain allows the production of 2-kg samples of active glass. The off-gas treatment system allows testing the DF of each equipment. The auxiliary cells are equipped with leach-rate tests, diffusion tests, and irradiation tests on the glass samples. The Atlas facility is a reproduction of AVM calcination and vitrification furnaces at 1/2 scale enclosed in a glove box. This facility is used for testing ruthenium volatility and containment in the vitrification process. The full-scale AVM inactive pilot facility is used for testing calcination and vitrification of new compositions of high-level waste and for developing new types of vitrification furnaces. The inactive test loop is for testing air cooling of glass containers. The full-scale AVH inactive pilot facility is used for testing AVH technology and has been in operation since late 1981

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

    International Nuclear Information System (INIS)

    Put, M.; Henrion, P.

    1992-01-01

    For the modelling of the migration of radionuclides in the Boom clay formation, the analytical code MICOF has been updated with a 3-dimensional analytical solution for discrete sources. the MICOF program is used for the calculation of the release of α and β emitters from the HIGH LEVEL RADIOACTIVE WASTES (HLW). A coherent conceptual model is developed which describes all the major physico-chemical phenomena influencing the migration of radionuclides in the Boom clay. The concept of the diffusion accessible porosity is introduced and included in the MICOF code. Different types of migration experiments are described with their advantages and disadvantages. The thermal impact of the HLW disposal in the stratified Boom clay formation has been evaluated by a finite element simulation of the coupled heat and mass transport equation. The results of the simulations show that under certain conditions thermal convection cells may form, but the convective heat transfer in the clay formation is negligible. 6 refs., 19 figs., 2 tabs., 5 appendices

  13. A methodology of uncertainty/sensitivity analysis for PA of HLW repository learned from 1996 WIPP performance assessment

    International Nuclear Information System (INIS)

    Lee, Y. M.; Kim, S. K.; Hwang, Y. S.; Kang, C. H.

    2002-01-01

    The WIPP (Waste Isolation Pilot Plant) is a mined repository constructed by the US DOE for the permanent disposal of transuranic (TRU) wastes generated by activities related to defence of the US since 1970. Its historical disposal operation began in March 1999 following receipt of a final permit from the State of NM after a positive certification decision for the WIPP was issued by the EPA in 1998, as the first licensed facility in the US for the deep geologic disposal of radioactive wastes. The CCA (Compliance Certification Application) for the WIPP that the DOE submitted to the EPA in 1966 was supported by an extensive Performance Assessment (PA) carried out by Sandia National Laboratories (SNL), with so-called 1996 PA. Even though such PA methodologies could be greatly different from the way we consider for HLW disposal in Korea largely due to quite different geologic formations in which repository are likely to be located, a review on lots of works done through the WIPP PA studies could be the most important lessons that we can learn from in view of current situation in Korea where an initial phase of conceptual studies on HLW disposal has been just started. The objective of this work is an overview of the methodology used in the recent WIPP PA to support the US DOE WIPP CCA ans a proposal for Korean case

  14. Measurement of optical glasses

    International Nuclear Information System (INIS)

    Nicolau-Rebigan, S.

    1978-11-01

    The possibilities of measurement of the optical glasses parameters needed in building optical devices especially in lasers devices are presented. In the first chapter the general features of the main optical glasses as well as the modalities of obtaining them are given. Chapter two defines the optical glass parameters, and the third chapter describes the measuring methods of the optical glass parameters. Finally, the conclusions which point out the utilization of this paper are presented. (author)

  15. Thermo-mechanical analysis for multi-level HLW repository concept

    International Nuclear Information System (INIS)

    Kwon, Sang Ki; Choi, Jong Won

    2004-01-01

    This work aims to investigate the influence of design parameters for the underground high-level nuclear waste repository with multi-level concept. B. Necessity o In order to construct an HLW repository in deep underground, it is required to select a site, which is far from major discontinuities. To dispose the whole spent fuels generated from the Korean nuclear power plants in a repository, the underground area of about 4km 2 is required. This would be a constraints for selecting an adequate repository site. It is recommended to dispose the two different spent fuels, PWR and CANDU, in different areas at the operation efficiency point of view. It is necessary to investigate the influence of parameters, which can affect the stability of multi-level repository. It is also needed to consider the influence of heat generated from the HLW and the high in situ stress in deep location. Therefore, thermo-mechanical coupling analysis should be carried out and the results should be compared with the results from single-level repository concept. Three-dimensional analysis is required to model the disposal tunnel and deposition hole. It is recommended to use the Korean geological condition and actually measured rock properties in Korea in order to achieve reliable modeling results. A FISH routine developed for effective modeling of Thermal-Mechanical coupling was implemented in the modeling using FLAC3D, which is a commercial three-dimensional FDM code. The thermal and mechanical properties of rock and rock mass achieved from Yusung drilling site, were used for the computer modeling. Different parameters such as level distance, waste type disposed on different levels, and time interval between the operation on different levels, were considered in the three-dimensional analysis. From the analysis, it was possible to derive adequate multi-level repository concept. Results and recommendations for application From the thermal-mechanical analysis for the multi-level repository

  16. Mechanically reinforced glass beams

    DEFF Research Database (Denmark)

    Nielsen, Jens Henrik; Olesen, John Forbes

    2007-01-01

    laminated float glass beam is constructed and tested in four-point bending. The beam consist of 4 layers of glass laminated together with a slack steel band glued onto the bottom face of the beam. The glass parts of the tested beams are \\SI{1700}{mm} long and \\SI{100}{mm} high, and the total width of one...

  17. NUCLEAR WASTE GLASSES: CONTINUOUS MELTING AND BULK VITRIFICAITON

    International Nuclear Information System (INIS)

    KRUGER, A.A.

    2008-01-01

    This contribution addresses various aspects of nuclear waste vitrification. Nuclear wastes have a variety of components and composition ranges. For each waste composition, the glass must be formulated to possess acceptable processing and product behavior defined in terms of physical and chemical properties that guarantee the glass can be easily made and resist environmental degradation. Glass formulation is facilitated by developing property-composition models, and the strategy of model development and application is reviewed. However, the large variability of waste compositions presents numerous additional challenges: insoluble solids and molten salts may segregate; foam may hinder heat transfer and slow down the process; molten salts may accumulate in container refractory walls; the glass on cooling may precipitate crystalline phases. These problems need targeted exploratory research. Examples of specific problems and their possible solutions are discussed

  18. The influence of particle size and fluorine content of aluminosilicate glass on the glass ionomer cement properties.

    Science.gov (United States)

    De Caluwé, T; Vercruysse, C W J; Fraeyman, S; Verbeeck, R M H

    2014-09-01

    Glass ionomer cements (GIC) are clinically accepted dental restorative materials mainly due to their direct chemical adhesion to both enamel and dentin and their ability to release fluoride. However, their mechanical properties are inferior compared to those of amalgam and composite. The aim of this study is to investigate if combinations of nano- and macrogranular glass with different compositions in a glass ionomer cement can improve the mechanical and physical properties. Glasses with the composition 4.5 SiO2-3 Al2O3-1.5 P2O5-(5-x) CaO-x CaF2 (x=0 and x=2) were prepared. Of each type of glass, particles with a median size of about 0.73 μm and 6.02 μm were made. The results show that the setting time of GIC decreases when macrogranular glass particles are replaced by nanogranular glass particles, whereas the compressive strength and Young's modulus, measured after 24 h setting, increase. The effects are more pronounced when the nanogranular glass particles contain fluoride. After thermocycling, compressive strength decreases for nearly all formulations, the effect being most pronounced for cements containing nanogranular glass particles. Hence, the strength of the GIC seems mainly determined by the macrogranular glass particles. Cumulative F--release decreases when the macrogranular glass particles with fluoride are replaced by nanogranular glass particles with(out) fluoride. The present study thus shows that replacing macro- by nanogranular glass particles with different compositions can lead to cements with approximately the same physical properties (e.g. setting time, consistency), but with different physicochemical (e.g. F--release, water-uptake) and initial mechanical properties. On the long term, the mechanical properties are mainly determined by the macrogranular glass particles. Copyright © 2014 Academy of Dental Materials. Published by Elsevier Ltd. All rights reserved.

  19. Formulation of Japanese consensus-building model for HLW geological disposal site determination. 2. Development of game for enhancement of social acceptance

    International Nuclear Information System (INIS)

    Okuno, Kenji; Oya, Yasuhisa

    2010-01-01

    Present project consists of two parts, model for institution design and education program. For the education program for public citizen, we have developed a game running on iPod touch based on Erehwon repository project proposed by Swiss ITC under OBRA project. (author)

  20. Development of antimicrobial optimum glass ionomer

    International Nuclear Information System (INIS)

    Angioletto, E.; Tezza, V.B.; Santos, M.J.; Montedo, O.R.K.; Pich, C.T.; Fiori, M.A.; Angioletto, Ev.

    2010-01-01

    The use of glass ionomer for restorations in dentistry for lower income population is a well established practice in public clinics of Brazil. However the average price of this kind of material and its low durability still have a negative impact on public health for being imported and frequently replaced it becomes expensive for the manufacturers and for public agencies. In glass ionomer the main antimicrobial agent is fluoride, which is released gradually. The material used for filling provides an average life of five years and its durability can be increased if the ionomer contains other oligodynamic elements. It was formulated, merged a new optimized glass ionomer which was characterized by X-ray diffraction, ion measurement and antimicrobial activity. This new product showed promising results, that pointed structural stability an increase of antimicrobial efficiency. (author)

  1. Granulated decontamination formulations

    Science.gov (United States)

    Tucker, Mark D.

    2007-10-02

    A decontamination formulation and method of making that neutralizes the adverse health effects of both chemical and biological compounds, especially chemical warfare (CW) and biological warfare (BW) agents, and toxic industrial chemicals. The formulation provides solubilizing compounds that serve to effectively render the chemical and biological compounds, particularly CW and BW compounds, susceptible to attack, and at least one reactive compound that serves to attack (and detoxify or kill) the compound. The formulation includes at least one solubilizing agent, a reactive compound, a sorbent additive, and water. A highly adsorbent sorbent additive (e.g., amorphous silica, sorbitol, mannitol, etc.) is used to "dry out" one or more liquid ingredients into a dry, free-flowing powder that has an extended shelf life, and is more convenient to handle and mix in the field.

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

    International Nuclear Information System (INIS)

    Kamizono, H.

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Yalmali, Vrunda S.; Singh, I.J.; Sathi Sasidharan, N.; Deshingkar, D.S.

    2004-11-01

    Long half life and easy availability from high level wastes make 137 Cesium most economical radiation source. High level liquid waste processing for 137 Cesium removal has become easier due to development of Cesium specific granulated ammonium molybdophosphate (AMP) composite. In such applications, resulting spent composite AMP itself represents high active solid waste and immobilization of these materials in cement may not be acceptable. Studies on immobilization of 137 Cs loaded AMP were taken up in order to achieve twin goals of increasing safety and minimizing processing costs of the final matrix. Studies indicated that phosphate modified sodium borosilicate SPNM glasses prepared under usual oxidizing conditions are not suitable for immobilization of 137 Cs loaded on AMP .Phosphate glasses containing Na 2 O, P 2 O 5 , B 2 O 3 , Fe 2 O 3 , Al 2 O 3 and SiO 2 as major constituents are capable of incorporating 6 to 8 % AMP. The Normalized Leach rates of these glasses for sodium, cesium, boron and silica are 10 -4 to 10 -6 gm/cm 2 /day which are comparable to or better than those reported for NBS glasses incorporating HLW. Homogeneity of the final matrix was confirmed by x-ray diffraction analysis. Further studies on characterization of these glasses would establish their acceptability. (author)

  4. 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

  5. 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...

  6. 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...... of different glass products can be quantified and ranked. This approach is based on determination of both the optical intensity and dimension of the striations in glasses. These two characteristic values areobtained using the image processing method established recently. The logarithmic ratio between...

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  8. 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

  9. Leaching of glass

    International Nuclear Information System (INIS)

    Hench, L.L.

    1977-01-01

    Understanding surface compositional profiles of glasses over a range of 0-2000 A with a variety of analytical instruments shows that five general types of glass surfaces exist. The surface character of a glass article depends upon bulk composition and environmental history during which surface dealkalization, film formation, and network dissolution can occur. Environmental-surface interactions generally result in complex compositional profiles of all the constituents in a glass. Durable glasses almost always develop a stable surface film which has a higher concentration of network formers than the bulk composition. Compositional effects that are used to improve glass durability usually improve the stability of the surface films. Durability tests or service conditions that lead to film destruction are especially severe for the most silicate glasses. 43 references

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

    International Nuclear Information System (INIS)

    Yi, Yun-Bo; Amme, Robert C.; Shayer, Zeev

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    Stefanko, D.; Herbert, J.

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  13. 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

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

    International Nuclear Information System (INIS)

    Kawanishi, M.; Komada, H.; Kitayama, K.; Akasaka, H.; Tsuchi, H.

    2001-01-01

    In Japan, the high-level radioactive waste (HLW) disposal project is expected to start fully after establishment of the implementing organization, which is planned around the year 2000 and to dispose the wastes in the 2030s to at latest in the middle of 2040s. Considering each step in the implementation of the HLW disposal project in Japan, this paper discusses the execution procedure for HLW disposal project, such as the selection of candidate/planned disposal sites, the construction and operation of the disposal facility, the closure and decommissioning of facilities, and the institutional control and monitoring after the closure of disposal facility, from a technical viewpoint for the rational execution of the project. Furthermore, we investigate and propose some ideas for the concept of the design of geological disposal facility, the validation and demonstration of the reliability on the disposal techniques and performance assessment methods at a candidate/planned site. Based on these investigation results, we made clear a milestone for the execution of the HLW disposal project in Japan. (author)

  15. Study of the glass formation of high temperature superconductors

    Science.gov (United States)

    Ethridge, Edwin C.; Kaukler, William F.; Rolin, Terry

    1992-01-01

    A number of compositions of ceramic oxide high T(sub c) superconductors were elevated for their glass formation ability by means of rapid thermal analysis during quenching, optical, and electron microscopy of the quenched samples, and with subsequent DSC measurements. Correlations between experimental measurements and the methodical composition changes identified the formulations of superconductors that can easily form glass. The superconducting material was first formed as a glass; then, with subsequent devitrification, it was formed into a bulk crystalline superconductor by a series of processing methods.

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  18. A Review of Glass-Ionomer Cements for Clinical Dentistry

    Directory of Open Access Journals (Sweden)

    Sharanbir K. Sidhu

    2016-06-01

    Full Text Available This article is an updated review of the published literature on glass-ionomer cements and covers their structure, properties and clinical uses within dentistry, with an emphasis on findings from the last five years or so. Glass-ionomers are shown to set by an acid-base reaction within 2–3 min and to form hard, reasonably strong materials with acceptable appearance. They release fluoride and are bioactive, so that they gradually develop a strong, durable interfacial ion-exchange layer at the interface with the tooth, which is responsible for their adhesion. Modified forms of glass-ionomers, namely resin-modified glass-ionomers and glass carbomer, are also described and their properties and applications covered. Physical properties of the resin-modified glass-ionomers are shown to be good, and comparable with those of conventional glass-ionomers, but biocompatibility is somewhat compromised by the presence of the resin component, 2 hydroxyethyl methacrylate. Properties of glass carbomer appear to be slightly inferior to those of the best modern conventional glass-ionomers, and there is not yet sufficient information to determine how their bioactivity compares, although they have been formulated to enhance this particular feature.

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

    International Nuclear Information System (INIS)

    Kaushik, C.P.

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Bertsch, P.M.; Zachara, J.M.

    1998-01-01

    'Cesium (137) is a major component of high level weapons waste. At Hanford, single shell tanks (SST''s) with high level wastes (HLW) have leaked supernate containing over 10 6 Ci of 137 Cs and other co-contaminants into the vadose zone. In select locations, 137 Cs has migrated further than expected from retardation experiments and performance assessment calculations. Deep 137 Cs migration has been observed beneath the SX tank farm at Hanford with REDOX wastes as the carrier causing regulatory and stakeholder concern. The causes for expedited migration are unclear. This research is investigating how the sorption chemistry of Cs on Hanford vadose zone sediments changes after contact with solutions characteristic of HLW. The central scientific hypothesis is that the high Na concentration of HLW will suppress surface-exchange reactions of Cs, except those to highly-selective frayed edge sites (FES) of the micaceous fraction. The authors further speculate that the concentrations, ion selectivity, and structural aspects of the FES will change after contact with HLW and that these changes will be manifest in the macroscopic sorption behavior of Cs. The authors believe that migration predictions of Cs can be improved substantially if such changes are understood and quantified. The research has three objectives: (1.) identify how the multi-component surface exchange behavior of Cs on Hanford sediments changes after contact with HLW simulants that span a range of relevant chemical (Na, OH, Al, K) and temperature conditions (23-80 C); (2) reconcile changes in sorption chemistry with microscopic and molecular changes in site distribution, chemistry, mineralogy, and surface structure of the micaceous fraction; (3) integrate mass-action-solution exchange measurements with changes in the structure/site distribution of the micaceous fraction to yield a multicomponent exchange model relevant to high ionic strength and hydroxide for prediction of environmental Cs sorption.'

  1. 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.

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

    International Nuclear Information System (INIS)

    Awano, Toshihiko; Kanno, Takeshi; Katsumata, Syunsuke; Kosuge, Kazuhiro

    2009-01-01

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

  3. Alternative biosphere modeling for safety assessment of HLW disposal taking account of geosphere-biosphere interface of marine environment

    International Nuclear Information System (INIS)

    Kato, Tomoko; Ishiguro, Katsuhiko; Naito, Morimasa; Ikeda, Takao; Little, Richard

    2001-03-01

    In the safety assessment of a high-level radioactive waste (HLW) disposal system, it is required to estimate radiological impacts on future human beings arising from potential radionuclide releases from a deep repository into the surface environment. In order to estimated the impacts, a biosphere model is developed by reasonably assuming radionuclide migration processes in the surface environment and relevant human lifestyles. It is important to modify the present biosphere models or to develop alternative biosphere models applying the biosphere models according to quality and quantify of the information acquired through the siting process for constructing the repository. In this study, alternative biosphere models were developed taking geosphere-biosphere interface of marine environment into account. Moreover, the flux to dose conversion factors calculated by these alternative biosphere models was compared with those by the present basic biosphere models. (author)

  4. A comparison of three methods for determining the amount of nitric acid needed to treat HLW sludge at SRS

    International Nuclear Information System (INIS)

    Siegwald, S.F.; Ferrara, D.M.

    1994-01-01

    A comparison was made of three methods for determining the amount of nitric acid which will be needed to treat a sample of high-level waste (HLW) sludge from the Savannah River Site (SRS) Tank Farm. The treatment must ensure the resulting melter feed will have the necessary rheological and oxidation-reduction properties, reduce mercury and manganese in the sludge, and be performed in a fashion which does not produce a flammable gas mixture. The three methods examined where an empirical method based on pH measurements, a computational method based on known reactions of the species in the sludge and a titration based on neutralization of carbonate in the solution

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

    International Nucle